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+    "query": "Would it be possible to make a thermal reactor with graphite and lead?",
+    "answer": "Yes, it is possible to make a thermal reactor with graphite, though lead is not typically used as a primary component in such designs. Graphite serves as an effective neutron moderator in thermal reactors, slowing down fast neutrons produced during fission to thermal energies that are more likely to cause additional fission reactions. Several reactor designs successfully utilize graphite moderation, including the RBMK design where nuclear fuel rods are encased in chambers with graphite walls, Magnox reactors that were graphite moderated and CO2 cooled, and modern pebble-bed reactors (PBR) which are graphite-moderated, gas-cooled nuclear reactors. In pebble-bed designs, each fuel sphere contains the nuclear fuel, fission product barrier, and graphite moderator all integrated together. Nuclear-grade graphite requires high purity levels, good crystallinity, and thermal conductivity around 140 W/m/K at room temperature to function effectively as a moderator, though graphite components do experience dimensional changes and eventual degradation under neutron flux over the reactor's operational lifetime.",
+    "search_results": [
+        {
+            "title": "[PDF] Nuclear graphite for high temperature reactors.",
+            "url": "https://www.nrc.gov/docs/ML0117/ML011770379.pdf",
+            "content": "1116xl106 Thermal conductivity (kcal/m.h.\u00b0 C) = 0.-1116x 106 Resistivity(4uQcm) The importance of good crystallinity to the nuclear industry is that it gives an indication of: a) a high virgin thermal conductivity b) lower crystallite irradiation induced growth than would be the case in poorly graphitised material c) lower irradiation induced material property changes (this follows from above) d) high purity levels e) good machinability Nuclear graphite should have a thermal conductivity of around 140 W/m/K, when measured at room temperature.",
+            "raw_content": "Nuclear Graphite for High temperature Reactors B J Marsden AEA Technology Risley, Warrington Cheshire, WA3 6AT, UK Abstract The cores and reflectors in modem High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design, see Table 1. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable. The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modem HTRs. HTR Graphite Component Design and Irradiation Environment The details of the HTRs, which have, or are being, been built and operated, are listed in Table 1. A feature of the present designs is that to optimise the power output, an annular core is proposed. This annular core configuration tends to increase the dose to the graphite reflector. The fast neutron flux reduces exponentially with distance into the reflector thus increasing the thermal flux. Table 1 High Temperature Gas Cooled Reactors Reactor Type MW(t) MW(e) Helium Inlet Outlet Criticality Shutdown Pressure Temperature Temperature (bar) Dragon Prism 20 20 350 750 1966 1976 Peach Bottom Prism 115 40 24 340 715 1967 1974 Fort St, Vrain Prism 842 330 48 405 780 1974 HTTR Prism 30 40 395 950 1999 GT-MHR Prism 600 71 288 704 AVR Pebble 46 15 11 260 950 1967 1988 THTR Pebble 750 300 40 250 800 1985 1989 HTR-10 Pebble 10 30 300 900 PBMR Pebble 265 110 70 560 900 1 During reactor operation, neutron flux and thermal gradients in the graphite components, including the reflector, can lead to component deformations, bowing and the build up of significant shrinkage and thermal stresses. In addition the operation of these reactors at high temperature for many years could lead to degradation of the graphite material properties. The design of the HTR graphite cores must account for these thermal, irradiation conditions. Some of the issues are common to both the prism and pebble bed designs, others are specific to the type of reactor. These design issues are discussed below. HTR Core - Prism Design The prism design is best illustrated by the Fort St. Vrain design as described by Neheig, (1972). The main features are the permanent reflector, surrounded by boronated carbon shielding, a replaceable graphite reflector and the hexagonal graphite fuel elements. Permanent Reflector Although the irradiation fast neutron flux is significantly reduced from the peak value by the time it reaches the permanent reflector, these components must last for the life of the reactor. Therefore the dose the permanent reflector sees may be significant towards the end of life. The permanent reflector components have to be designed in such a way that the structure remains \"gas tight\" and that thermal and neutron streaming are minimised. This has to remain the case throughout life and during all thermal and pressure transients, for both normal and emergency operation. Most designs use graphite keying systems and dowels to keep the components located together. The columns of graphite bricks are usually free standing as individual columns of blocks. This is important as it avoids gapping due to differential thermal expansion and irradiation growth between columns arising from variations in material properties and irradiation behaviour that would occur in a \"brick wall\" type bonded system. It is important to provide side restraints for the columns using systems of springs or garters or other support systems. The first major issue connected with the restraint system is the low coefficient of thermal expansion of graphite compared with steel (-4 x 10-6 K-1 for graphite compared with -18 x 10-6 K1 for steel). The second is the large axial and radial temperature distributions within the graphite components that would cause the columns to bow outwards if no radial restraint were provided. The third is the irradiation shrinkage and growth that also leads to column bowing. As changes in the dimensions of graphite components due to thermal gradients and irradiated-induced dimensional changes cannot be prevented, the result of restraining the distortion leads to column kinking which could lead to gas leakage and thermal and neutron streaming if the design did not take this into account. It was small movements of the reflector and fuel element in Fort St. Vrain that led to redistribution of the coolant flow. These resulted in significant changes in fuel outlet temperatures and steam generator inlet temperatures that in turn caused power fluctuations that prevented the achievement of full power operation. To resolve this problem constraining devices were installed on the upper elements (Brey et al., 1982).\n2 In the prism type of reactor design, the permanent reflector receives much less dose than the replaceable reflector. For this reason lower grade graphites were used in Fort St. Vrain and in the HTTR. However, care must be taken in selecting these lower grade graphites as the higher levels of impurities may lead to an increase in decommissioning costs, as discussed later. Replaceable Reflector The replaceable reflector is normally constructed from blocks of graphite of similar dimensions to the fuel elements. The grade of graphite used in Fort St. Vrain was of a lower grade than that of the fuel. The side replaceable reflector elements were solid, but the upper and lower replaceable reflector had coolant holes and control rod holes. As these components have large flux, and possibly large temperature gradients, across their width and length it is possible that they will become significantly distorted and bowed after several years of operation. This may require prompt replacements, as it was bowing of the replaceable reflector and delaying its replacement that caused problems with fuel removal in the final years of the DRAGON project. In deciding on the graphite grade for the replaceable reflector, it is important not only to consider the implications for activation of the graphite, but also to consider the cost of obtaining irradiation data to sufficient dose for use in life extension and safety cases. This exercise may be so costly that use of good quality fuel element graphite for the replaceable reflector may be a cheaper option in the long term. For early permanent moderator prism reactor designs (no longer considered) Blackstone (1969) gives a peak dose of 200 x 1020 n/cm 2 EDND for a 30 year life time at temperatures between 450 8000C. Clearly a reflector element changed once during its lifetime could only see one quarter of this dose on the side against the fuel. Fuel Elements There have been various fuel element designs for prism fuel. The behaviour of the individual fuel particles and the compact material in which they are encased are not considered here. However, it should be noted that although the graphite technology associated with the particles and compact is related to that of the main moderator graphite there are differences as non-graphitised materials and possibly natural graphites are used in these fuel items. The graphite fuel moderator blocks remain in the reactor for much shorter periods than the reflector. However the temperature and flux that they see is more onerous. The blocks may contain passages for the coolant gas and the control rods, as well as holes for the fuel elements and possible burnable poisons. The large temperature and flux gradients between holes can give rise to significant shrinkage and thermal stresses leading to cracking. For this reason the design and through-life stressing of the graphite prism fuel blocks is important. Three -dimensional assessments are required due to possible end effects in the length of the blocks: Table 2 gives the doses and temperatures quoted for the lifetime of HTR prism fuel taken from various sources.\n3 Table 2 Doses and temperatures for HTR prism fuel.\nReactor Temperature Range 'C Dose x 1020 n/cm 2 EDND UK Mark II Fort St. Vrain Maxiumum0 1150 53.6 Fort St. Vrain Median# 700 -950 16.75 Peach Bottom- 400-800 30-40 (max) NP-MHTGR\" 200 and 1300 30 HTTR'\" 800-1000 10 Graphite data requirement for the whole of the reactor. Ishihara et al (1998) **. Everett et al. (1969) #Nehrig et al. (1972) Upper and lower graphite structures Fort St. Vrain is supported on keyed support blocks and post structures in the lower gas outlet plenum. However modifications were suggested for later designs to better resist seismic loading and thermal strains (Peinado, 1982). Above the core there is an upper plenum for the inlet gas, however there is no graphite lining for this structure. HTR - Pebble bed design A typical design of a pebble bed reactor is described by Lohnert and Reutler, (1982). Side reflector In the pebble bed design the main graphite component to consider from a life-time point of view is the side reflector only. In some earlier designs, as in AVR, there were also graphite \"noses\" which protruded into the core. The purpose of these noses was to house the control rods, allowing them to be positioned some distance into the core. These noses are no longer a feature of the latest pebble bed designs which have annular core configurations. One of the reasons for the removal of the noses from the design was the high dose environment that they would operate in would probably lead to the need to replace them several times during the reactor life. The replacement of the reflector in a pebble bed design would be a major undertaking. However, this is being considered in some of the latest designs. The following information is reported on the lifetime dose for pebble bed type cores: 1. From the Dragon programme 140.0 x 1020 n/cm2 EDND with a peak temperature of 9000C for a 30 year life (Blackstone, 1969) 2. For the AVR reflector 50 x 1020 n/cm2 EDND at 6500C, 12 x 1020 n/cm2 EDND at 1000I C (Haag et al, 1986) 3. For the later designs of German pebble bed reactors 300-400 x 1020 n/cm2 EDND at 730-8800C (Schmidt, 1979) In AVR the coolant flow was upwards with the return flow passing outside the reflector blocks as was the case in the THTR, where the flow was downwards. However, in the concept for the modular HTR (Lohnert and Reutler, 1982) very large blocks of graphite are used for the reflector, containing holes for the return coolant flow, control rods and small absorber spheres (KLAK system). The control rod holes are very close to the core boundary where the flux and temperature 4 gradients are high. This can lead to unacceptable shrinkage and stresses at these holes. Various ways have been investigated of overcoming this problem, including slitting the hole, through to the core, to avoid component failure (Schmidt, 1979). This may necessitate the use of graphite sleeves to prevent bypass of the control rod coolant gas. In addition graphite sleeves may be used in the return gas holes to prevent large temperature gradients and excessive cooling of the large graphite reflector blocks. Another feature of the blocks in the pebble bed design is saucer shaped indentations in the reactor side face of the lower core. These indentations act as \"disturbances\" to prevent bridging of the outer layers of fuel that may cause these layers to stay longer in the reactor than would be desirable. The blocks are joined together by a system of graphite dowels and keys and have similar restraint problems related to thermal expansion and dimensional change as discussed for the prism reactor core as discussed above. The high dose at the surface of the reflector block may cause some of the graphite to be irradiated through shrinkage and \"turn-around\" until the swelling is much greater than the original volume. At this stage the graphite structure will have started to disintegrate and it is possible that some of the graphite will become so friable that the fuel balls will rub the surface away. Several solutions have looked into solving this problem including sacrificial layers. Another way may be to return to the single zone core with noses and replace the noses at regular intervals. Lower core, inlet and outlet plenum The present modular pebble bed designs have both the inlet and outlet ducts at the bottom of the reactor pressure vessel. This arrangement avoids any chimney effect in the extremely unlikely event that the inlet/outlet duct should shear and allow air ingress. Although the irradiation dose is low at the bottom of the core, there may be large temperature gradients that have to be accounted for. In addition there are gas outlet holes, or slots, that may be subject to compressive loading from the pebbles. Deeper into the lower core structure the inlet and outlet plenum may pose design problems related to the temperature differences between the inlet and outlet gas. In addition the many paths through the graphite taken up by the inlet and outlet holes, the KLAK systems and the control rod holes can also lead to design problems. Reflector roof and upper inlet plenum The irradiation dose to the upper structure is low, however there may be stressing problems related to temperature gradients. The present pebble bed designs use graphite components for the roof structure and the upper plenum. Roof designs have been either cantilevers as in the HTR module or structures hung from upper structures by metal hangers as in THTR. Both designs have advantages and disadvantages related to component reliability, connectivity with the side reflector and the many holes required for the inlet gas and the control mechanisms. Graphite manufacture 5 The starting point in production of graphite is the selection of a suitable coke. These cokes are produced as by-products from the petroleum or coal industry or from naturally occurring pitch sources. These cokes vary considerably in their structure, size and purity. After production the cokes are broken up and calcined at temperatures between 900-1300TC to drive off volatile material and reduce the amount of shrinkage in the later processes. The calcined cokes are then crushed, milled and graded before being supplied to the graphite manufacturer. It is the choice of the particular coke size, purity and structure that decides the virgin and irradiated properties of the final product. A suitable blend of coke grades are then mixed with a binder, usually a coal tar pitch. In addition a crushed graphite flour may be added. The coke particles are often referred to as filler particles. The mixture is then formed into blocks often referred to as the \"green article\". Various methods of forming are used and the method chosen has an influence on the properties of the final product. The methods are discussed below: 1. The most common method of production is by extrusion. In this method the mixture is forced through a die under pressure. This method can be used to produce blocks of various sections and of reasonably long lengths. Blocks of the order of 500mm square by 3600mm long can be produced in this way. It is important that the extrusion pressure and rate is carefully controlled in order to maintain the desired quality. Graphites produced in this way have anisotropic material properties due to alignment of the filler particle grain, however it is possible to produced reasonably isotropic using this method. 2. Moulding or pressing. This method is used to produce a very isotropic product. The blocks are moulded or pressed from one or two directions at the same time. The AGR graphite moderator .blocks were produced using this technique. 3. Iso-static moulding is a more sophisticated method in which the coke and binder mixture is contained in a rubber bag and external pressure applied to give a uniform pressure from all sides. 4. Finally there is vibration moulding. In this method the graphite mixture is placed in a mould, which is vibrated to compact the mixture. Next the graphite mixture is pressed from one side and vibrated again whilst under load. There are other variations on these methods. Having formed the 'green article', which is reasonably soft, it is rapidly cooled by immersion in water. The green article is then baked at a temperature of around 800TC to drive off more volatile material and 'coke' the binder. To prevent oxidation, the blocks are encased in a granular packing, usually a coke. This allows for expansion and helps to support the shape of the green article. This is a long process and may take 30-70 days. One difficulty that arises at this stage is that the thermal conductivity of the graphite is very low, -30 W/m/K and on cooling thermal gradients in the blocks may lead to internal cracking. One method of overcoming this problem is to add crushed scrap graphite to the mix. However, this has implications for the irradiation behaviour of the final product as discussed later. The baking process will produce gas evolution pores throughout the structure as volatile gases are driven off. Much of this porosity will be open. To increase the density the baked blocks may be impregnated with a pitch under vacuum in an autoclave. This pitch is much less dense than the binder pitch. To help with this process the surface of the block may be broken by rough machining 6 or by grit blasting. This allows the pitch to enter the open porosity more readily. After impregnation the blocks are re-baked for a much shorter period. There may be up to four impregnations used; however the gain in density for each subsequent impregnation is much less. The product can now be regarded as carbon blocks, which can be used as an insulation material or furnace liner. However for this application they are usually baked at a higher temperature -1 100\u00b0C. The carbon blocks are now ready for graphitisation. There are two methods of graphitisation commonly used. The original method is to use an Acheson furnace. This is a large open furnace, which may be up to 7m wide by 20m long, into which the carbon blocks are stacked and covered in an electrical conducting coke. A large electric current is applied to each end of the bed through water-cooled electrodes and the blocks are taken through a temperature cycle to -3000'C. This process can take about 15 days. Another, more modem, quicker and cheaper method of graphitisation is to stack the carbon blocks in long lines so that they touch. Again the blocks are covered in coke to prevent oxidisation, but this time the current is applied directly through the carbon blocks and not through the packing material. This method can only be used for blocks of similar cross-sections. During this graphitisation period the graphite crystals are formed and the material becomes much softer and more easily machined. The electrical and thermal conductivity dramatically improve and many more impurities are driven off. If a more pure product is required the graphite blocks can be reheated to -2400'C in an Acheson furnace with a halogen gas passed through it. However, this final process could add up to 30% to the cost of the graphite. This process of graphite production is summarised in Fig. 1 Polycrystalline Graphite Microstructure As discussed above, the final polycrystalline microstructure is determined by the structure of the coke and the binder phase and also by the manufacture process. At the crystallite level the graphite has strong hexagonal basal planes with much weaker bonding between the planes. For perfect graphite crystals the 'd', or inter-layer, spacing has been measured to be 3.3539 A (0.33539 x 10-9m) with an 'a' spacing of 2.46 A (0.246 x 10m). The 'c' spacing is twice that of the 'd' spacing. The size of the crystallites can be measured by x-ray diffraction and has been found to give values of between 400 and 800 A for La and L, is well graphitised material (Reynolds, 1968). In the 'c' direction the size of the crystallite is limited due to so call \"Mrozowski\" cracks which are formed during cooling from the graphitisation temperature (,-3000' C), see Fig. 2. The mechanism, which leads to the formation of these cracks, is due to the large difference in the coefficient of thermal expansion in the two crystallographic directions (a. =-26.5 x 10-6 K\"1, % =---1.5 x 10-1 K7 1). At a temperature of around 1800*C the structure hardens and the much larger shrinkage in the 'c' direction coupled with the restraining affect of the rest of the structure leads to horizontal cracking in the basal planes. Various estimates of the size of Mrozowski cracks have been given, however, in practice there are probably cracks of a variety of widths ranging from less than 250 A upwards.\n7 It is this cracked structure that gives graphite its good thermal shock resistance, allowing large crystal expansion in the 'c' direction without leading to inter-crystalline cracking. These cracks also provide accommodation spaces that can be taken up by irradiation-induced crystal growth and play an important role in determining component property changes in reactor. Beyond this scale, the cracked crystallite structures are jointed together and follow the general shape of the coke particle. There are also many larger cracks and fissures which also tend to follow the coke particle shape. Examples of this are illustrated in Figs 3. It can be envisaged that the general alignments of the 'a' axis is with the \"flow\" of the coke particle with the 'c' axis perpendicular to this direction. It can also be envisaged that the shape, size, distribution and orientation of the coke particles will strongly influence the material property of the final graphite component. During manufacture the graphite coke particles may have been ground and blended to give a uniform mix. They may be lenticular or needle like in shape or in some special cases spherical. In the UK the Pile Grade A graphite, used in the Magnox reactors was manufactured from a needle coke. Graphite produced from these needle type coke particles is very anisotropic with an anisotropy ratio in the region of 2. The most famous of the graphites with spherical structures is Gilsocarbon graphite which as manufactured from naturally occurring pitch found in remote parts of Utah in the USA. These coke particles formed spheroids with the 'c' crystallographic direction lying mainly in the radial direction and the 'a' crystallographic direction mainly in the hoop. Graphites formed from Gilsocarbon coke had semi-isotropic properties. Some modem graphites are manufactured from finely ground coke particles, this also can produce semi-isotropic properties. The coke particles are usually bound together using a pitch binder. The binder itself will probably be mixed with a \"flour\" consisting of finely ground coke and possibly scrap graphite. The baked structure may have also have been impregnated once or twice to increase the overall density. Features that are of interest are the randomly ordered, well-structured, small particles and the large gas evolution pores. The orientation of the coke particles in the bulk product is strongly influenced by the forming process. Many types of graphite are manufactured by extrusion which tends to align the coke particles. Other processes described above that can give more isotropic structures are compression moulding and isostatic moulding. Graphite requirement specification Density For a given amount of output, the higher the density of the moderator, the smaller the volume of the core, thus a high-density graphite is desirable. The theoretical density of graphite crystals is 2.265 g/cm 3, however polycrystalline graphites have a much lower density due to inter-crystalline porosity. Early nuclear graphites had typical densities of 1.6 g/cm 3, the second generation of graphite moderated reactors used graphites with densities around 1.72 g/cm 3 and the third generation 1.82 g/cm 3. Modem nuclear graphites have not improved over this latter figure.\n8 The importance of crystallinity In theory, from a nuclear point of view, any carbon could be used as a moderator if it could be packed to the required density in the required nuclear configuration. However, the availability, and structural properties, of relatively pure artificial graphite led to its use as a moderator in the first graphite moderated nuclear reactors. These artificial graphites were essentially electrodes used in the steel industry and their purity relied on the choice of raw materials (filler coke and binder) and the manufacturing process and heat treatments to drive off unwanted volatile impurities. The method used by the graphite manufacturers to determine the degree of crystallinity (or graphitisation) is to measure the electrical resistivity. At room temperature the electrical resistivity can be directly related to the thermal conductivity by the Wiedemann Franz law: 0. 1116xl106 Thermal conductivity (kcal/m.h.\u00b0 C) = 0.-1116x 106 Resistivity(4uQcm) The importance of good crystallinity to the nuclear industry is that it gives an indication of: a) a high virgin thermal conductivity b) lower crystallite irradiation induced growth than would be the case in poorly graphitised material c) lower irradiation induced material property changes (this follows from above) d) high purity levels e) good machinability Nuclear graphite should have a thermal conductivity of around 140 W/m/K, when measured at room temperature. Dimensional Stability In addition to good crystallinity, polycrystalline graphite components have another requirement to ensure dimensional stability. This is related to the way the randomly orientated crystals in a porous material interact as they grow and shrink with irradiation. Simmons (1965) demonstrated that there was an empirical relationship between dimensional change rate and the Coefficient of Thermal Expansion (CTE) (The higher the coefficient of expansion the less the dimensional change rate). He also derived a theoretical relationship between dimensional change rate and the coefficient of thermal expansion (Marsden, 1998). However this relationship breaks down at comparatively low doses. It was also found that extruded course grain graphites that have anisotropic material properties tended to have lower CTEs than more isotropic graphites. However, it was also observed that some isotropic graphite with very high CTEs expanded with irradiation. From experience it is desirable to choose a graphite with a coefficient of expansion between 4.0 and 5.5 x 106 Kl' measured over the range 20-120'C. Air reactivity (Thermal oxidation in air) 9 In the unlikely event of an incident involving air ingress, it is important that the air reactivity, that is the rate at which graphite can oxidise in air, is as low as possible. Irradiation increases the air reactivity rate in graphite, however this effect is minimal compared to the increase in air reactivity caused by catalytic impurities. For the AGRs the mean reactivity in air of samples was specified as not exceeding 3 x 10-6 g/g/k at 400 TC (Hutcheon and Thorne, 1965). Absorption cross-section For the particular HTR design the nuclear physics considerations will lead to the values of absorption cross sections required. However, historically there has been no systematic effort put into determining a specification for the absorption cross-section in graphite. The early plutonium production reactors built in the UK used various graphites with nuclear abortion cross-sections of between 4.7 and 5.1 mbam. The later military and civil carbon dioxide cooled reactors used Pile Grade A (PGA) graphite with an absorption cross-section of 4.0 mbam. For this reason an upper limit for boron of 0.2 ppm was specified. The method used to measure the absorption cross section was to place graphite samples, or whole blocks in some cases, in a test reactor such as GLEEP in the UK. Later when the AGRs were designed it was recognised that the major impurity atom contributing to the absorption cross section in graphite was 1 0B which burnt up rapidly early reactor life. For these later reactors the absorption cross-section was calculated from the chemical inventory, however the value arrived at depended on what impurities were measured and the choice of measurements was somewhat arbitrary. As previously discussed it is clear that in the past no systematic effort has been made to define a specification related to the impurity levels related to absorption cross-section. The author considers that a list of elements with significant absorption cross-sections should be identified as part of a graphite specification. The graphite manufacturer could then be asked to analyse their product for these elements. Impurities related to operational and decommissioning problems The radioactivity associated with graphite components arises from initial impurities and from subsequent contamination within the reactor circuit. Probably the most important isotopes related to initial contamination will be 60Co, 1 54Eu, 3H, 36C1, 41Ca and 1 4C , however this is not an exhaustive list and a more rigorous approach is required as discussed below. An IAEA TechDoc (Marsden, 2001) is being prepared at present on the subject of decommissioning graphite reactors which gives much more detail concerning the problems related to disposal of nuclear graphite. Releases during operation 10 In the UK, graphite reactors such as the AGRs and Magnox have to monitor the release of certain isotopes. Some of these radionuclii, such as 36C1, 14 C and 3H, may be related to the release of products from the graphite core. Again a definitive list of undesirable isotopes, which may be released from graphite, is required, taking account of regulations in the country where the reactor is to be constructed. The likely origins of these radionuclii can then be determined to give a list of undesirable impurities from an operational point of view. Information will also be required as to how easily these radionuclii can be released from the graphite structure during operation. Decommissioning Similarly a list of the most undesirable radionuclii with regard to the long-term risks associated with waste disposal can be obtained from the appropriate authorities, such as NIREX in the UK. Then the route that can produce these radionuclii can be determined to give a comprehensive list of undesirable elements from a decommissioning point of view. Strength In a modem reactor, the strength of the graphite is important ,as it may be subjected to shrinkage and thermal stresses, as well as restraint loads and possible seismic impact loads during the life of the reactor. Irradiation modifies graphite strength, as does thermal and radiolytic oxidation. Graphite is stronger in bend than tension and stronger in compression than bend. As with many brittle polycrystalline materials, the failure strength depends on the component geometry, loading configuration and component size. Unfortunately at present there is no satisfactory failure model for unirradiated, or irradiated, graphite which makes it difficult to predict the behaviour of graphite components in the future. Proposed Specification for HTR Graphite I. The graphite should be reasonably dense -1.8 g/cm 3. 2. It should be well graphitised as indicated by a thermal conductivity of -145 W/m/K measured at room temperature. 3. It should have a low absorbtion cross-section, between 4 and 5mbams. (This can be calculated from knowledge of the chemical impurities). 4. Impurities that could possibly lead to operational problems and high decommissioning costs must be kept to a minimum. 5. The graphite must have a high irradiation dimensional stability. This is indicated by a relatively high CTE(20-120\u00b0C) between 4.0 and 5.5 x 10-6 K71. 6. The irradiation time, over the irradiation temperature of interest, for the graphite to return to its original volume should be as long as possible. (In the long term a Material Test Reactor (MTR) programme can only confirm this.) 7. The graphite must have a moderately high strength (A tensile strength of about -20 MPa). 8. The air (moisture) reactivity should be measured to ensure that the rates are acceptable (0.2 to 0.01 mg/g-h). 9. A suggested list of chemical impurities that should be minimised is given in Table 6. This list is based on past experience and should be reviewed in the light of local requirements for decommissioning and operation. Table 6 Graphite impurities that are considered to be incompatible for reactor operation and final decommissioning and disposal 11 Element Symbol Element Symbol Aluminium Al Mercury Hg Antimony Sb Manganese Mn Arsenic As Molybdenum Mo Beryllium Be Nickel Ni Barium Ba Chlorine Cl Boron B Potassium K Bismuth Bi Phosphorous Pb Cadmium Cd Platinum Pt Caesium Cs Selenium Se Calcium Ca Samarium Sm Chromium Cr Silver Ag Cobalt Co Silicon Si Copper Cu Sodium Na Gold Au Tantalum Ti Indium In Tin Sn Hafnium Hf Sulphur S Lead Pb Titanium Ti Dysprosium Dy Tungsten W Europium Eu Vanadium V Iron Fe Zinc Zn Gadolinium Gd Strontium Sr Lithium Li Halogens Magnesium Mg Rare earth metals Conclusions Unirradiated and irradiated graphite material properties depend strongly on the choice of the raw materials and the manufacturing process. Suitable graphite for modem HTRs can be designed based on the choice of coke, binder and manufacturing technique. However a compromise is always necessary. A suggested specification for HTR graphite is given. References Blackstone R, Graham L W, Everett M R, Delle W. Irradiation data on Gilsocarbon graphites for high temperature nuclear reactors. DP Report 646. April 1969. Brightman F G. Development of sampling and assay methods for WAGR radwaste. CEC Report EUR 13254, 1991 Brey H L, Olson H G. The Fort St. Vrain quality assurance programme. Gas Cooled Reactors Today, BNES Conf. Bristol, Vol. 1, pp. 103-105, 1982. Brey H L. Fort St. Vrain experience. Gas Cooled Reactors Today, BNES Conf. Bristol, Vol. 1, pp. 35-39, 1982. Engle G B. Assessment of Grade H-451 graphite for replaceable fuel and reflector elements in HTGR. GA-A14690, UC-77, 1977. Everett M R, Blackstone R, Graham L W, Manzel R. Graphite materials data for high temperature nuclear reactors. DP Report 699, 1969 12 Haag G, Delle W, Kirch N, Nickel H. Results of visual in-pile inspection of the inner graphite reflector of AVR. IAEA Specialist's meeting on Graphite Component Structural Design. Japan, July 1986 Hutcheon J M, Thome R P. Improved graphite for nuclear purposes. Second conference on Industrial carbon and graphite. Soc. Chem. Ind. pp. 441-445, London 1965. Ishihara M, Hanawa S, Iyoku T, Mogi H. Study on structural integrity of core graphite components in nuclear reactor. Int. Sym. Carbon 1998, Tokyo. Lohnert G H, Reutler H. A new design of a HTR pebble bed reactor. Gas Cooled Reactors Today, BNES Conf. Bristol, Vol. 1, pp. 265-269, 1982. McKee D W. The catalysed gasification reactions of carbon. Chemistry and physics of carbon. Vol. 16, 1965 Marsden B J, Arai T, Hopkinson K L. Validation of a model on irradiation induced porosity evolution in gas cooled reactor graphites. AEAT-3212, 1998 (unclassified). Marsden B J. Characterisation, treatment and conditioning of radioactive graphite from decommissioning of nuclear power plants. IAEA TechDoc. To be published 2001. Nehrig D A, Neylan A J, Winkler E 0. Design features of the core and support structures for the Fort St. Vrain nuclear generation station. Graphite Structures for Nuclear Reactors. Inst. Mech. Eng. March 1972. Peinado C 0, Wunderlich R G, Simon W A. HTGR nuclear heat source component design experience. Gas Cooled Reactors Today, BNES Conf. Bristol, Vol. 1, pp. 107-112, 1982. Reynolds W N. Physical Properties of Graphite. Elsevier 1968. Schmidt A. Analysis of the stresses in a side reflector block of the Pebble reactor. IAEA Specialist meeting on mechanical behaviour of graphite for HTRs. France June 1979. Simmons J H W. Radiation damage in graphite. Pergamon Press. 1965. White I F, Smith G M, Saunders L J, Kaye C J, Martin T J, Clarke G H, Wakerley M W. Assessments of management modes for graphite from nuclear reactor decommissioning. EUR 9232 En., 1984 13 FIG. 1 GRAPHITE MANUFACTURING PROCESS 13000C MIXED EXTRUDED OR MOULDED COOLED BAKED AT - 10000 C IMPREGNATED WITH PITCH FURTHER IMPREGNATION AND BAKING AS REQUIRED GRAPHITISED AT -2800 0C 14 MICROCRACKS IN THE GRAPHITE CRYSTALLITE STRUCTURE 'LATIL 7. Electron micrographs of reactor graphite by the replica method, (Magnilication x 5,000) (After Thrower and Reynolds. 1963) (a) Unirradiated.\n(b) Irradiated at 200-C to a dose yj = 2-7 x 1020.\n(c) Irradiated at 200-C to a dose y\u2022 = 25 x 10'\".\n15 FIGURE 2 FIGURE 3 MICROGRAPH OF PITCHCOKE GRAPHITE 16"
+        },
+        {
+            "title": "[PDF] Graphite Technology Course. - Nuclear Regulatory Commission",
+            "url": "https://www.nrc.gov/docs/ML0120/ML012080125.pdf",
+            "content": "The principal difficulty in the use of graphite as a nuclear reactor moderator material stems from the large and continuing changes in dimensions and physical",
+            "raw_content": "GRAPHITE TECHNOLOGY COURSE C_ Dr TB -Marsden :7 \"4 c~r , ... *......': s contact: \u2022 - .\n.. -' .,\": .\nGRAPHITE TECHNOLOGY COURSE LECTURE 1 1. Introduction and Aims of the Course The principal difficulty in the use of graphite as a nuclear reactor moderator material stems from the large and continuing changes in dimensions and physical properties which occur die to the combined effects of radiation damage and radiolytic weight loss that are suffered by the material during long-term reactor exposure. These changes can be severe and are difficult to predict with certainty for the designed lifetime of modem reactors. The nature of the changes is such that they lead to problems in attempts to design stable and long lasting moderator structures, and their structural effects may become life-limiting for some moderator designs. Assessment of the effects of dimensional and physical property changes on the integrity of graphite moderator structures has thus become an important factor in the preparation of safety cases that are written in support of moderator design, reactor operation and life extension. In order to carry out such assessments it is necessary to have a good basic understanding of the phenomena of radiation damage and radiolytic weight loss in graphite, and the various complex factors which influence the resultant dimensional and physical property changes they induce in the material. Secondly, it is necessary to be able to understand the extensive database which is available on graphite behaviour under irradiation, and in particular to know how to interpret that data correctly for the specific type of graphite, radiation conditions and physical environment appropriate to each application. Thirdly, it is essential that the correct methodology should be fully understood and properly applied throughout the assessment process used to evaluate all parameters such as radiation damage dose, dose rate, equivalent temperature, radiolytic weight loss, dimensional property change and physical property changes as a function of all relevant basic reactor parameters. Finally, it is essential to know how to apply the evaluated dimensional and physical property changes in the assessment process, and to have a basic understanding of how these changes may affect moderator structural integrity. \"One of the aims of the course will be to give those attending a fundamental understanding of the basic principles, data and methodology used in the assessment of graphite moderator structures including the evaluation of: (i) Radiation damage dose and dose rate. (ii) Equivalent temperature. (iii) Dimensional changes of PGA and Gilsocarbon graphites. (iv) Physical property changes of PGA and Gilsocarbon graphites. (v) Radiolytic weight loss. (vi) Effects of radiolytic weight loss.\nI The course is confined to Magnc'x and Advanced Gas-cooled reactors and will introduce members to graphite moderator structural design of both types. Important aspects of the assessment of the effects of in-pile exposure on graphite moderator components and structural integrity will also be discussed, including such effects as the evaluation of thermal and 'shrinkage stresses, brick distortion and structural loading. The use of computer codes in the assessment process will be addressed. 2. Graphite Manufacture The early nuclear power reactors (Magnox) were constructed from an anisotropic extruded pure petroleum coke graphite known as Pile Grade A (PGA). The irradiation conditions in the later Advanced Gas-cooled Reactors precluded the use of PGA and a new material was developed by the manufacturers based on a specification produced by the UKAEA. This material was an isotropic moulded graphite manufactured using a coke prepared from Gilsonite, a naturally.occurring asphalt mined in Utah, USA. This improved graphite, which has better all round properties and dimensional stability under fast neutron irradiation, is known as Gilsocarbon graphite. The properties of PGA and Gilsocarbon graphites are shown in Table 1.1 Fig 1.1 illustrates the process of graphite manufacture. Coke may be derived from naturally occurring asphalt or as a by-product of the oil refining process and may be used to make nuclear graphite providing that its purity is acceptable. The coke is first calcined, that is heat treated to 1300'C to remove residual volatile hydrocarbons, and is then crushed, ground and blended to give a specific particle size distribution. The coke particles are then mixed with a pitch binder and then either extruded or moulded to the required shape. The \"green\" article is then baked at a temperature of about 1000*C to carbonise the pitch binder and then impregnated under pressure with a suitable pitch to increase the density. The brick is then graphitised by heating electrically in a furnace packed with coke dust and sand to a temperature of about 2800\"C, taking three or four days to reach temperature, a day or so at temperature and about 14 days to cool. Crystal development takes place during graphitisation and as a result its pro:perties change markedly, particularly with respect to ease of machining and improved conductivity. The purity of the product is also improved since many of the contaminants are volatilised out at these temperatures. The brick may receive further impregnation and graphitisation steps to improve density, strength, etc. The extrusion process used in the manufacture of PGA graphite tends to align the needle shaped petroleum coke particles parallel to the direction of extrusion which results in a highly-orientated structure. Furthermore, the crystallographic layer planes of the eventual graphite crystallites tend to lie parallel to the long aixs, thus producing a product having anisotropic properties, ie properties measured parallel to the extrusion direction differ from those perpendicular to it. Gilsocarbton graphite on the other hand, being moulded and made from coke containing spherical structures in which the graphite crystallites tend to lie in an \"onion skin\" structure, has very isotropic properties. Figures 1.2 and 1.3 illustrate these differences.\n2 3. Radiation Damage and Annealing A neutron-induced fission in nuclear reactor fuel produces two to three new neutrons with energies which occur over a wide range. The spectrum of fission neutrons peaks at an energy of about 2 MeV and extends to energise greater than 10 MeV. The role of the graphite moderator is to slow these neutrons down to thermal energy levels by collisions between the neutrons and the nuclei of carbon atoms. In a collision between a neutron and a carbon nucleus a substantial fraction of the neutron energy is transferred to the nucleus. The binding energy of the carbon atom in the graphite lattice is about 5eV, and, if the energy transferred is only a few times greater than this, it is irreversibly displaced from that site. The primary displaced carbon atom will collide with other carbon atoms causing them to be displaced also, and so on, thereby creating a displacement cascade, Fig 1.4. A graphite crystallite-consists of layers of carbon atoms with the layers stacked one upon the other, and with the carbon atoms in each layer located at the vertices of a hexagonal network, Fig 1.5. Displaced atoms may move to positions between the layers, known as \"interstitial\" positions, leaving behind them holes or \"vacancies\". The interstitial atoms group together to form interstitial loops which push the atomic layers apart, causing crystal growth perpendicular to the layers. The vacancies form vacancy lines which cause the atomic layers to contract in the plane of the layers, causing crystal shrinkage parallel to the layers. Fig 1.4 illustrates the damage process. The large majority of displaced atoms find their way back to normal lattice positions, but a significant proportion form defects which change the properties of the graphite in various ways. Some of the displaced atoms return by a process of annealing, the number of atoms so affected being a function of both irradiation temperature and duration. Thus, a given fast neutron damage dose will produce more net damage if the irradiation temperature is reduced or the dose rate is increased (ie a given dose is received over a shorter period of time). As will be seen later, this aspect of radiation damage is particularly important in both the interpretation of data and in its application during the assessment process. 4. Damage Function The damage received by an element of graphite from an irradiation source is a function of a source strength, the distance from the source, and the attenuation in damage with distance through intervening graphite. The attenuation of damage through graphite was first measured by Kinchin(\" in the graphite reflector of the BEPO reactor at Harwell. The resultant curve of relative damage through graphite is shown in Fig 1.6. The curve permits damage in any graphite lattice to be computed providing that an appropriate adjustment is made for differences in graphite density between BEPO (1.6 g/cm3) and the reactor of interest. Subsequent Monte Carlo calculations by Boden and Russell(2) have led to refinements of the Kinchin curve, and Fig 1.7 shows the damage absorption curve currently recommended for use in Magnox reactors, this curve also being based on a graphite density of 1.6 g/cm3. The curve may be used to derive the value of an integrated damage function at any position in a Magnox reactor core taking account of varying source strengths, the distance from the various sources and the thickness of intervening graphite. The damage function so derived 3 enables the damage dose received by the graphite in a given time to be calculated, and this dose may in turn be used to evaluate the effect of fast neutron irradiation on the component in question. The calculation of damage function may be performed by hand calculation or by computer calculation using a code such as GRAFDAM which incorporates the damage -absorption curve shown in Fig 1.7. The damage function, df, for a single uniform line source is defined as: df = (4(Rg)/R) (1) where Rg equivalent thickness of BEPO graphite (density 1.6 g/cm3) between the line source and the target, ins. R -distance between line source and target, ins. ck(Rg) = damage absorption curve value from Fig 1.7 corresponding to intervening graphite thickness Rg. For multi-source situations with sources of differing strengths, = \u20224{(R)i) (2) where Bi -accumulated fuel bum-up of the it source, MWd/t. B = accumulated fuel burn-up of a reference source, MWd/t. 41)(Rg)i= damage absorption curve value from Fig 1.7 corresponding to intervening graphite thickness Rg for the ith source. Ri = distance between ith source and target. The same equation may be used to calculate an integrated damage function value in AGRs, but here different damage absorption curves have been derived to represent the multi-pin fuel configuration of these reactors. Damage functions for AGRs are calculated by the FAIRY code using two different damage absorption curves, one representing the damage contribution from single pins within a fuel cluster contained within the target graphite brick, and another representing the contribution from all relevant channels outside the target channel, each fuel cluster being treated as a single line source located at the channel centre. Graphite damage 4 kernels for both source types, which may be used as input to FAIRY, have been derived by Boden and Russell\"). 5. Equivalent Rating and Dose In order to correlate data obtained from several different experimental reactors, and to apply. these data to the assessment of power reactors having various geometrical lattice arrangements and operating parameters, it is convenient to convert all fuel ratings and integrated doses to a common standard. The standard first adopted in the UK was the Calder equivalent rating and dose. The Calder equivalent rating is defined as that fuel rating, Pe, in a Calder reactor which would produce the same rate of production of damage at a standard position as would be produced at the given position in the reactor under consideration where the fuel rating is P. The standard position chosen in Calder is the wall of a fuel channel 3.55 in diameter, in a lattice having a uniform square pitch of 8.0 in, containing fuel elements with a natural uranium rod diameter of 1.15 in, and at a point which lies on the line joining the centre of the fuel channel to its closest neighbour. In reference 3, it is shown that: Pe = A . df. P (3) A c4,. - df j,., where Pe = Calder equivalent rating, MW/t P = total local fuel rating in the reactor under consideration (ie including that proportion of channel heat output which appears in the graphite), MW/t A = uranium cross-sectional area per channel in the reactor under consideration, in' Ac.,\u2022 = uranium cross-sectional area per channel in Calder (= 1.04in2) df = damage function at the chosen point in the reactor under consideration (obtained from equation 2 above) dfcwr = damage function at the standard position in Calder (= 1.395) 5 Substituting values gives: Pe = 0.689. A . df. P (4) In reactors where fuel materials other than natural uranium are used it is necessary to apply a correction factor to the above relationship to allow for the fuel density change. Equation 4 then becomes: Pe = 0.689.A.df.P. Pu (5) 18.78 where pu \" grams of uranium in a cubic centimetre of fuel rod material in the reactor under consideration 18.78 = density of natural uranium, g/cm3. For Magnox reactors, pu is taken to be 18.78 g/cm3, whereas for AGRs, pu has a recommended value of 9.388 g/cm. Calder equivalent dose is obtained by multiplying rating values by the irradiation time in days. Thus, De = 0.689. A . df . B Pu (6) 18.78 where De = Calder equivalent dose, MWd/t (. B = accumulated fuel bum-up in reference fuel channel, MWd/t. It is current practice to convert Calder equivalent doses to equivalent DIDO nickel doses (EDND, n/cm2) since all graphite data are now presented as a function of EDND, largely because the majority of high dose data were obtained in the DIDO experimental reactor facility where damage fluxes and doses were measured by nickel foil activation. The conversion is made using the relationship(). EDND = 1.0887 x 10'7. De (7) 6 Substituting in equation (6) gives: EDND = 3.9942 x 1015 .A. df .B. pu (8) 6. WIMSE Code The evaluation of damage doses (EDND, n/cm2) using the damage function routines incorporated in GRAFDAM and FAIRY has now been superseded by use of the Winfrith code WIMSE which is capable of calculating damage doses directly. The WIMSE code is a powerful collection of methods for carrying out within core calculations on any thermal reactor type by solving the Boltzmann transport equation for the neutrons in the core. This includes all graphite moderated reactors. The code is the reference code for lattice cell calculations for all thermal reactor core calculations in the UK. There is a variety of methods available within the code varying from simple modelling of fuel geometry in which the fuel is smeared out into a single central zone through to explicit modelling of the fuel pins, graphite sleeve and the graphite brick. All methods are linked to the same data library and this contains a response function for damage. Using this response function users can obtain damage anywhere in the structure. Calculations may consist of a single fuel channel and its associated graphite or an array of channels can be used. The choice of model is determined by the user. 7. Calder Effective Dose During the early 1960's experimental data on stored energy and thermal conductivity changes were rather limited, and it became necessary to extrapolate the available data in order to make predictions for the higher irradiation doses appropriate to the Civil Magnox reactors. In order to make the required extrapolations it was necessary to establish a suitable correlation between the relevant parameters of dose, temperature and property change. It was well known that damage accumulation occurred at a slower rate when the irradiation temperature was increased, and a detailed study indicated that the data at different temperatures could apparently be made to correlate with a single curve if the Calder equivalent dose was reduced by a factor R(O) which was temperature dependent. Thus, high irradiation temperatures required low values of R(O) to reflect the slower rate of damage accumulation, and a single curve of property change versus Calder effective dose was produced in which the latter is defined as: Calder effective dose = Calder equivalent dose x R(O) where R(6) is a function of Calder equivalent temperature as shown in Fig 1.8. The assumption was made that a given property change would be reached regardless of temperature providing the irradiation damage dose was sufficiently high. This assumption is almost certainly pessimistic, particularly for higher irradiation temperatures where saturation in the build-up of stored energy and thermal conductivity changes appear to occur. Nonetheless, Calder effective dose has proved to be a very useful tool for correlating certain 7 property changes at low irradiation temperatures and its use persists to the present time. For most purposes, however, more reliable predictions can be obtained using the current improved data which are presented as curves of property change at various DIDO equivalent temperatures as a function of DIDO nickel dose. 8. Equivalent Temperature It is now necessary to consider the net rate of production of damage in the graphite, ie the difference between the rate of production of damage and the rate of annealing of damage. The rate of annealing of damage increases with increase in temperature, but at any given temperature the amount of damage which is annealed decreases with increase in the rate of production of damage. In other words, annealing of damage is a function of both temperature and irradiation time:, and if either of these parameters is reduced then less annealing of damage will occur. In c-der to correlate data from a number of sources having widely different rates of production of damage and irradiation temperature, it is convenient to establish a standard rate of production of damage and to evaluate an equivalent temperature at which the graphite must be irradiated at this rate in order to produce the same net damage for a given total production of damage. The reference first chosen in the UK for this standard damage production rate is that rate of damage which wouldc be produced at the standard position in Calder when the local fuel rating is 3.12 MW/t. The Calder equivalent temperature is then defined as the temperature in a Calder reactor which, for a standard fuel rating of 3.12 MW/t would result in the same net damage at the standard position (see Section 5) as would be produced at the given position in the reactor under consideration, for the same total production of damage. The relationship is governed by the equation: I 1 k log\" (Pe (9) Te 77 E 3.12 where Te = Calder equivalent temperature, K Ti = irradiation temperature in reactor under consideration, K k = Boltzmann's constant (= 8.617 x 10. eV/K) E -Activation energy (usually taken as 1.2eV for low dose effects such as initial increase in Young's modulus and thermal resistivity, but a value of 3.0eV is considered more appropriate when the effects of high dose structural changes become important). Pe = Calder. equivalent rating at selected point in reactor under consideration,, MW/t.\n8 From equation 9 it can be shown that values of Pe greater than 3.12 MW/t would necessitate irradiation in Calder at a lower temperature than Ti in order to achieve the same net damage, and vice versa. As in the case of equivalent doses (see Section 5) it is now more usual practice to refer equivalent temperatures to a standard position in DIDO where the Equivalent DIDO Nickel. Dose Rate is 4 x 10 3n/cm2.s. This is again because the large majority of high dose data has. emanated from DIDO, and the resultant database uses DIDO equivalent temperatures calculated at this standard position in DIDO. The appropriate relationship for calculating DIDO equivalent temperatures is: S 1I - k loge 03 (10) 0 771 E 4\u00d71t where 0 = DIDO equivalent temperature, K 4) = equivalent DIDO nickel dose rate at selected point in reactor under consideration, n/cm2.s. Ti, k and E are as defined in equation (9). 9. Important Reactor Parameters The damage dose (De, MWd/t) or (EDND, n/cm2) should be calculated using equations (6) and (8), respectively, when the reference fuel burn-up (B, MWd/t) is calculated as follows: (a) For predicting material property changes, self-stress and cross-sectional dimensional changes: B Bt where B t = local fuel bum-up over life (b) For predicting column length changes: B = B.,. L/L, where B = mean fuel burn-up in channel over life Lf = total length of fuel in channel L = length of column in active core (c) For predicting brick length changes: B = Bmb. Lf'/Lt 9 where Bin, = mean burn-up of fuel located in brick bore Lr' = length of fuel within brick .\n= length of brick References 1. Bell J C, Bridge H, Cottrell A H, Greenough G B, Simmons J H W and Reynolds W N. Phil Trans Roy Soc A254, p361 (1962). 2. Boden W A and Russell. P D D. Mk II graphite damage in a flux gradient. APC/R1424 (1972). 3. Prince N. The calculation of stored energy and thermal conductivity changes in graphite due to neutron irradiation. TRG Report 388(R). 4. Compendium of CAGR core and sleeve data and methods. CSDMC/P28.\n10 Lecture 1 Table 1. 1 Cnrnnarison nf\" 1\u2022'ir-tie\" ni' ThI\u2022 (r~ar1e A an ;ic +\u2022hn ('r.aht Property Units Pile Grade A Gilsocarbon (Hexagonal symmetry) graphite (Isotropic) g. cm\" Open pore volume Closed pore volume 3 -3 cm cm cm3.cm-Thermal Expansion Coefficient K-1 (20-120\"C) Thermal conductivity w.m'\u2022K' (20\u00b0C) Young's Modulus (20-C) Strength Tensile Bend GNm-MN.m'2 MN. m-2 Compression MN. m\" Poisson's ratio Electrical resistivity A ohm.cm\" Diffusivity Permeability Viscous flow coeff B. m Slip flow coeff K. m 1.74 0.198 0.01 Parallel to extrusion 0.9x10\" Perpendicular to extrusion 2.8xl0\" Parallel to extrusion 200 Perpendicular to extrusion 109 Parallel to extrusion 11.7 Perpendicular to extrusion 5.4 Parallel to extrusion Perpendicular to extrusion Parallel to extrusion Perpendicular to extrusion Parallel to extrusion Perpendicular to extrusion 17 11 19 12 27 27 Multiple values -0.07 Parallel to extrusion 620 Perpendicular to extrusion 1100 13.6x10\"' Parallel to extrusion Perpendicular to extrusion Parallel to extrusion Perpendicular to extrusion 712x101 6 147x10\"' 6 108x109 21xt10' 1.810 0.11 0.086 4.3x10\" 131 10.85 17.5 23.0 70.0 0.21 900 3x10* 3 6.5x10 5 5.8x10\" 14\u0002 = These properties are representative of the materials but do not represent mean values. GRAPHITE TECHNOLOGY Density Lccturc I GROUND & BLENDED CMIXED SCOOLED SEXTRUDED OR MOULDED TICLE1 CBAKED AT = 10000C ,TICLE C> IMPREGNATED WITH PITCH T C> GRAPHITISED TO = 28000C [GRAPHITE S>FURTHER IMPREGNATIONS AND GRAPHITISATIONS IF REQUIRED Figure 1. Materials and Process used in Graphite Manufacture GRAPHITE TECHNOLOGY '.,,tt r; I -26' --~ --...\nI-Figure 2.\nFigure 3.\nPolarised Light Micrograph of British Reactor Grade 'A' Graphite Cut Parallel to Extrusion. (Scale 185 microns to 1 cm) Polarised Light Micrograph of IM1-24 (Gilso Carbon) Graphite. (Scale 185 microns to 1 cm) ... ...\nITE ........ GRAPHITE TECHNOLOGY Lccturc I 0 Vacant lattice site X Interstitial atom x x X Neutron Displacement cascade (1) Sub-microscopic cluster of 4\u00b12 atoms ----Layer planes Dose increase (2) Interstiti Inte tial diffusing to loop D1 vacancy (3) Vacancy line (schematic) (4) Vacancy-loop (circular or hexagonal in-plane) Figure 4. Lattice defects in graphite introduced by radiation GRAPHITE TECHNOLOGY L,-cure I a Figure 5. The Graphite Crystal Lattice GRAPHITE TECHNOLOGY A C y I i.c~ctiirt I 1.10 1.10 .\n............................................\n.. ..................... ................ ............... 1.00 0.70 \"0.60 -e S 0.60 0 .3 0 -. ------------. ... ........ .. ... 0.20 0.10 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 rg (cm) Figure 6. Variation of Graphite Damage Flux with Distance Through Graphite GRAPHITE TECHNOLOGY * -..\nLecLtrc I 1.40 ---1.30 -. .\n..-..... .... .......... .\n............ . ...... ............... J.20 \"\u00b0\u00b01- -' 0.90 -S 0 .8 0 ......... .............. .............. .............. .............. .............. ............. .............. .............. .............. ............. .............. 0.70 ----- _ >\" 0.60 -_ 0 .50 - - - ......... ........ 0 .4 0 ...... ............. .............. ...... .... .- ' ....... ................ .............. .............. ............. .............. ............\ni .. .\n............ .............. ............. ............... .............. 0.40 0.30- ---- -----.. 0.20 -,_ 0.10 -------- ---- ........ .............. .\n.... ........-. .\n.\n.\n.\n.\n.\n.. .\n.\n-..... 0.00 -1 \"-0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 Dist.eot\u2022.o rough graphite, ins Figure 7. Relationship Between Graphite Relative Damage and Distance GRAPHITE TECHNOLOGY Lecture I VALAONCc $ W!TH AZ OIAE\"~ ~E~T I.c ------------- r -' scurALiEN7 7EAP CALZ*L7: .. 1 . SiNC \u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020\u2020 r:tQWALS.T TE \u00fdAP S S.~.lANC CNS-.A.\na *r 7 A- I V zC r V 7 C t '.':;Z G;: C ! =IC SPC::%AVeuNL-j 5-CI 1 I V6 'C- C \u00fdC! 1 i.S 'C-C SO ; S~ a- at CCasCCC ;45C i 3 2.75 *C.C(:s-CC6 *~~~ ..\n* * ~ I ..... ..... ............\nII:I4I 1 11 i s- !C .\nCCC \u00a3AZ-:'EUIAL-EN7 -, Figure 8. GYRAPHITE TECHNOLOGY :\u0002 S Icc GRAPHITE TECHNOLOGY COURSE LECTURE 2 1. Effect of Fast Neutron Irradiation The dimensional and physical property changes which occur in graphite due to the combined effects of fast neutron irradiation and radiolytic weight loss are very complex and do not always conform to a logical pattern. The magnitude of the changes is different for each graphite type and method of manufacture, some graphites exhibiting highly anisotropic behaviour whilst others are completely isotropic. Interpretation of the available data on any given graphite often presents difficulties, not only in terms of mean behaviour and scatter about the mean, but also in terms of extrapolating the available experimental data to reactor lifetime exposure values. Notwithstanding these difficulties, agreed interpretations and extrapolations of the available data on PGA and Gilsocarbon graphites do exist in the form of data manuals and reports, and these are continually updated in line with the latest thinking, observations and requirements in this technological field. For all practical applications it is therefore essential that the data user should make reference to the latest data manual or report when attempting to evaluate the dimensional or property change of any given graphite and to confirm from an appropriate specialist in the graphite field that the recommendation is current. The methodology employed in the evaluation of dimensional and property changes is also very complex and highly detailed. Great care is required when using the methodology to ensure that all known factors have been taken into account and correctly applied. Some of the changes are a function of several variables (such as dose, equivalent temperature, absolute temperature and weight loss) some of which are interdependent with one another. The irradiation conditions may also vary during reactor life, and this must be taken into consideration when the effects of in-pile exposure are evaluated. As with graphite data, the rules employed in graphite methodology also appear in data manuals and reports. and these cover both PGA and Gilsocarbon graphites where different rules apply. Here again it is essential that the user should seek the advice of a graphite specialist to ensure that the latest endorsed methodology is used for each application. The graphite database and methodology rules are indeed so extensive and detailed that it would be practically impossible within the available space of these lecture notes to summarise them in a manner that would permit them to be used in practical assessments. The following notes are intended as a general guide to the graphite data and the methodology used in PGA and Gilsocarbon component and structural assessments, with emphasis placed on summarising the graphite data and the most important rules employed in the methodology. A brief synopsis of the data and methodology used to evaluate each of the various changes induced in both PGA and Gilsocarbon graphites is given below.\nI 2. Dimensional Changes As was mentioned in Lecture 1, radiation damage by fast neutrons causes carbon atoms to be displaced into interstitial positions where they form interstitial loops which push the carbon atom layers apart, resulting in crystal growth in the c-axis direction. The displaced atoms leave behind vacancies which form vacancy lines, and these result in crystal shrinkage. in the a-axis direction. This behaviour continues over all measured doses. Between room temperature and 300\u00b0C there are large changes in crystal volume, this volume change being associated with stored energy. Above 300TC the graphite crystals change shape with little or no change in volume. During the manufacture of polycrystalline graphites, because of the large anisotropy of thermal expansion in the graphite crystal a and c directions, intercrystalline cracks are formed as the graphite cools after graphitisation. This leads to the formation of microporosity within-the graphite structure. These imperfections are usually referred to as Mrosowski cracks after the scientist who suggested that mechanism by which they are formed. This microporosity is the reason for the lower density of bulk graphite (1.70 to 1.80 g/cm3) when compared with the crystal density (2.23g/cm3). During irradiation, part of the c-axial crystal growth can be accommodated by the pore space available within the Mrosowski cracks. Since PGA is an extruded graphite manufactured from needle-shaped petroleum coke particles, the graphite tends to have the crystal basal planes aligned parallel to the extrusion axis. As a result, between 150\u00b0C and 300TC PGA behaves in a similar manner to the crystal, growing perpendicular to the extrusion direction and shrinking parallel to the extrusion direction. The changes in crystal dimensions can be very large as may be seen in Fig 2.1 which presents data on pyrolytic graphite where the carbon atom layers are highly orientated. As will be seen later, the dimensional changes of PGA graphite are significantly smaller because the crystal layers in nuclear graphite are more randomly orientated. Above 300'C PGA shrinks both parallel and perpendicular to extrusion. Gilsocarbon graphite on the other hand, being moulded and made from coke containing spherical structures in which the graphite crystallites tend to lie in an \"onion skin\" structure, has very isotropic properties. As a result, the dimensional changes in the parallel and perpendicular directions to the direction of moulding are the same. At all temperatures of practical interest, ie 300WC and above, Gilsocarbon graphite exhibits initial shrinkage during which stage any c-axis growth is accommodated by a-axis shrinkage and voidage available in the microporosity. This process continues until all the microporosity has been taken up and the graphite begins to swell, this point being known as \"turnaround\". Radiolytic oxidation increases the microporosity of graphite and is predicted to delay turnaround in Gilsocarbon graphite. The effect of radiolytic oxidation on the dimensional changes of PGA graphite is considered to be negligible. 2.1 Data and Methodology - PGA Dimensional Changes Figs 2.2 and 2.3 summarise the dimensional change data for PGA graphite obtained from ) high flux irradiation experiments in DIDO'1. Dimensional change, %, is plotted as a function of equivalent DIDO nickel dose (EDND), n/cm 2, for DIDO equivalent temperatures ranging from 150 to 6500C. Fig 2.2 shows that shrinkage occurs parallel to extrusion at all temperatures, and that shrinkage rate generally decreases with increase in temperature. Fig 2.3 shows that at temperatures below about 3000C graphite grows perpendicular to extrusion, the growth rate increasing with decrease in temperature. At temperatures between 300 and 650'C initial shrinkage in the perpendicular direction is followed later by growth. For PGA dimensional changes, DIDO equivalent temperatures are calculated using an activation energy of 1.2eV('). Radiolytic weight loss should be assumed to have no effect on dimensional changes. 2.2 Data and Methodology - Gilsocarbon Dimensional Changes The solid lines in Fig 2.411 are the recommended mean dimensional change curves for Gilsocarbon graphite in the absence of radiolytic oxidation as a function of dose (EDND) n/cm2 at DIDO equivalent temperatures of 350, 450 and 550\u00b0C. The dotted lines in Fig 2.4 present the corresponding recommended \"free\" shrinkage curves in the absence of shrinkage reversal as would be expected to occur in the presence of very high rates of radiolytic oxidation when creation of microporosity would delay the onset of turnaround. The dimensional changes are applicable to both parallel and perpendicular directions to the direction of moulding. It will be seen that the initial shrinkage rate increases with increase in temperature, and that in the absence of radiolytic oxidation the onset of turnaround occurs earlier the higher the irradiation temperature. In practice the dimensional change curves will be somewhere between the extreme values represented by the solid and dotted lines in Fig 2.4 since the radiolytic oxidation rate will be finite but small. In Gilsocarbon graphite, the rate of radiolytic oxidation is inhibited by the planned inclusion of methane and carbon monoxide as constituents in the coolant gas, thereby suppressing oxidation in the majority of the micropores. Since radiolytic oxidation is confined to what are referred to as the \"reactive pores\", a method of calculation has been devised whereby an intermediate curve may be evaluated between the solid and dotted lines by assuming that the intermediate curve is situated at a fractional distance between them represented by the ratio: initial reactive pore volume, VRPV, divided by initial total open pore volume, VTOT, in accordance with the equation: Dimensional change, % = (VRo) Curve A + 1(- Curve B (1) Curve A is the dotted line in Fig 2.4 Curve B is the solid line in Fig 2.4 VRPv is the initial reactive pore volume (at zero time) in which all of the radiolytic oxidation is assumed to occur, cm 3/cm 3. VTOT is the initial (unirradiated) total open pore volume cm3/cmI.\n3 The initial reactive pore volume is assumed to be directly proportional to the initial radiolytic oxidation rate, and from experimtental data it has been determined that Vpiv should be evaluated from the equation: IVRpV = 0.15 108 RR (2)' where: RR is the initial radiolytic oxidation rate at 6730K and 41 bar, g/g, hr (mW/g), and is coolant composition dependent as defined by the expression: RR = [ Ae-k(Cn 4 + 0.05 IIo-(3) where: (CH4) is the in-pore methane concentration, vpm A and k are constants taken from the following table: Table 2.1 and Fig 2.5 give values of RR for a convenient range of CO and CH4 values. For Gilsocarbon graphite, DIDO equivalent temperatures are calculated using an activation energy of 3.0eVI21. 3. Coefficient of Thermal Expansion The coefficient of thermal expansion (CTE) as used in the UK nuclear industry is usually measured in the range 20-120'C. 'he volume expansion of bulk graphite is much less than the volume of the crystal, largely due to the accommodation of crystal expansion by microporosity. The variation of the component crystal coefficients of thermal expansion ca and ca has been studied by irradiating highly oriented pyrolytic graphite. This shows that below -300'C ao decreases from 26 x 10.6 K\" to 14 x 10-6 K` while a. changes from -1.25 x 10\" K` to -+1 x 10.6 K\" over the same temperature range. Above -300\u00b0C both ca and cL, remain close to unirradiated values.\n4 ( The CTE of PGA graphite exhibits orthotropic behaviour with Q,.,2 perpendicular to extrusion significantly higher than that parallel to extrusion, the respective values being -2.8 x 10.6 and 0.9 x 10.' K\". Gilsocarbon graphite on the other hand exhibits isotropic behaviour with respect to the expansion coefficient, 2, being significantly higher at -4.3 x 106 K. Clearly both graphites have much lower CTE values than the ot, value of -26 x 10.6 K-1. measured in highly oriented pyrolytic graphite. The changes in CTE with irradiation of bulk graphite are a complex function of the changes which take place within the crystal and within the structure of the graphite which are not fully understood. PGA generally shows an increase in expansion coefficient with irradiation in both perpendicular and parallel directions to extrusion, the increase being more rapid and to a higher level the lower the irradiation temperature. Gilsocarbon graphite, on the other hand, exhibits a significant reduction in expansion coefficient with irradiation following a smaller increase at low doses. It is important to note that the thermal expansion coefficient of graphite is temperature dependent and exhibits a significant rise in expansion coefficient with increase in temperature. Values determined in the temperature range 20-120\u00b0C following fast neutron irradiation must therefore be adjusted to allow for this increase. It has been shown experimentally that the instantaneous coefficient of thermal expansion (CTE), a,, can be determined from the CTE value measured in the range 20-120\u00b0C, a( 10.12 0), using the relationship: ai = Ai.c'..oo120) + Bi (4) where Ai and Bi are dependent on temperature only, ie equation (4) is applicable to both unirradiated and irradiated graphite. The parameter Ai may be taken as unity for all temperatures of practical interest (in both Magnox and AGRs). The parameter Bi as a function of temperature is shown in Fig 2.6. The mean CTE over the temperature range 20 to T\u00b0C , .\nis of more practical use than the instantaneous coefficient, ai, and may be obtained from the relationship = (T-20)- T ai . dT (5) \u00a2z\u2022'7) (T-20) 20 S 1 foT (a od+Bi) aT (6) (T-20) f20 ie a(2 0 _ = ar, 0 __ 2 0 ) + Bi (7) where is the mean value of Bi over the temperature range 20-T1C.\n5 Figure 2.7 has been produced using equation 7 for discrete values of a 1 :\u2022 1 :f)j between 1.5 x 10` and 5.5 x 10\" 'C', and gives mean CTE values in the range 20 to T\"'C for T values up to 750\"C. 3.1 Data and Methodology - F'GA Coefficient of Thermal Expansion Figures 2.8a and 2.8b summarise the PGA experimental data\") on %0-120) changes as -a function of dose (EDND), n/cm2, for DIDO equivalent temperatures in the range 150 to 650'C. At the lower irradiation temperatures there is a rapid rise in expansion coefficient in both parallel and perpendicular directions to extrusion. As the irradiation temperature increases so the dose at which the thermal expansion begins to rise increases. Furthermore, the level to which it rises decreases with irradiation temperature. For PGA graphite, the irradiation-.induced changes in C)2.IM20C are evaluated from Figs 2.8a and 2.8b using DIDO equivalent temperatures calculated with an activation energy of 1.2eV\"'. The changes so evaluated are additive to the unirradiated a,:.o value for the graphite under consideration. However, irradiated thermal expansion values may be read directly from Figs 2.8a and 2.8b if unirradiated values are not significantly different from the starting values shown (0.8 x 10.6 and 3.0 x 1I06,,Ct). Radiolytic weight loss effects on CTE should be assumed to be insignificant. Adjustments for temperature dependence of CTE are as detailed in Section 3 above. 3.2 Data and Methodology -Cilsocarbon Coefficient of Thermal Expansion Figure 2.9a shows the recommended mean curves for calculating the changes in a%_,o-12o) of Gilsocarbon graphite as a function of dose (EDND) n/cm2 at DIDO equivalent temperatures of 350, 400, 450, 500 and 550'C. The initial rise in CTE is temperature dependent, the level to which it rises decreasing with increase in irradiation temperature. The subsequent fall in CTE with dose also occurs earlier with increase in irradiation temperature. As for PGA, the evaluated change in CTE is additive to the unirradiated c,_> 1 ,20 , value for the graphite under consideration, but irradiated values may be read directly from Fig 2.9a if unirradiated CTE is not significantly different from the shown starting value of 4.35 x 10\"6,0C\" 1. For AGR applications, a small additional correction to the irradiated a%.\n120) value is recommended to allow for a dependence on local strain. The sum of the elastic and creep strains should be evaluated and used to determine the appropriate correction from Fig 2.9b which is directly additive to the irradiated value. The correction leads to an increase in a( 2 o.t~o) for compressive strains and a decrease for tensile strains. The irradiation-induced changes in a% 2 0.12 0) for Gilsocarbon graphite are evaluated from Fig 2.9a using DIDO equivalent temperatures determined using an activation energy of 3.0eV. It has been found that radiolytic oxidation of Gilsocarbon does not produce significant changes in CTE.\n6 As for PGA, adjustments for temperature dependence of CTE are as detailed in Section 3. 4. Thermal Conductivity The principal thermal conductivities of the graphite crystal, that is K. parallel to the basal planes and K, perpendicular to them, are dominated by phonon lattice wave conduction. The presence of lattice defects introduced by irradiation damage reduces both principal\" conductivities, but for most polycrystalline graphites the conductivities are dominated by the component K, because K, > > K,. As will be seen later, further reductions in thermal conductivity may also occur as a result of pore opening due to radiolytic oxidation and pore generation at high doses due to crystal straining after turnaround. The data are traditionally presented in terms of the fractional change in thermal resistivity measured at room te~mperature. However, in practice it is the thermal resistivity at the operating temperature which is required, and so it is necessary to develop an appropriate relationship from which this may be evaluated. The thermal resistance of irradiated graphite is given by: 1 _ 1 + 1 (8) K(7) 7,o(7) K,(7) where 1 K(7) UnIrradiated thermal resistance at T *C 1 Ko(7) = Unirradiated thermal resistance at T\u00b0'C 1 -Thermal resistance at T'C introduced by fast neutron irradiation It is convenient to introduce thermal resistance at 30 0C and consider ratios: K,(30) _ K,(30) + K0(30) (9) K(7) K0(7) 14(7) K/(30) = Ko(30) + K,(30) K1(30) (10) K(Y) Ko(T) __(30) Ki(T) 7 The first term in equation 10 involves unirradiated values and depends on temperature only. The ratio K,(30)/K(T) has been shown to be dependent on temperature only. The ratio Ko(30)/Ki(30) involves measurements at 30\"C after irradiation at specified conditions and is shown to be dependent upon dose and DIDO equivalent temperature. Two additional factors need to be introduced into equation 10 to account for structural changes in the graphite following turnaround and to account for pore opening effects due to radiolytic weight loss. The final equation for the thermal resistance of irradiated and oxidised graphite at T*C then becomes: 1 [K, (30) K 0(o0O [ 0 7 + f.8 T] StK[..\nf SK K.(30) X,(30) _K,(30) K 5(30) K,(7) = Structural factor to allow for structural changes following turnaround.\n[{ =Factorial increase in thermal resistance due to radiolytic weight loss. Equation 11 is a general relationship which may be used to evaluate the thermal resistance of irradiated and oxidised graphite, 1/K(T), at any desired temperature up to the irradiation temperature; above this temperature changes in thermal resistance may occur due to annealing. 4.1 Data and Methodology - P(GA Thermal Conductivity The data from which the various parameters in equation 11 may be obtained in order to evaluate the thermal conductivity of irradiated and oxidised PGA graphite, K(T), are 8 1 K(Y) where: (11) summarised in Figs 2.10, 2.11, 2.12 and 2.13.\nFig 2.10 shows the fractional changes in thermal resistance, f, (= K,(30)/K(30)-I), as a function of neutron dose for different irradiation temperatures from 150 to 650\"C (DIDO equivalent). The changes are strongly temperature dependent with the largest increases. occurring at the lower irradiation temperatures. Fig 2.11 gives the temperature dependence of thermal conductivity (K,(T)/Ko(30) for unirradiated PGA graphite. Fig 2.12 gives the temperature dependence of thermal resistivity, 6T, for irradiated PGA graphite. The structure term Sk is unity for PGA graphite applications. Fig 2.13 shows the data on the change in thermal resistivity with radiolytic weight loss for PGA graphite from which use of the relationship [K]KI,,. = -3.1x is recommended, where x is the fractional weight loss. When Fig 2.10 is used to determine the fractional change in thermal resistance of PGA graphite, DIDO equivalent temperatures are calculated using an activation energy of 1.2eV. 4.2 Data and Methodoloav - Gilsocarbon Thermal Conductivity The parameters in equation 11 for evaluating the thermal conductivity of Gilsocarbon graphite in the irradiated and radiolytically oxidised condition may be obtained from Figs 2.14 to 2.19, inclusive.. Fig 2.14 shows curves of f as a function of dose for DIDO equivalent temperatures ranging from 325 to 550TC. It can be seen that f increases with dose, eventually reaching a saturation level which reduces with increase in irradiation temperature. Fig 2.15 gives the relationship between (K,(T)/Ko(30)) and temperature for Heysham II/Torness Gilsocarbon graphite. Fig 2.16 gives the same relationship for Gilsocarbon graphite in all other AGRs. Fig 2.17 shows the temperature dependence BT for all irradiated Gilsocarbon graphites. Fig 2.18 shows the structure term Sk as a function of dose for DIDO equivalent temperatures in the range 350 to 550\u00b0C. Fig 2.19 shows the data on the change in thermal resistivity with weight loss for Gilsocarbon graphite from which use of the relationship [K,,K],, = e2.7x is recommended, where x is the fractional weight loss. DIDO equivalent temperatures used to derive thermal resistivity changes should be calculated 9 using the following activation energy values: f should be determined from Fig 2.14 using 1.2eV. Sk should be determined from Fig 2.18 using 3.0eV. 5. Stored Energy The crystal lattice of virgin graphite possesses the lowest potential energy of the constituent atoms, and the presence of irradiation-induced defects increases the total energy. If the temperature of irradiated graphite is raised above the irradiation temperature, the lattice defects reduce their energy by mutual annihilation and the excess energy appears as heat. This phenomenon, known as stored energy, was first predicted in 1943 by the physicist E P Wigner and is sometimes referred to as Wigner energy. Stored energy may be characterised by several parameters of which the most commonly used are the total stored energy, E, and the rate of release of stored energy, dE/dT. As dE/dT is dimensionally equivalent to specific heat (J/g.K) its use permits a rapid assessment of the thermal stability of the graphite to be made. For fault study purposes it is assumed that the irradiation temperature must be exceeded by 50'C before significant energy release commences. At higher temperatures the release rate. dE/dT goes through a maximum (dE/dT)M, after which a reduction to lower levels may occur, although some high temperature measurements show shallow peaks. The magnitude of both E and (dE/d!T)m. are dependent upon the irradiation damage dose and the DIDO equivalent temperature. 'Their magnitude is directly related to the fractional change in resistivity. f. measured at room temperature. The following relationships were obtained from experiments on PGA graphite. It is important to note that they do not apply to graphite irradiated at the lower temperatures pertaining to the Windscale Piles. E = 31.1f (12) and ( (dEM E (13) d- , 1910 where 10 Ko(30) 1] f K(30) = Irradiation induced fractional change in thermal resistivity measured at room temperature. E = Total stored energy, J/g. (dE/dT),, = Maximum rate of release of stored energy, J/g\u00b0C. 5.1 Data and Methodology - PGA Stored Energy The fractional change in thermal resistivity, f, should be determined from Fig 2.10, with DIDO equivalent temperatures obtained using an activation energy of 1.2eV. No adjustments to either E or (dE/dT),,.. are required for structural changes or radiolytic weight loss. 5.2 Data and Methodology - Gilsocarbon Stored Energy Formal recommendations assume the use of PGA data. 6. Specific Heat The specific heat of graphite varies with temperature in a unique relationship which is the same for all well graphitised nuclear graphites. The relationship is shown in Fig 2.20. It should be assumed that the specific heat is not changed by fast neutron irradiation and is unaffected by graphite structural changes and radiolytic weight loss. 7. Young's Modulus The stress-strain curves for unirradiated graphite in both tension and compression are non linear, the material exhibiting hysteresis on unloading after which there is permanent set, Fig 2.21. The value of Young's modulus will therefore depend upon stress level and will be less than the dynamic modulus at significant strains. However, if the static modulus is measured at a low enough strain the static and dynamic values should be the same. Following irradiation the stress-strain curves become more linear and the permanent set is decreased, the strains to failure decrease, but the Young's moduli and strengths increase considerably. Experiments to determine the magnitude of irradiation-induced changes in Young's modulus have concentrated on dynamic modulus as the basis of measurement. In orthodox polycrystalline graphites irradiated at temperatures below about 300\u00b0C the dose 11 dependence of Young's modulus change is very complex, an initial large rise is followed by a fall, then a second substantial rise followed by another fall, eventually reaching zero as the graphite disintegrates. For irradiation temperatures above 3000C an initial rise is followed by a constant value for a period, then a substantial rise and again a decrease to zero. Increasing irradiation temperature diminishes the dose at which the changes occur, except for the initial rise which decreases in magnitude with increasing irradiation temperature. There are two distinct mechanisms involved in these changes, firstly due to changes in the elastic constants of the graphite crystals and secondly due to changes in the bulk structure of the graphite. Furthermore, there are two distinct mechanisms for changes in the elastic constants of the crystals. The first of these, which is responsible for the initial large rise in Young's modulus, is due to the pinning of mobile dislocations in the crystals, and this effect soon saturates. The second, which is responsible for the fall in modulus during the second irradiation phase, is caused by the changes in crystal modulus which result from the large changes in the c-axis lattice spacing which are particularly significant below 3000C. The effects of structural changes are also twofold. Differential straining of the crystal changes the pore structure in a way which at first \"tightens\" the graphite structure by reducing the average shear stress on the crystals (ie the modulus increases). However, at large crystal strains new porosity is generated which reduces the elastic moduli, eventually to zero, when the graphite disintegrates. Finally, it is necessary to take into account the reduction in Young's modulus which results from the increase in porosity due to radiolytic oxidation. This effect is dependent upon weight loss and graphite type. The combined effect of fast neutron irradiation and radiolytic oxidation is the product of the factorial changes due to each effect: E =( E) .k (14) where: E Factorial change in dynamic Young's modulus due to the combined E, -effect of fast neutron irradiation and radiolytic oxidation. (E, i oFactorial change in dynamic Young's modulus due to fast neutron =irradiation. S Factorial cha!ge in dynamic Young's modulus due to radiolytic oxidation. E12 7.1 Data and Methodology - PGA Young's Modulus The factorial change in Young's modulus due to irradiation above may be determined from Figs 2.22 and 2.23 which give the fractional change (E/Eo-l) as a function of equivalent' DIDO nickel dose for DIDO equivalent temperatures ranging from 150 to 6500C. The data are respectively for the parallel and perpendicular directions to extrusion, from which the various effects described in Section 7 above may be observed. DIDO equivalent temperatures should be determined using an activation energy of 1.2eV. Data covering Young's modulus changes due to radiolytic oxidation are shown in Fig 2.24, from which use of the following relationship for evaluating the factorial change as a function of weight loss is recommended for PGA graphite: E eJ = e (15) where x is the fractional weight loss. The combined effect of these two changes may be determined using equation 14. 7.2 Data and Methodology - Gilsocarbon Young's Modulus The higher irradiation doses and temperatures in AGRs give rise to a modified treatment of the factorial change due to fast neutron irradiation. This becomes two terms, firstly the pinning term which rapidly saturates at low dose and then shows no further change with dose, and secondly the structure term which first rises with dose and later starts to fall at the onset of turnaround. For AGRs, therefore, equation 149 is usually written as: where: PE. (16) E. P = Saturated pinning term to account for pinning of mobile dislocations. S = Structure term to allow for changes in the graphite structure due to irradiation damage. W = Weight loss term.\n13 je E E ( ( (\u00fd. (17) The pinning term is a function of DIDO equivalent temperature only, and may be determined from Fig 2.25 using an activation energy of 1.2eV. The structure term is a function of both dose and DIDO equivalent temperature, and may be obtained from Fig 2.26 using an activation energy of 3.0 eV. The weight loss term for Gilsocarbon graphite is also different from that applicable to PGA, and may be obtained from the relationship: (-3J4 (18) where x is the fractional weight loss. 8. Static Stren2th The strength criteria against which graphite failure under stress should be assessed depends upon the particular loading conditions to which the material is subject. Like all brittle materials, graphite has different failure apparent stresses depending on the type of loading and the component geometry. It is stronger in compression than bend, and stronger in bend than tension. For moderator bricks and keys, the stress distribution is a complex one to which neither the ultimate tensile strength nor bend strength may be used to calculate failure under tensile loading. In these cases the apparent stress at the point of fracture initiation must be determined by finite element modelling using the results of destructive tests on appropriately loaded moderator bricks and keys or slices cut from these components. The strength of a moderator brick or key which has suffered radiation damage and radiolytic oxidation may then be evaluated by adjusting this apparent failure stress by the appropriate factorial changes in graphite strength derived from experimental data on these effects. Fast neutron irradiation produces substantial increases in the strength of graphite except at very high doses where the fall in :he elastic moduli is accompanied by a loss of strength. The mechanism for the strength changes is the same as that causing changes to Young's modulus, being the result of within-crystal effects at low doses and structural changes at high doses. In a brittle material such as graphite a crack growth mechanism of failure would be expected to operate, and a Griffith-type relationship between strength a, modulus E, work of fracture g, and inherent flaw size c, viz: 14 C implies that changes within the crystal lattice without changes in the inherent crack population should lead to a strength change proportional to the square root of the modulus' change, ie: -- (20) where: Factorial change in strength due to fast neutron irradiation (EJ = Factorial change in Young's modulus due to fast neutron irradiation However, irradiation-induced structural changes external to the crystal may influence the work of fracture or the effective flaw size, as well as the average crack/pore spectrum sampled by the modulus, and so equation 20 cannot be expected to be obeyed exactly. Nevertheless, the relationship of Equation 20 has been shown to represent the data reasonably well up to the point of turnaround where the structural changes give rise to a fall in Young's modulus beyond which the fall in strength is more severe. Radiolytic oxidation also causes changes in strength, and the overall strength change is given by: (~~j (21) o IG 0ox where (Cy Factorial change in strength due to radiolytic oxidation.\n15 I2 E a- ag--(19) 8. 1 Data and Methodology - PGA Strength Determination of the factorial change in strength due to fast neutron irradiation should make \"use of equation 20, with factorial changes in Young's modulus evaluated by the method described in Section 7. 1. An activation energy of 1.2eV should again be used to determine DIDO equivalent temperatures in Figs 2.22 and 2.23 which give values of (E/Eo-1) as d function of dose. Determination of the factorial change in strength due to radiolytic oxidation should make use of the following relationship derived from the data of Fig 2.27: -~ (.....J (22) where x is the fractional weight loss. The combined effect of these two changes may then be determined using equation 21. 8.2 Data and Methodology - Gilsocarbon Strength As was the case for Young's modulus, the higher doses and temperatures in AGRs give rise to a modified treatment, and for such applications equation 21 is usually written: = = (P.S')Ir'W (23) where: P = Saturated pinning term to account for pinning of mobile dislocations. S\" Modified structure term to allow for structural changes due to irradiation damage. W = Weight loss term = ( ). _ ( .\n(24) C y ( m\u2022 0 0'.a' 16 The pinning term is the value of (E/E,,) read from Fig 2.25 using an activation energy of 1.2eV. The modified structure term is given in Fig 2.28, for which an activation energy of 3.0eV is appropriate for calculating values of EDT. These curves have been calculated assuming. that strength is proportional to the square root of the modulus change for doses up to the maximum modulus change. Above this dose, strength changes are assumed to vary linearly with modulus change. The modified structure term has thus been evaluated as: S =S Y < Ymx S* -y Ymx S. where S is the structure term shown in Fig 2.26. The weight loss term for Gilsocarbon strength changes may be obtained from the relationship: where x is the fractional weight loss. 9. Impact Strength UKAEA studies of unirradiated graphite specimens subjected to repeated impacts carried out on a variety of graphite types including PGA have shown that there is a one to one correlation between the dynamic stress at failure in a single impact and the static 3-point bend \"strength, Fig 2.29. From this work it was concluded that the impact energy to fail a rod was proportional to ob 2 -/E, Fig 2.30. Much less work has been done on the impact behaviour of irradiated graphite. Probably the most detailed work has been carried out on AGR fuel sleeves, but this work is very specific in being used to support safety cases for on-load refuelling. Although no firm rules exist for determining the impact strength of irradiated graphite, it is generally accepted that the above relationship between impact energy, strength and modulus may be used as a means of estimating the factorial change in impact strength due to fast neutron irradiation, (U/U,,),. Thus: 17 (C/c )2 (U/LI), -((26) Substituting the relationship between strength and modulus changes due to fast neutron irradiation given in equation 20 gives: (U/ 0)i -(E/E 0 o), This implies that the impact strength is unchanged by irradiation up to the point of turnaround (see Section 8). However this approach is simplistic since it assumes that the stress/strain behaviour is linear before and after irradiation. In reality, unirradiated graphite exhibits non-linear behaviour and significant hysteresis whereas irradiated graphite shows more linear behaviour and less hysteresis. These factors would be expected to affect the energy absorption capacity of the material but are difficult to quantify. It would be expected that radiolvtic oxidation would reduce the impact strength of graphite since both strength and modulus are reduced. UKAEA have studied the impact strength of various sleeve graphites, from which is was concluded that radiolytic oxidation up to 19% weight loss did not have any untoward effect on impact resistance over and above that indicated by the changes in static strength and Young's modulus. Thus the factorial change in impact strength due to radiolytic oxidation, (U/Uo)o,. may be obtained from the equation: (\u00b0IO\u00b02o (U/Uo)o, (EIE 0 9.1 Data and Methodology - PGA Impact Strength PGA graphite is not irradiated beyond the point of turnaround and so for PGA: -. (U/Uo). = 1 (28) ie there is no change in PGA impact strength due to irradiation alone. Equation 27 may be used to estimate the factorial change in impact strength due to radiolytic oxidation by substituting the relationships for static strength, and Young's modulus changes from equations 22 and 15 respectively. Thus for PGA graphite: -(e-L\u2022 -5e-.6 (29) Pu/,,)ox e -4.\" 18 9.2 Data and Methodology - Gilsocarbon Impact Strength Up to the point of turnaround Gilsocarbon static strength change due to fast neutron irradiation is directly proportional to the square root of the modulus change, as given by equation 20. For this dose range, therefore, there is no change in Gilsocarbon impact strength due to irradiation alone since equation 28 will apply. Beyond turnaround, Gilsocarbon strength is assumed to vary linearly with modulus change. Since Young's modulus is seen to fall beyond turnaround, this implies there is a corresponding fall in strength. From equation 26, it will be clear that a reduction in both strength and modulus beyond turnaround at the same rate will result in a fall in Gilsocarbon impact strength. This fall in both strength and modulus beyond turnaround is given by lE/,., where E, is the Young's modulus value beyond turnaround and E..., is the maximum value of modulus reached at the point of turnaround where the dose is , Thus for Gilsocarbon graphite: (U/U) = 1 y < MX (30) (L/UU) -(Et/E\u00fd)- = Et/Emji y a YrI (31) Et/ Em. where Et/E,,., is taken from Fig 2.26 using an activation energy of 3.0 eV. For Gilsocarbon graphite, the factorial change in impact strength due to radiolytic oxidation may be obtained from equation 27 by substituting the relationship given by equations 25 and 18. (UU\\)O -(e3\"7x\u00fd) -e-'. (32) e-3..u The combined effect of irradiation and radiolytic oxidation may be obtained by assuming that the two effects are multiplicative, ie that: 19 U/N- = (/Uo). (U/U 0) (33) 10. Irradiation Creep When graphite is subjected to an applied stress in the presence of fast neutron irradiation it exhibits creep strain. This behaviour is extremely important since without irradiation creep the stresses in moderator bricks and keys would become unacceptably high at relatively low doses. Thermal creep is negligible at all temperatures of practical interest. Experiments at constant stress showed that the creep strain during irradiation of graphite under stress is separated into primary and secondary contributions: C, EP + E (34) where E, total creep strain -= primary creep strain (recoverable) E, secondary creep strain These experiments have shown that when structural changes within the graphite are not significant, the total creep strain under a constant externally applied stress a is given by: 1E -(1 - e-4)) + a . 0.23y (35) where: 7 = Equivalent DIDO nickel dose in units of 102' n/cm2 , = Initial (unirradiated) Young's modulus obtained in a static test up to stress a. Thus both primary and secondary creep terms are linear with applied stress and inversely proportional to the initial Young's modulus. This applies for different graphites, for different directions within the same graphite, or for density variations in the same material. The creep strain may be uniquely expressed in elastic strain units (esu); 20 E (esu) = (1 - e-4y)_ 0.23y \"where: esu 3E 0 Data covering several types of graphite are shown in Fig 2.31. The upper two plots demonstrate the linearity of creep strain with dose and the fit with equation 35. The lower plots demonstrate that despite major differences in properties, all graphites conform to the single relationship of equation 36 when creep strain is expressed in elastic strain units. The experimental data indicate that creep is substantially independent of temperature in the range 300 to 600'C. Below 300'C, the data are not sufficiently well established to doses at which structural effects may influence creep rate. Under constant stress, the secondary creep term in equation 35 is modified to allow for structural changes in Young's modulus at high doses and for radiolytic weight loss by substituting for ,Eo the effective modulus: = 1 Eo. S.W (37) where S W (E LE) E,uc,.jO Hence, the secondary creep strain between dose limits -,, and -y may be written 0.23c f 2 1 dy Which may be further modified to allow for variable stress: (38) 21 (36) 0.23 fY2 -d (39) ,E, Y, S. W The concept of an effective modulus should also be applied in the consideration of primary creep and its recovery. 10.1 Data and Methodology - p_.A Irradiation Creep The irradiation creep data at temperatures > 300\u00b0C are reasonably well defined and indicate the use of a structure term, S, of 1.0. The complex modulus changes < 300'C preclude definition of a structure factor in this temperature range. The weight loss term, W, is represented by the relationship given in Section 7. 1, equation 15, ie: W = (EJO = -.\n8 E.ox 10.2 Data and Methodology - Gilsocarbon Irradiation Creep The structure term, S, recommended for Gilsocarbon may be obtained from Fig 2.26 using an activation energy of 3.0eV. The weight loss term, W, is given by equation 18 in Section 7.2, ie: -3.4 11.0 Activation Energy Values Table 2.2 gives a convenient summary of activation energy values recommended in the Data and Methodology sections of this lecture for use in calculating DIDO equivalent temperatures when determining the various property changes of PGA and Gilsocarbon graphites. References I. Birch M and Brocklehurst J E. A review of the effects of irradiation on the physical and mechanical properties of Pile Grade \"A\" graphite. ND-R-1236(S), August 1986. 2. Brocklehurst J E. Irradiation damage in CAGR moderator graphite. ND-R-1117(S), July 1984 and Addendums.\n22 Lecture 2 Attack Rate Parameter for RPV Model in/gim per hr per mW/gm x 10*'8 I Carbon Monoxide Methane 2500 5000 -75000 10000 I 15000 20000 25000 0.0 I 0.822 0.641 0.505 0.411 0.312 0.270 0.247 25.0 I 0.656 0.548 , 0.4-4,2 0.366 0.286 0.253 0.235 50.0 0.526 0.470 0.388 0.327 0.263 I 0.237 0.224 75.0 I 0.423 0.404 0.341 0.293 0.24.1 0.222 0.214 100.0 0.3a3 0.348 7 0.301 0.262 0.222 0.209 0.204 125.0 0.280 0.301 0.267 0.236 0.205 0.196 0.19.1 150.0 i 0.20 .262 I 0.237 0.213 0.190 0.185 1 0.186 175.0 0.192 0.228 0.211 0.193 0.176 0.174 1 0.178 200.0 0.161 0.200 0.189 j 0.175 0.16- 1 0.164 0.170 225.0 0.137 0.177 0.170 0.160 I 0.152 0.155 0.163 250.0 0.118 0.157 0.153 0.146 1 0.142 0.147 0.156 275.0 0.104 I 0.140 0.139 0.13-l 0.133 0.140 0.150 300.0 I 0.092 0.126 0.127 0.124 1 0.125 ; 0.132 0.144 325.0 0.083 0.114 0.116 0.114 .\n0.117 0.126 0.138 350.0 0.076 I 0.104 0.107 0.106I 0.111 I 0.120 0.133 Table 2.1 - Attack Rate Parameter as a Function of Methane and Carbon Monoxide Concentration for Gilsocarbon Graphite at 400 degrees C and 41 bar ........ ...... ...... ...... .G R.P IT E .E ........ GRAPHITE TECHNOLOGY ..\njN Lecture 2 Table 2.2 - Activation Energy Values for Calculating DIDO Equivalent Temperatures when Determining the Irradiation - Induced Property Changes of PGA and Gilsocarbon Graphites Property Change Dimensional Changes Coefficient of Thermal Expansion Thermal Conductivity Stored Energy YM versus dose Young's Modulus YM Saturated Pinning Term YM Structure Term YM versus dose Static Strength YM Saturated Pinning Term YM Modified Structure Term Impact Strength, Et/Emax Irradiation Creep, YM Structure Term Activation PGA PGA 1.2 1.2 1.2 1.2 1.2 1.2 Energy, eV Gilsocarbon 3.0 3.0 3.0 1.2 3.0 1.2 3.0 3.0 3.0 ..\nt I, GRAPHITE TECHNOLOGY Lecture 2 N N x ---I Figure 1.O~esincL changS 0 ,-'r: ric 5rczhit iow lmerature GRAPHITI\"E TCHNOL OGY Q:l: Lecture 2 rAAS7 NEUTRCMJ 005:E n,:r,- iEn-Ni -I KIYr Iq;\u00fd'!A7iCN SymSCL 7EMMATRu x 250 '4 A-Figure :2.\nIRRAMMO~N INDUCED O[MENSiGNAL C:-,ANCES PARALLEL 70 EXT,1RUsCN STANOARO 0100 fMk E HOLLOW FUEL ELEMEN0ATA' GRAP'HITE TECHNOLOGY 2 3 4.\n1 -.\n1 !so ADO rioNHDj 31 31H avas 1) NBV5B 3S MflO4 2~ DOD 7VONI& NDI%1--X2 02. &ffJINddSREDNVIH- '1VNO!SN\u00fd3WNK Dn3JfN1I N0Ol V10-V IObI 0-CIS z 61Z DC 3 n Ifl Sb --d; NDLLV lll'\u00fd t I oE JnfiA C 'A K .7 A 0 x -V .7 t ),-:; Z ii -(Nr\u00fdm) t -10 G;4 ;Is Lecure 2 1.00 0.00 -1.00 -2.00 -3.00 -4.00 -5.00 -6.00 -7.00 -8.00 -9.00 -1 0.00 2.00 I Fast Neutron Dose, n/cm\"2 (EDN) 0.50 1.00 1.50 Figure 4. Dimensional Changes and a 'free' Shrinkage of CAGR Moderator Graphite at Different DIDO Equivalent Temperatures GRAPHITE TECHNOLOGY '0\u0002 0' L Constant State from 1.5x10-22 / 450 35 S\"'. \"'450 \u00b0: \".550 0.00 J .,cclure 2 [CO] vpm --x 2500 4 5000 -e- 75000 -8- 10000 -15000 -- 20000 - 25000 Figure 5. Attack Rate Parameter as a Function of Methane and Carbon Monoxide Concentration for Gilsocarbon Graphite at 400 degrees C and 41 bar .. ......... .\n........\n\u2022..\u2022.......... .:.:.:.::.:.... ::.. .. :.:.:. ...... .\n... .\n.\n.\n.\n........ GRAPHITE TECHNOLOGY Lcltere 2 1.80 1.60 1.40 1.20 1.00 '0 W 0.80 0 0.60 0.40 0.20 0.00 -0.20 -0.40 i I , I I I I I 0 100 200 300 400 500 600 700 Temperature (QC) Figure 6. Parameter Bi v Temperature for Graphite .......... ... .. .. ...... .... .. .... :.: .. : .: :. :.:.::.:::::::: :::: :::H I:::: T E C H N O L O GY.:::::.:...... . ..\n:..,, ':::::::::! GRAPHITE TECHNOLOGY '-in 8.00 7.50- I X( 2 0 _1 2 0 ) X 1.0 E6 7.00 6.50 6.. 5.50 S5.50 -5.00 S5.00 -4.50 o4.50 -4.00 S4.00 3.5- 3.00 2,.50 2.00- 1.50 1.50 1.00- I I 0 100 200 300 400 500 600 700 800 Temperature (QC) Figure 7. Mean CTE v Temperature & a(2 (0_ 120) for Gilsocarbon Graphite Lccture 2 ........ ..... .\n.\n...... ...... ........... 0 ......... .\n.... ..... .. ... .. .\n...... \u00fd.,x . ... .\n. ....................... ..... ............ . . ............. .......... .. .... ..... ... .. ..................... . .... .\n.\n.\n........ .\n.... ..... .... ................ ............................... ...... ... .. . ........... .\n.... .. ... .. .. ....... ........... ... ......... .................. ........ ............ ......... ................................. ........... ..... ............ ................... ............... . ......... . ....... . .\n.... .\n............ ........... ............ ........... .. ....... .\n........................ ............... ........... GRAPHITE TECHNOLOGY Figure 8(a) & (b).\nKy S \", P _IA \u00fd7iON x 0 x'0o 3 1200 zoo 220 300 350 Lso 650 U N I PA ITE 0 VALU ES' 2 pAL L --- CEX RL US CN I 2 A 0 F AST NEUTPON COSE, n/Cim- ;-,-N) TH. PAL EXPNSO0N C0E=,i CIE.NT(20 -0C) CHANGES WITH ES7 NEUTRZON DOSE. 43TANOARO 0100 Mk~ El HLLOW FUEL -ELEIMEM7Z;ATA, . . ... . . . . . . . . .\nGRAPHITE TECHNOLOGY Lecture 2 L.ecslurt 2 6.0 EDT 5.5 -(3eV) 350 54.0 40 4 3.5 , 30 .0- 0 w C) 5 3.5 -..\n: . el 3..... ~2.5 2.0 1.5 1.0 -T r T T I 0 20 40 60 80 100 120 140 160 180 Dose (n/cm 2 x 10 20) Figure 9(a). a( 20 _120) v Dose & EDT for Gilsocarbon Graphite GRAPHITE TECHNOLOGY o\u2022 i Lecture 2 Figure 9(b). Change in a((20 _120) v Total Strain for Gilsocarbon Graphite \u2022~~ ~~~~~~~~~~ .. .. ................... .. .. .. .. .. .. . .. . . .. . .... ..... .......... .. .. .. ........ ....\n:::. .-......\n::... ... .: ...: ...::..::;..; ; : : ::.. GRAPHITE TECHNOLOGY Lecture 2 SY\u00fd1CL l.QA0(ATICNk 7;- PEPR.JURE, C x ISO zoo 0 A 2OO'C 225 250 300 3550 650 x x 2502-1 A A A NO00 .3SO *C FAS7NEUTRCN OOSE, nic~mi (EON1 I 0 Figure 10. FqAC7ICNAL C'HANGES IN 7lijECMAL \"-\u00fdE'iSTV17Y Wi7H'FAIS7 N-U7hON OOSE: S7ANDARO 0100 MklIII hJOLLOW FUE ELEMENTW0 A 7. GRAPHITE TECHNOLOGY IG0 7.-150% 20 :. -^ !Q Z-.\n7.\nLIec ttre 2 -J 200 1.00 MEASUREMENT TEMPEPATURE, OC 600 800 Figure I1. THERMAL CONDUCTIVITY TEMPERATURE DEPENDENCE OF UNIRRADIATED PGA GRAPHITE ............\nI.. ... GRA1PInIfrE TEciNOLOGY 1.0 rn a I a 05 -0 I !\n/112 Lecture 2 Ii-I ,.1!tI I 900 0- 100 200 300 1600 500 600 700 800 900 TEMHPMIFAT1JPF, K( Figure 12. THIERMAI- RESISTANC\u2022E IN IRRADIATED FiRAPIIITE NORMALISED] TO 300 K( .\n.......... ......................\n*\" \u2022 . . :\u2022 ., ;\u2022 .\n.::: : :\u2022 \u2022\u2022: ::;:::::: .:: -',.. .. ........ .... ....... I.............. ............ ........... ............ .(,I mI i irlE mlciI NOILOG V \"i Lo \u00a3 Lecture 2 3 0 2 1 0 o 0 7/ 7 \"I I/ / 0 0 0 0 1 0 .2 0 .3 FRACTIONAL WEIGH1T LOSS, x Figure 13. CHANGE IN THERMAL RESISTIVITY WITH RADIOLYTIC WEIGHT LOSS .\nT E .I.NO...O G GIZA IIII TrErc TECNOLOG Y :....\n.Lecdure 2 EDT (1.2eV) 11.00 10.00 9.00\" 8.00. 7.00 6.00 5.00 4.00 3.00 2.00 1.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Dose (n/cm2 x 10 20) 70.00 80.00 90.00 100.00 Figure 14. f-Factor v Dose & EDT for Gilsocarbon Graphite s-I 0 U 0 U 4 tI-I .\n.. .?:\u0001..-'. :,:::.:::. .-.\n: :.:: :.:.:. ....\n::.:. .$.!: .:.:.' $ ..... . :: ..\n:4. :.: ...\n'$ .. .: x ?::: GRAPHITE TECHNOLOGY Leciu rc 2 1.00 . 0.95 0.90 0.85 \" 0.80 S0.75 0.70 0.65 0.60 0.55 0.50 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 Temperature ('C) Figure 15. Ko(T) / Ko(30) v Temperature, for Gilsocarbon Graphite from Heysham II / Torness Only .. .~........... .\n......\n~-..............'................... GRAPHITE TECHNOLOGY Lecture 2 1.00o 0.95 0.90 0.85 o 0.80 0.75 0.70 0.65 0.60 0.55 0.50 -I i I I I I I I 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 Temperature ('C) Figure 16. Ko(T) / Ko(30) v Temperature, for Gilsocarbon Graphite for AGR's other than Heysham II / Torness .......... .. .\n. .. ...... .. ...... .. ..... .. ...... ......... .\n... ... .... ........ ........... ..... ........ ..... ..\n.\n.... .... ....... ..... ... .. .\n.\n..\n.\n. ..... ... ... ...... .. .... .... .\ng ...... ..... .. ..... . ....... ... .\n................ ....... ....... ........ ............ ...... ..... . .. ............... ...... ...... .... .. ... ......... .. .. .. ..... .\n............................. ........... ................. .. .. .... .\n................ ....................... ....................... . .............. .... .. ....................... ....... .... .\n.... .................... ... ............. ...... .\n.\n...... ... .. .. .. .\n.\n. .. ... ......................... GRAPHITE TECHNOLOGY 1.20 1.15 1.10 1.05 1.00 0.95 0.90 S0.85 0.80 0.75 0.70 0.65 -U.oU '~ I I I I I I I I II III 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 Temperature (QC) Figure 17. Delta (T) v Temperature for Gilsocarbon Graphite .. TEC.NO .. GRAPHITE TECHNOLOGY Lecture 2 I /(mN ,I; ILechire 2 80 100 120 140 160 180 200 Dose (n/cm 2 x 10 20) Figure 18. Sk Factor v Dose & EDT for Gilsocarbon Graphite 2.0 1.9 1.8 1.7 1.6 1.5-1.4 1.3-a-. 0 C) w EDT 550 , 500 450 400 350 1.2 -1.1 1.0 0.9-0.8 \u0002.1 0 0 20 0 40 6 60 ... .~ .*.*.. .\n.**.~. .* .*.* . .~ .*.*. .*..~. ..~ ~. *.* ~. .. .\n. ...\n~.... ...........\n*.*.*.....* .*.. ... ... .\n.\n.\n.\n.\n.\n.\n. GRAPHITE TECHNOLOGY (\u0002N Lecture 2 o 0.2%/. 101(\u00fd0? & 2%Y CO/OOV'/ fill, 3 ~-2 10 20 30 Iet W I (idIII I.OSS,0!u Figure 19. FRACTIONAL INCREASE IN i[P'I- 17RAL 1RESISTIVITY OF (iR*APIIlTE CAU SEID B3Y RADIOLY'ric OXIHIAT ION GRAT'liI TErCHINOLOGY I 2.20 . 2.10 2.00 1.90 1.80 1.70 1.60 1.50 1.40 1.30 1.20 1.10 1.00 0.90 n An I I I I I I I 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Temperature (OC) I 1 1 1400 1500 1600 Figure 20. Specific Heat v Temperature for Gilsocarbon Graphite \u2022:: := ::\u2022:.\u2022.\u2022\u2022 \u2022::\u2022::::::\u2022.\u2022:;\u2022.:...::\u2022:::.::::::\u2022::::: \u2022 \u2022.q\u2022\u2022.q:\u2022\u2022 \u2022\u20225.\u2022\u2022Y.:\u2022:\u2022:\u2022:.\u2022. \u2022\u2022.Y.\u2022: \u2022 \u2022:\u2022 \u2022\u2022:\u2022 :\u2022:\u2022+..\u2022\u2022.\u2022::.:\u2022::\u2022::........... .\n.\n...... ...........- ........ : .. . S. .. .. . .. .. .. ... .. ... .. .......\u2022 ..... w \u2022--...-.....-\u2022.. -.\n\u2022 w... ..-. \u2022 . .. .. .. . ..\u2022 .. . .... .\u2022..... .\u2022.\u2022\u2022. ..w...-..-. w\u00a2-w~ \u2022\u2022w..,..v \u2022.v\u2022.w \u2022...\u2022.\u2022.~ w .\u2022...+: ::+ +: :+ :+ ,: \u2022 ++: :.x :+ +: :x :+:+::::::::::.........::::: ......... ..... ..................... ...\n: :::::::::::::::::::::::::::::: :: :::::::::::::::::::::::::::::::: .... IT T EC ........... ...............\n\" Lecture 2 9.\nI I 1400 1500 1600 V \u2022l v*,,/ Lecture 2 i000 _ Loading J-Untcading //// 700 -/ 0 (Z) Tension I 0.02S 0.05 0.075 -rrin, r/ -Loading 3000 Un~cadinc 2000 -/ // 0 0.1 0.2 0.3 0. 0.5 (b) Comcressior _train, % Figure 21. Stress-Strain Diagrams for Graphite 4-.' LI (A M C_ -0.10 0. 05 -0.1100 -0.2 - 0., GRAPHITE TECHNOLOGY f-rn> KW I I...\nL.ecture 2 S Y III10t. IOO1AI)IAriont IF ~f Il A~TURE x A v. v A 2000C ISO 200 250 300 350 1, 50 650 I '.4'..' A *A I * v tVVV xt A A A 0)0 10 Is 20 25 30 3 5XlO 2 1 rASr tiI IJ!Jj0N14 )()%1 j/tjj1 2 I[ID)N Figure 22. FRA(TIONAI. (IIANi[S IN YOUNWS MODJULUS WITHi FAST NEUTRfON DJOSE: Sf ANDAPRJ 0100 Ilk III HOLLO1.W WEIJ[ ri EJIENI DATA. PAPALI Ell TO EXTRUSION ........ .\n.... .\n.\n.\n.\n.\n.\n.\n.\n.\n15 KEY SYBLIRRADIATION SYM3O. MPERAVUflE a x x X ,x 0 I A A A Isu 200 2S0 1. 50 (IS0 8' 2SA A f.50cC L650%C 0 A A A A A 10 is .0 FAW MrA IT HON I)O\"A, 111(F11WEN 25S 20 115 Figure 23. FRACTIONAL. MIAKES IN YOUNOiS FHOI'.tJI.US WITHI FASIr NUi~tPON DOSE: STANDA[RD 1)100 MkId!llHOLLOW FUEL ELEFMENTr IATA. 1PERPENI)ICU[AP [0 EXTRUSION ......... .\n...... ... .... .... ........... ............... .... .. ...... ....... ..\n.... ........ .\n. .......... .........\nL~ecture 2 3 OX 1021 Ox '43 10 A a 05 05 \u00fd00 c0 ........... ... ...................... SSO- 1019m]\u2022M 11,]OIUVU IIm SfifIlION SM.iNflOk NI JUNVII] \"17Z OIP!LI to 0[ -110 (- 9 #3I -to 1-0 I I -I o() 0) %.0 > 61 VI VV 0(110 M 60 Ll F0 (0 z 0 1 0 0 X SS() I I l(.13M 1VHOI I iVilJ \u00021II1p\u0002Yj Lclure 2 Figure 25. Saturated Pinning Term v EDT for Gilsocarbon Graphite GRAPHITE TECHNOLOGY I.ccture 2 2.50 2.404 2.30 400 2.20 2.10 ......... 2.00 -350 1.90 1.80 -55O S1.70 S1.60 1.50 1.40 1.30 1.20 1.10 -1.00 I 0 20 40 60 80 100 120 140 160 180 200 Dose (n/cm 2 x 10 20) Figure 26. Structure Term v Dose & EDT for Gilsocarbon Graphite ... .................. .. ........... .\n.\n. .. .......... ............ .\n........... .......... M ..... X . .\n...... .\n............\n.............\n...... ............................ .. . ..... ............ ............ ... . .... ......... ............\n.\n. ............\n... ....................... ............. ... ......... .\n. .\n. .................... . ......... .......... .................... ........................... ......... ... .......................... ........ ......... .. .............\n. GRAPHITE TECHNOLOGY o: ::: 0 **o 0 0 0 0 \u2022:!ii~is.J![ .\n. !i~iiliii! \" -- \"4 .I \u2022\u2022. \" \"-::..i\u2022ii- A g\u202277iCi' iii4ii ' \"J / !!!!!!! .a , \u2022 -4:<::::0 ::::::-./ ii~iiiii \u2022 \u00a2 _ \u2022/ \", .,,,.., ,. , !\u2022!ii7!iii 4 /4 iiiii;i:7 \u2022 \u2022 -,,o/ i!!!ii~iii \\ 4 /\u00a2 o ,,.,. +. / ,i~!: : -*J, ri!:!!!: =- = iii~iio,/ / / * -]~i: !- -;:::<::': iiii~i~i8 Sii'\u2022'\u2022\u2022= ::::;:::: ;:\u2022::: /: ii7;- > -, \"i:Y-: -/': : ,\u2022 \":: \u2022 ILctlurc 2 Figure 28. Modified Structure Term v Dose & EDT for Gilsocarbon Graphite .\n. ..... ...... .. .. .......... ........... .. ...... ........ ....................... . ... .... ............ .... ..... .\n.\n.\n.... .... ........ .\n.\n.\n.\n..... .\n.\n................. ..... ....... .\n........ .\n... . .............. . .......... .\n..\n................. ..... ........... ............................\n.... ..... . ..... .... ... ........ ................ .\n............... ....... ......... ... .\n.\n....... ................. ..... .. .... ... .\n.\n.... .... .. .... ..... ........ ........... :.l. ........ \" 11 ..... .:.. . .......... .\n....... ......... ............. ................ .\n... .............. ' .. .: . ......... GRAPHITE TECHNOLOGY Lcclucre 2 Fig 29. Stress required to fail In a single Impact test versus the 3-point bend strength for different graphites \u2022 X PGA 100.00 , 90.00 (_ +-- EY9106 80.00 Co 70.00 E 0 ATJ 60.00 S50.00 -f 13AEL 40.00 30.00 .\n.\n.\n............... -7V CG W 20.00 --x A Gilsocarbon U) 10.00 0.00 -, , , , , , ' ' , , , , , ,' H D G 0.00 20.00 40.00 60.00 80.00 100.00 STATIC 3-Point Bend Strength zx POCO SIII...... ............. .\n.\n:: .\nY (GR A Ih I'I'IE 'TE(1N(h I,\")( Y I ,t.v'ttrtc 2 Fig 30. Energy required for failure of rod specimens after 50 Impacts versus energy to fall 3-point bend specimen. IX PGA -5 0.70 -- EY9106 \"1) C 0.60 C_ 0 BAEL Sleeve L. 0.50 -IIATJ a 0.40 .-0.30 V GlIsocafbon o\u2022 0 .20 ........ .. ... , . . ... . . .\n... _ _ _ _ A CGW 0.10 0) X w 0 .0 0 ------ -r -- r --1 -' - F -r 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 3-point Bend Strength z POCO C.... .......\n(. CAP1 I ITE lE C I I N 1, G(N' AOOIONHDai alIHc[VIID c 0. x E-t-IlLi'drdn- P VDC c !1-2r. *2SOO ND!:.Lri3N ISTz' -- C E-0 Z-C. I-C -'D! x '--c, XVV x xff I s o L, N, o e -, r, 3 N I S T E I/ Noe 53-,1w.ert\u00fd-, LIT Dd c zL:;:/u ILZOI.X9 s z L z -3jm-.)-3-j r, D,-: DN)Ori-,DNI S3.LlHdVL4D IN3113-'\u00fd 10 -JD NOI.LIVW d 6\u00fd- K D U ONV 6336) ;\u00fd FN-0511-v 1 3\u00fd\u00fd GRAPHITE TECHNOLOGY COURSE LECTURE 3 1. Thermal Oxidation in Air Thermal oxidation of graphite in air is an important safety consideration in any situation where hot graphite may become exposed to air since the oxidation process is exothermic. A self-sustaining reaction, or even worse a runaway reaction, may ensue - particularly if stored energy is released as was the case in the Windscale Pile No 1 fire in 1957. Situations of interest in current safety studies include depressurisation by duct fracture of steel-pressure vessel Magnox reactors, and waste disposal of irradiated graphite components by burial or other means. Three possible reactidns may occur in the presence of oxygen, and these may be represented by the following equations: C + 0 2 = CO2 C + 1/2O. = CO CO- \" -CO 2 All of the above reactions are exothermic, the first has the heat of reaction of 94k.cals/mole and the other two combine to give the same total end product and heat generation. With an excess of oxygen present mainly CO2 will be produced, but with oxygen depletion such as may occur at high temperatures in large bricks due to diffusional control, the reaction may be limited to formation of CO. 1.1 Virgin Graphite Oxidation In the presence of air, graphite temperatures have to be at least 3000C before any appreciable reaction will take place. The rate of oxidation increases with temperature, and numerous studies have shown that air reactivity is related empirically to temperature by the Arrhenius law: AR = Aer' (1) where: AR = Rate of graphite oxidation in air( air reactivity) usually expressed as fractional mass of original graphite consumed per unit time, ttg/g.hr.\nI A = Pre-exponential constant, which is dependent upon both graphite type and the state of the graphite (ie whether virgin, irradiated, oxidised, contaminated, etc). E = Activation ertergy, cal/mole, which is also dependent upon the state of the graphite. R = Universal gas constant (= 1.986 cal/mole *C). T = Temperature, *K The standard measurement temperature for air reactivity is 400\"C, and the value of air reactivity at 400\"C for virgin graphite is largely dependent upon the purity of the material. For the commercial Magnox stations the mean virgin reactivity at 400*C is taken to be 3.2 ;Lg/g.hr, but graphite variability is recognised by assuming a \"rogue brick\" reactivity of 10 ;Lg/g.hr in fault studies using the RHASD computer code. It is important to recognise, however, that the graphite for each application will be different and dependent upon the method and raw materials of manufacture, and these may affect both the mean and spread values of air reactivity. 1.2 Activation Energy Pure carbon has an activation energy of 60,000 cal/mole (usually expressed as 60 k.cal/mole), but quite low levels of contamination above about a few tens of ppm will reduce the activation energy to about 40 k.cal/mole. Very severe contamination may reduce the reaction rate to lower levels of the order of 30 k.cal/mole, but current studies generally assume a standard value of 40 k.cal/mole. However, whenever data are available on measured values of activation energy for a given application these should always be used. Figure 3.1 shows the effect on air reactivity rates at temperatures above 400\"C of assuming different activation energy values.. Graphite is a porous material with a high surface area within the pore structure-typically 0.1 to 0.2 m2/g. This can vary from one graphite to another leading to large variations in the r pre-exponential constant A, and hence reactivity, while the activation energy, E, remains substantially constarrt. However, it should be noted that the nature of this in-pore oxidation leads to a breakdown in the Arrhenius law relationship (Equation 1) at high temperatures due to oxygen starvation within pores that are remote from external graphite surfaces. The reason for this is that in the majority of cases oxygen can only gain access to the interior pores by the process of diffusion. Furthermore, the by-products of oxidation (CO and C0 2) can only escape by diffusion. The diffusivity of gases within graphite is very much lower than in free space, and it is as a result of this low diffusivity that diffusional control can reduce the rate at which air reactivity increases with temperature. The transition to this stage depends upon a number of factors including component geometry, air reactivity and diffusivity, and would also be influenced by the existence of any pressure drops which induce permeable flow.\n1.3 Effect of In-Pile Exposure In-pile exposure causes the air reactivity of graphite to increase, and there are three possible ways in which this may occur: i. By crystal damage due to fast neutron irradiation which creates crystal imperfections thus making more reaction sites available at which oxidation can occur. ii. Due to radiolytic oxidation creating more porosity and hence surface area at which thermal oxidation in air may proceed. iii. By contamination of the graphite during operation by materials which may either increase or inhibit the oxidation rate. In this particular section we shall concentrate on the first two of these effects. Whenever possible monitoring data should be used to establish relationships which define the increase in air reactivity of virgin graphite with fast neutron dose and radiolytic weight loss. Such data are available from Nuclear Electric's Commercial Magnox reactors from which the following equation has been derived for use in RHASD to calculate graphite heating under fault conditions of loss of coolant and air ingress: Ri = Rv[ 1 + 0.0035 D.Re} + 12.5 {1-exp(-3)} (2) 3 where: R, = Irradi.ated air reactivity at 400'C in /g/g.hr Rv Virgin air reactivity at 4000C with a value of 3.2 tg/g.hr for regular use and a value of 10 -g/g.hr for the RHASD \"rogue brick\". D.R(0) -Calder effective dose in MWD/Te(u) 7 -Graphite percentage weight loss due to radiolytic oxidation. Data at other temperatures are then derived using an activation energy, E, of 40 kcal/mole since the combined effects of irradiation damage and radiolytic oxidation are assumed to have no effect on activation energy. The use of Calder effective dose, D.R(0), rather than Calder equivalent dose, D, is to take account of the fact that fast neutron irradiation at higher temperature produces less damage in graphite than is produced at lower temperatures due to the higher rate of annealing which takes place at high temperatures. The term Calder effective dose was coined in the early 1960's when it was believed that it correlated well with the measured increases in stored 3 energy and thermal resistivity. The relationship between R(O) and Calder equivalent temperature is shown in Fig 3.2 from which it will be seen that R(O) falls with increase in temperature up to a Calder equivalent temperature of -380'C at which point a saturation level is reached. The relationship permitted reasonably accurate extrapolation of the limited data available at that time. Whilst the use of Calder effective dose has now been largely. superseded, use of the term still persists in certain Magnox reactor applications such as air. reactivity. It is again emphasised that whilst equation 2 may be applicable to the Commercial Magnox Stations, a different relationship may well hold for other systems where appropriate data should be sought. 1.4 Contamination Impurities create defects in the hexagonal lattice of graphite crystals which then act as sites for either enhancing or inhibiting the rate of air oxidation. Elements appearing in lower groups than carbon in the periodic table tend to be catalysts, notably metals. Many investigations have attempted to rank metals in order of catalytic effectiveness with very inconsistent results suggesting variability of experimental conditions and of the chemical form in which the metal was introduced., not to mention uncertainty in the initial impurity content of the graphite used. Some investigators have found lead to be the most catalytic metal which can give rise to increases of several orders of magnitude in graphite-air reactivity, whilst sodium, vanadium, manganese, silver, copper and ferrous metals all appear high on classification lists. Both lead and sodium (introduced as salt carried in the cooling) are suspected as contributors to increaSed air reactivity of the Windscale Pile graphite and hence as a possible cause of the 1957 fire. Very small amounts (- a few ppm) of some contaminants can markedly affect air reactivity. As has already been stated, E for pure carbon is -- 60 k cals.mole, and with almost any contamination beyond a few tens of ppm causes a reduction in E to about 40 k cals/mole. Further reductions in E are possible for particular contaminants or very severe contamination. Materials which inhibit air reactivity are defined as those materials added to the graphite which either seal the surface or which slow down the oxidation reaction by, for example, increasing the activation energy, E. The most significant inhibitors are the halogens (fluorine, chlorine, etc) and phosphorous, sometimes alone or in the form of compounds. One of the most effective inhibitors is POC13 , whilst CC 4 and CC1 2F2 are known fire retardants. Inhibitors are mainly of interest in the context of controlling or extinguishing graphite fires. Gaseous inhibitors are more appropriate in this context, and an inert gas such as nitrogen or argon would be suitable as a means of excluding oxygen. Carbon dioxide would inhibit graphite oxidation to negligible levels below about 7006C, but may not adequately suppress oxidation if other flammable materials are also present.\n4 2. Radiolvtic Oxidation in CO, 2.1 Introduction Two types of oxidation can occur in a CO2 environment, namely thermal and radiolytic.. Thermal oxidation is the purely chemical reaction between graphite and CO 2 which proceeds' at a negligible rate below 625\u00b0C and is not important even for the hottest parts of AGR fuel sleeves operating at temperatures up to 675\u00b0C. Radiolytic oxidation, on the other hand, occurs when CO 2 is decomposed by fast neutron and gamma radiation (radiolysis) to form CO and an active oxidising species which reacts with the graphite. Graphite is a porous material and the oxidation takes phase predominantly within the graphite pores, proceeding at a rate which is proportional to the mass of CO2 contained in the pores and the energy deposition dose rate absorbed by the CO 2, but is substantially independent of temperature within the range of practical interest. An explanation of the process of energy deposition in graphite is given in Appendix 3.1. Radiolytic oxidation is important because the resultant weight loss may affect reactor economic performance by reducing moderating power and increasing permeability, and can cause significant changes in graphite properties and loss of strength. It is usually necessary to control the rate of oxidation since the weight loss in pure CO, would be intolerably high even in the relatively low-rated Magnox reactors. The oxidation rate is more than halved if the carbon monoxide concentration is allowed to build up naturally to about 1 % by volume, and may be further reduced by the planned injection of minor concentrations of hydrogen and moisture. Whilst such measures are found to be adequate for controlling the rate of radiolytic oxidation in Magnox reactors, the more highly-rated AGRs require a more effective inhibitor such as methane which can reduce the rate of attach by about a factor of 10. 2.2 Standring Equation for Weight Loss Radiolytic oxidation in Magnox reactors is generally assumed to occur uniformly throughout the open pore structure at a rate proportional to the radiation dose rate to the CO2 contained within those pores. Standring\") has derived a general expression for calculating the oxidation rate which has formed the basis for all graphite weight loss assessments and experimental analysis in the radiation chemistry field for many years. The derivation is as follows: Consider a 1-g specimen of open porosity e cm3/g being oxidised by CO2 at pressure p lb/in2 absolute, temperature TOK and energy deposition dose rate, D W/.g, to graphite (and therefore to CO2). [Pressure is expressed in lb/in3 since these were the units most commonly used in reactor specifications.] Then the weight of CO, in the open pores: ep p 273 14.7 T where p. = density (s.t.p) of CO, (1.98 x 10-1 g/cm3).\n5 The dose rate to the gas in the specimen: =eDp P 273 W 014.7 T = Dp, P 273 10\"9 14.7 T 1.6 If G., represents the number of atoms of carbon gasified per 100 eV of energy absorbed by the gas, then in the 1-g specimen the rate of gasification :DpP 273 10T \" -c atoms/sec 14.7 T 1.6 100 eDp, - p 273 1019 G- 12 14.7 T 1.6 100 6x 1023gSOec 2.32 10-5 eGcDp pa g1se, T substituting for po = 1.98 x 10-3 g/cm 3, this becomes Rate = :L.656 x 10- x T gg,>h T EGGcDp 145 T % per year T (3) (4) 2.3 Controlling Parameters in the Standring Equation 2.3.1 Dose Rate The energy deposition dose rate term, D, is a measure of the total energy absorbed in unit time from the scattering of -y- radiation and fast neutrons in the derivation of equation 3 it is assumed that the rate received by the graphite is the same as that absorbed by CO, in the pores of the graphite. In 1962, using a theoretical approach it was shown that a fraction k of the fission energy from the fuel caused heating in the moderator. From this the following equation for the mean dose rate to the moderator was derived: 6 (.\nk.P D -E wig D=W/ w where: P = Reactor thermal power, MW W = Weight of active core graphite (excluding reflectors), tonnes k is typically 0.056 for the Calder reactors but higher at about 0.086 for -AGRs It is important to note that significant spatial variations in dose rate will occur within the graphite moderator and these are from two main sources. Firstly, there are axial and radial variations in dose rate within the core, so that axial and radial form factors must be applied to the mean dose rate, D, when evaluating peak values. The peak value may be determined from the equation: D D fafr where: fa mean fuel rating in a channel maximum fuel rating in a channel mean channel rating maximum channel rating Secondly , there wiI. be a marked radial variation in dose rate within the unit cell of any given reactor and this must also be taken into account when calculating the distribution of radiolytic weight loss within the unit cell. The variation is such that the dose rate will decrease in a radial direction away from the fuel element. It is possible to calculate the shape of the dose distribution curve for any given reactor parameters using reactor physics codes such as WGAM and MCBEND; a typical result for an AGR unit cell is shown in Fig 3.3. Finally , it is important to recognise that the dose rate, D, in equations 3 and 4 above is the dose rate to the graphite moderator at the start of reactor life. Loss of moderator due to radiolytic oxidation will reduce the weight of graphite in the active core, and hence the dose rate will increase with time. The effect on weight loss of the progressive increase in dose rate with time is dealt with later in Section 2.3.5. The reduction in dose rate with time which occurs in every unit cell due to the fall in fuel rating attributed to reactivity loss is usually dealt with by averaging out such effects over the life of the fuel.\n7 2.3.2 Pressure and Temprn..ature Terms The variations in pressure and tempe:-rature change the density of the gas in the graphite pores in the normal manner described by the Gas Laws. Experiments have shown that no other corrections for pressure and tempterature are required for dependency of other terms in equations 3 and 4. 2.3.3 The G-c Term The oxidation rate constant G-c is strictly only applicable to oxidations in pure CO2 and is independent of pressure, temperature and graphite type. For a fixed set of conditions, different graphites exhibit different oxidation rates due to their different open porosity as demonstrated earlier. By definition, G-c is a measure of the number of atoms of carbon gasified by the oxidising species produced adjacent to the graphite surface by the absorption of 100eV of energy in the CO2. G-c varies with coolant gas composition. Although estimates vary, the generally accepted value for pure CO2 is 2.35. The reason for any uncertainty is the experimental difficulty of ensuring the complete absence of carbon monoxide from the CO\u00fd. The mechanism of radiolytic oxidation is represented by the following equations: Gas Phase CO2 -radiation > CO + O* CO + O* -- CO2 Graphite Surface O + C > CO where 0* is an active oxidising spe.-cies formed by radiolysis of CO0. Carbon monoxide thus inhibits racliolytic oxidation by gas phase de-activation of the active ( oxidising species. The presence of only 1 % of CO can reduce G-c by a factor of about 2 , as may be seen from Fig 3.4 which is a plot of the experimental data that fits the relationship: G-.c = 1.05 + 1.32 1 + 8.5[CO] where [CO] is the carbon monoxide concentration in volume percent. \u2022 A--8 Further rcductions in G-c arc possible by the addition of low concentration-. of moi.Nture to coolants containing carbon monoxide. The data from experiments at 1.2% CO are shown in Fig 3.5 from which reductions in G-c due to moisture effects of up to 35% are apparent. Much larger reductions in G-c are possible by the use of methane as an inhibitor. The effect is shown in Fig 3.6 which demonstrates that methane is more effective in reducing G-c thaa \" is the case with moisture. The use of methane as an inhibitor of radiolytic oxidation has beer largely confined to AGRs where high inhibition factors are required to negate the effects of higher dose rates. For AGR applications, the use of G-c is not appropriate since with methane inhibition the oxidation process no longer proceeds at a uniform rate throughout the pore structure. This topic is covered in Section 3 of these notes. 2.3.4 The Porosity Term The initial open porosity, E., of any individual piece of graphite can be determined by measurement, and the value so obtained used in equations 3 and 4 to calculate the oxidation rate. However, as is described below, synthetic graphites usually exhibit in rapid increase in open porosity during the very early stages of oxidation and a more reliable estimate of the initial oxidation rate and its subsequent increase with weight loss is obtained by making allowance for this effect. To do this it is necessary to use the \"effective\" initial open porosity, F-,, instead of E. in equations 3 and 4. It should also be noted that the convention of expressing oxidation as a percentage of the original weight automatically requires porosity to be expressed in cm 3/g. However, when studying increases in porosity with in-pile oxidation it is more suitable to use cm3/cm 3, it. The total porosity, mtr, of graphite comprising both open pores, 7,, and closed pores, it, and these are related to bulk density, PB, by the following equations: \u00b1 C m /CM 3 (5) PC) =c P (i-M c 3 1CM 3 (6) PB (ra - 11 C m3 1CM 3 (7) ( Pffe P C 9 where: p = crystal density of graphite (2.23 g/cm\u2022) PH: = Helium density of graphite. g/cm' Note that the helium density is based on the total volume of crystals and closed pores contained within a given mass of graphite. Helium density is therefore lower than crystal density, but is higher than bulk density. The treatment of porosity changes in Magnox reactors where PGA is radiolytically oxidised in coolant gas comprising carbon dioxide with additions of carbon monoxide and moisture but with little or no methane present, and where the oxidation may be considered to occur uniformly throughout the porous structure, can be carried out using the following approach developed by Standringct Unoxidised PGA has an open porosity amounting to about 20% of the bulk volume of the graphite whilst a further 4 to 6% is closed porosity which is inaccessible to the coolant gas. Experimental work has shown\"'\u2022 that approximately 40% of this closed porosity is opened up before the graphite has sustained 2% weight loss, whilst subsequent oxidation only increases the o.p.v. in proportion to the weight loss, Fig 3.7. The constant of proportionality is the reciprocal of the helium density which is constant from 2% up to at least 30% oxidation, since this allows for enlargement of open pores plus an opening up of the closed pores, both occurring in proportion to the graphite consumed. There is evidence that much of this initial 40% of closed-pore volume opens up at very low oxidation (0.04%) and it has therefore been recommended that it should all be assumed to open up instantaneously at the commencement of oxidation. The subsequent increase in open pore volume can be represented by the equation: + PB (8) PHe 100 where: - = Open pore volume at x% oxidation, cm 3 /cma( 7te = Effective initial open pore volume, cm 3/cm 3 PBo = Initial mean bulk graphite density, g/cm 3 PHe = Mean helium density after 2% oxidation, g/cm3 Thus the initial rate of graphite oxidation given by equations 3 and 4 would use effective open pore volume, F, and become: 10 I eG~ Dp g0 = 1.656 x 10-'* T \"D gig hr (9) T cG.Dp 145 -c % per year (10) T where: ,, = initial rate of graphite oxidation per unit time It can be shown that allowance for the effect of increasing porosity on cumulative weight loss, Ct, can be made by use of the following equation\") , which assumes that: All gas in an open pore of any size is equally effective in gasifying graphite at the walls of that pore. ii. Dose rate to the graphite is constant. C: =a A xpA -V (1 where: Ct = Percentage weight loss in t years CPO Initial oxidation rate at t = o, percent per year, from equation 10 A = 107t (0 - rTO. Curves of cumulation weight loss, Cp, versus g&t for constant dose rate to gas are shown in Fig 3.8 for selected values of r;. 2.3.5 Cumulative Weisht Loss and Constant Reactor Power Equation 10 may be written in a different form in which the dose rate term, D. is substituted by kP/W as in equation 3. Thus: = 145kP % per year (12) II 'From equation 12 it will be apparent that the rate of weight loss will increase with loss of moderator mass, WV, if reactor power remains constant. Standring has again shown''' that cumulative weight loss, C,, for a reactor operated at constant power is governed by the equation: A log, (1 J C = go. (13) 100\"n-'\" Ao,1+\u2022 100 The use of this equation yields higher cumulative weight losses than are obtained using equation I 1 as may be seen from Fig 3.9 which gives the resultant curves of C, versus gjt, and which may be compared with the corresponding curves for constant dose rate seen in Fig 3.8. 2.4 DIFFUSE Code The treatment radiolytic oxidation in Advanced Gas-Cooled Reactors is radically different and much more complex than that employed in Magnox reactor weight loss calculations, and the reasons for this are as follows. Methane is a significantly more effective inhibitor than carbon monoxide. It inhibits by forming a protection species on the surface of the graphite pores which then competes with the graphite for the active oxidising species that escape deactivation by the carbon monoxide. The mechanism for radiolytic oxidation and inhibition can be represented by the following equations: Gas Phase CO, > CO + O* CO + O* > CO 2 CH, --> P where 0* is the active oxidising species formed by radiolysis of CO2 and P is a protection species formed from methane oxidation.\n12 0:1: + C > CO 0* + P > OP where OP is the deactivated gaseous product of methane destruction. Experimental data indicate that the protective species is extremely effective in suppressing radiolytic oxidation, and indeed for methane concentrations of practical interest, oxidation in the large majority of the pores is virtually eliminated. However, it appears that methane is denied access to some parts of the pore structure, and it is in these pores - known as the reactive pores - where the oxidation is concentrated. It is for this reason that the Standring equations no longer aVply and, as will be seen later, a different approach is required. The methane that is destroyed within the graphite pores must be replaced if inhibition is to remain effective. However, the rate of methane destruction is quite high in relation to the gas flow rates due to permeation and diffusion, and hence there may be a significant reduction in methane concentration at positions within the bricks which are remote from external surfaces. A further complication arises from the fact that methane destruction and radiolytic oxidation result in the formation of water and carbon monoxide within the pore structure, and for similar reasons the concentrations of these constituents builds up to higher levels within the bricks. Since both methane destruction rates and radiolytic oxidation rates are functions of all these minor constituents in the coolant gas, it is necessary to determine their distribution within the graphite bricks taking into account not only pore structure changes but also the numerous external factors which influence these complex processes. The governing equations for evaluating coolant composition distribution within two dimensional geometries must be solved by computer techniques, and the DIFFUSE code was developed for this purpose. Having solved for methane, carbon monoxide and water profiles the code is then required to determine radiolytic oxidation and weight loss profiles as a function of time, and this is done using finite elements. Several versions of DIFFUSE have been developed, the more important of which are DIFFUSE 4 by Russell and DIFFUSE 6 by Davies\"'. Details of:the DIFFUSE 6 code are given in Section 2.5. 2.5 Controlling Parameters in DIFFUSE 6 Calculations 2.5.1 Diffusion Equations The basic unknowns in DIFFUSE 6 are the methane, C,. moisture, C 2, and carbon monoxide, C 3, gas concentration profiles. These are calculated from the following set of second order differential equations: 13 VT (D 1o V(CQ)) - V(v.Cl) - K( VT (D20 V(C 2)) - V(v.C\u2022) + K, STOX + 0 VT (D30 V(C 3)) - V(v.C) + K1 STOX I + K2 STOX 2 + 0 (15)\" (16) The first term in each equation is the pure diffusion contribution, and the diffusion coefficients Dto, D 20 and D3o are the effective diffusion coefficients in graphite of methane in CO,, moisture in CO, and carbon monoxide in CO,, respectively. (See Section 2.5.2). The second term in each equation is the contribution from porous flow (due to permeation), and v. is the velocity vector for CO2 flow through porous media. (See Section 2.5.3). The second term in each equation is the contribution from porous flow (due to permeation) and is defined in Section 2.5.3. The last term(s) are the sink and source terms which define methane destruction, and moisture and carbon monoxide formation, and are defined in Sections 2.5.4 and 2.5.5. 2.5.2 Diffusion Coefficients The free gas diffusion coefficients of gas I and gas 2 (eg methane in and carbon monoxide in CC 2), D 1,, are obtained from the expression: BT- 1 1 ,j \" + \" 2 CO, moisture in CO2 (17) where: B [10.7 - 2.46 x 10-4 14 C (.\n-- 0 (14) T= Temperature, K M, Nil = Molecular weights of the two gases P = Absolute pressure in atmospheres r= Mean collision diameter for the two gases, Angstrom ID= Collision integral for diffusion given as a function of kT/E 1, The collision diameter and integral used in equation 17 are obtained from standard tables using the mean value of -,A/k defined by: where: k = Boltzmann constant. 1.38 x 10.6 ergs/k E = Energy of molecular interaction Standard values of e/k are available for carbon dioxide, methane, carbon monoxide and water. The effective diffusion coefficient in graphite is very much lower than in free gas and is given by: De. = IDII\" (18) where X is the ratio of diffusion in graphite to diffusion in free gas. Dto, D30 and D3 0 in equations 14, 15 and 16 are the effectivediffusion coefficients in graphite of methane in CO,, moisture in CO,, and carbon monoxide in CO2, respectively. The diffusivity ratio, X, will increase with weight loss due to the increase in open porosity. DIFFUSE 6 incorporates several options whereby X is allowed to vary either linearly or quadratically with either weight loss or dose. The variation can also be controlled by use of threshold values below which a linear variation is obtained and above which a quadratic variation is obtained. 2.5.3 Porous Flow When graphite is subjected to a pressure gradient, VP, gas will permeate through the graphite pores (porous flow) in accordance with Darcy's equation: 15 T v1 VP (19) where: v = Velocity vector for CO, flow through porous graphite (see equations 14 to 16) Bjp = Permeability coefficient for virgin graphite, cm 2/atmos, s. Under impressed flow, the pressure profile under steady state conditions is determined from continuity (conservation of mass) considerations using the equation -V(pv) = 0 ie V (.FL) (1.0VP) 0 where: R = Universal gas constant. which may be written: A *.V (P.VP) = 0 where A B .,)(I or or K.XV (P2) = 0 where K A2 Cr.. The finite element method is used to calculate the pressure profile which is then used to calculate v during th6 gas concentration calculation which is in turn fed into equations 14, 15 and 16. In DIFFUSE 6 the graphite perme2bility coefficient may be calculated using options which allow BIp to vary either linearly or quadratically with either weight loss or dose. The variation can also be controlled by The use of threshold values below which a linear variation is obtained and above which a quadratic variation is obtained. 2.5.4 Methane Destruction. K, Methane is destroyed by radiolysis at a rate which is proportional to the local values of dose rate, open pore volume and G(-CH4) value, the latter being known as the methane destruction rate parameter and is defined as the number of molecules of methane destroyed per HeV (HeV = 100 eV) of energy absorbed by the CO, in the graphite open pores. The destruction 16 of methane gives rise to the formation of carbon monoxide and water in accordance with the equation: CH 4 + 3CO, -- 4CO + 2 H,O This equation defines the normally accepted values of the stoichiometry for H:O formatioh from CH4 destruction (STOX = 2), and for CO formation from CH, destruction (STOXI = 4), used in equations 15 and 16. The sink term for methane destruction, K,, in equations 14 to 16 is defined as: K = B2.M.W(p).V(p).G(-CH,) x 106 (20) where: B, = Physical constant (= 1.036 x 10W) HeV Mole/J Molecule M = Molecular weight of CO, W(p) = Dose rate to graphite at mesh position (p) watts/g V(p) = Total open pore volume at mesh position (p) cm-/cm G(-CH,) = Number of methane molecules destroyed per 100eV of energy absorbed by CO, contained within the graphite pores. G(-CH 4 ) is a complex function of the methane, carbon monoxide and moisture concentrations, and temperature. The recommended expression is: 2.2 G(-CH4,) -. 1 + [CO] .xp (4900) 5 +k6. [H20] } (21) [CHi, [C O] where: [CH 4], [CO] and [H20] are the concentrations of methane, carbon monoxide and moisture, respectively, vpm. R = Gas constant (= 1.9869) cals/*C Mole T = Temperature, *K k5 and k, are constants (k, = 0.00208, k6 = 0.0868) K, (equation 20) is then used as a nodal load vector in the finite element formations. In DIFFUSE 6 the total open pore volume is assumed to vary with weight loss according to 17 the relationship: V(W) = V0 + 1(p) (I -Vd) where: V. = Initial total open pore volume, cm3/cm 3 i(p) = Mean weight loss at a point over the current time increment, %. The dose rate variation with weighT: loss is defined in Section 2.5.6. 2.5.5 Graphite Oxidation. K, Carbon monoxide is also produced from graphite oxidation in accordance with the equation: C + CO2 -) 2CO Again, this equation defines the normally accepted value of the stoichiometry for carbon monoxide formation from graphite oxidation (STOX2 = 2) used in equation 16. The source term for the production of CO from graphite oxidation , K:, is defined as: Y2 = B.,.M.WB).Vo.AR 1 ..-!p- ) x 106 (22) From a comparison with equation :20, it may be seen that: K2 = B2 .M.W(p).V(p).G(-c) where G(-c) = AR B,-ips \"Vs and V() = P) AR. = Initial graphite attack rate at 674K and 41 bar for the coolant composition at mesh point (p), glg.h. mw/g, B, = Constant for DIDO rig (= 0.069302 x 108 when rig temperature and pressure are! 673K and 41 bar, respectively) 18 p, = Graphite density in DIDO rig specimens (= 1.8) g/cmr V, = Open pore volume in DIDO rig specimens (= 0.121) cm'/cm3 V0 = Initial total open pore volume, cm'/cm, fi(p) -Factorial increase in rate of weight loss due to all relevant factors at mesh position (p) , as defined in Section 2.5.7. The graphite attack rate parameter is a function of methane and carbon monoxide concentration which may change with respect to time. The recommended values of AR. for Gilsocarbon graphite at 400'C and 41 bar are given in Table 3.1 and illustrated in Fig 3.10. 2.5.6 Dose Rate Variation In order to include as much flexibility into DIFFUSE 6 as possible, several options are included to allow for increases in local dose rate as a function of mean brick weight loss at any given time. The dose rate, W, at any position in the brick cross-section is thus assumed (at time zero) to be governed by the equation: W = W [p' + A + v'.g] d(r) (23) where: W = Mean local dose rate to graphite, watts/g S = Constant fraction of dose rate A = Incremental fraction of dose'rate which may vary with time \" \"= Fraction of dose rate associated with moderator/fuel ratio, g d(r) = Dose rate factor in brick Increases in dose rate due to weight loss are assumed to occur in the ratio: Factorial increase in dose rate 1 (1 -\u2022 100 where: = mean section weight loss, % The options available in DIFFUSE 6 for increase in dose rate due to radiolvtic vweight loss are: 19 W =i + V.kgd(r) 100 W = 1l[gi, + A + V, dr 100 W = WV[p' + A +v.'g] d(r) The latter equation represents no effect of weight loss on dose rate. 2.5.7 Reactive Pore Volume (RPV) Model The RPV model is based on the theory that radiolytic oxidation in inhibited coolant is concentrated within a small fraction of the total open pore volume known as the reactive pore volume. The oxidation rate within the RPV is a function of gas composition, and the rate of graphite weight loss is assumed to increase proportionately with increase in RPV. Although the RPV model can be specified in terms of several pore groups in which each group is characterised by its pore entrance diameter (PED) or pore entrance width (PEW). DIFFUSE 6 is limited to a single pore group. For RPV calculations, the initial RPV at a given mesh point p, RPV'o(p), assumed to be directly proportional to the initial ra.diolytic oxidation rate: AR RPDV' 0 (P) = OnV.. '0 Experiments have shown that for a standard initial graphite attack rate, ARs, in the DIDO rig of 0.2 x 10.8 g/g.h.mW/g at 41 bar and 673K, the initial RPV value, OPVo, for a single pore group is 0.03 cm 3/cm 3. AR. is the initial graphite attack rate at 41 bar and 673K for the coolant in question (from Table 3.1 and Fig 3.10). It should, however, be noted that DIFFUSE 6 has the option of performing a Strandring type calculation for which jRPV'Q (P) = OPVO High weight loss experiments in inhibited coolant have shown that the graphite oxidation rate does not go on increasing expontentially as predicted by the Standring relationship for PGA 20 radiolytically oxidised in CO.. The results indicated that the rate (and by implication RPV) increased with weight loss by about a factor of 3 after which there was no further increase in attack rate, ie RPV had saturated at 3 times the initial value, as may be seen in Fig 3.11. This rule has been assumed to apply regardless of initial RPV, and the calculation routines therefore incorporate a pore efficiency factor, F, to adjust RPV in accordance with one of the following relationship RPV(p) = RPV (p).F(PED) for cylindrical pores RPV(p) = RPV ' (p).F(PEW) for slab-shaped pores where F(PED) and F(PEW) define the reactive pore efficiency factor versus pore entrance diameter or width. Tables are available giving values of efficiency factor to give saturation of RPV following oxidation to 3 times the initial RPV. The initial rate of griphite oxidation at mesh point (p) is given by: P Rate,(p) = 9.92076 . 10 ARo(p) -f W\u00b0(p) % per year (24) T where the numerical constant is obtained from: 100 T,. 24. 365.25. 10 3 1%.14.504 with PS = Pressure in experimental rig (41 bar) T= Temperature in experimental rig (673K) ARo(p) = Initial graphite attack rate at 41 bar and 673K for the coolant composition at mesh position*(p), glg.h.mW/g. P and T are pressure, psia, and temperature T, K, at position (p) W.(p) = Initial energy deposition dose rate at position (p), W/g. Having defined the initial rate of graphite oxidation. Rateo(p), in equation 24, the weight loss in the it, interval of time Ax.(p) is given by: Ax,(p) = Rate, (p).fi(p)dt X i(p) = xil(p) + Axi(p) where fi(p) is the mean factorial increase in rate of weight loss over the i1h time increment and is defined as: ( RPV, (AR 'o (25) fi= V' -(i) - ARQ) 21 where: ( RPV = Mean factorial increase in RPV averaged over the ith RPV,,1 time increment (including an allowance for pore efficiency changes). ( \" ) = Mean factorial increase in dose rate averaged over the ith time interval (including allowance for weight loss effects) ( AR. = Mean factorial change in initial attack rate averaged SARo over the ith time increment due to changes in coolant composition. Appropriate techniques are included in DIFFUSE 6 to ensure correct integration of weight loss in each time step. RPV is assumed to increase with loss of graphite at crystal density, ie no allowance is made for opening up of closed porosity, hence: PB. Xj( ) RPV. RPVo + -.\n( PcC I00 However, an option is available for allowing reductions in RPV due to fast neutron pore closure: PB 0 x 1(p) RPVi (p) = RPV* + PB\" -closure PC 100 Summarising, the essential steps in DIFFUSE 6 are as follows: For each time step calculate: i. Diffusion and permeability coefficients updated to allow for the effects of weight loss. ii. New pressure profile in brick section. iii. Factorial increase in dose rate due to weight loss. iv. Gas composition distribution. v. Initial graphite attack rate at each mesh point corresponding to new coolant gas composition distribution.\nvi. Factorial changes in initial graphite attack rate (since startup) due to changes in coolant gas composition. vii. Factorial increases in RPV due to weight loss. viii. Overall factorial increase in the rate of graphite weight loss due to changes in dose rate, gas composition and RPV. ix. Total weight loss at each mesh point by appropriate integration of the weight loss in each time step. 3. Effect of Radiolvtic Oxidation on the Physical and Mechanical Properties of Graphite Experimental data on PGA graphite radiolytically oxidised in nominally pure CO.,, and on AGR moderator and fuel sleeve graphite radiolytically oxidised in methane-inhibited coolants, has indicated that the thermal conductivity, Young's modulus and strength changes due to oxidation alone were changed by an exponential factor e\u2022', where x is the fractional weight loss and a is a constant which is dependent upon both graphite type and property change. The following table compares values of the constant a for three property changes of three types of graphite. Property Change Gilsocarbon AGR Fuel PGA Graphite Moderator Graphite Sleeve Graphite (Nominal Pure CO,) (inhibited coolants) (inhibited coolants) Thermal +2.7 +2.3 +3.1 Resistivity \"K\" Young's -3.4 -3.4 -4.8 Modulus Strength -3.7 -3.9 -5.2 Ha _ _ _ _ _ _ _ _ _ _ In the above table K,. E 0 and q, are unirradiated values, whereas values of the properties.\nK, E and a are oxidised 23 Radiolytic weight loss should be assumed to have no effect on the coefficient of thermal expansion of graphite. :Studies of the effect of radiolytic oxidation on dimensional changes of graphite indicate no effect before the onset of shrinka;ge reversal (ie turnaround). However, the onset of shrinkage reversal was shown to be increasingly delayed with increase in weight loss. The effect is only relevant for Gilsocarbon graphite in AGRs. The rules for calculating the effect on dimensional changes of radiolytical weight loss inhibited coolants are detailed in the notes for Lecture 2. References I. Standring J. Calculation of the graphite weight-loss in Civil Magnox and Advanced Gas-cooled Reactors. Journal of Nuclear Energy, Parts A and B, 1966, Vol 20 pp2 0 1 to 217. \" 2. Russell P D D. A user's guide to the APC graphite weight loss computer code DIFFUSE 4. 1979. APC/R 1514 3. Davies M A. A theory manual for DIFFUSE 6. 1991. AEA-RS-5117. C.\n24 APPENDIX 3.1 ENERGY DEPOSITION IN GRAPHITE During the maintenance of the chain reaction in a reactor, energy is deposited in the: moderator because of the following neutron cycle. Fast neutrons due to fission in the fuel escape from the fuel after some small energy loss due to inelastic scattering. During the inelastic scattering process small fractions of energy escape from the fuel in the form of gamma radiation. The fast neutrons escaping from the fuel will be elastically scattered in the moderator, losing energy to the moderator in the process. The choice of the graphite pitch is such that a much degraded fission neutron spectrum diffuses back to the fuel. This spectrum predominates in neutrons having thermal energies. A small fraction of the thermal neutrons are captured by moderator atoms, producing gammas; a larger fraction are captured by canning materials, releasing high-energy gamma rays; the majority of the thermal neutrons, however, induce fission in the fuel, producing prompt fission and fission product decay gammas in addition to high-energy fission fragments. fast neutrons. betas, etc. Some of the neutron spectrum diffusing back into the fuel will be in the energy band in which resonance capture occurs in the fuel; high-energy gamma rays will be produced by neutron capture. Some of the high-energy neutrons returning to the fuel will produce fast fission in the fuel, resulting in further fast neutrons and gammas. To sum up, the energy deposition in the graphite is due to slowing down of fast neutrons and gamma rays (prompt fission, capture, fission product decay and inelastic scattering) arising in the fuel plus a small amount due to capture gammas arising in graphite. The components of the heating due to capture gammas and the majority of the prompt fission and fission product decay gammas will be proportional to the thermal neutron flux in the channel. The components of the heating due to slowing down of fast neutrons and inelastic scattering gammas will be proportional to the fast neutron flux. ).\nLecture 3 Attack Rate Parameter for RPV Model gi/gmn per hr per mW/gm x 10**8 Carbon Monoxide_ I Methane 2500 5000 [75000 10000 15000 20000 25000 0.0 I 0.822 0.641 0.505 I 0.411 0.312 0.270 I 0.247 I 25.0 1 0.656 0.548 0.442 0.366 0.286 1 0.253 0.235 1 50.0 T 0.526 0.470 1-0.388 0.327 0.263 0.237 0.224 175.0 0.423 0.404 0.341 0.293 0.241 0.222 0.214 100.0 0.343 0.348 0.301 0.262 0.222 0.209 0.204 125.0 0.280 0.301 0.267 0.236 0.205 0.196 0.194 150.0 0.230 0.262 0.237 0.213 0.190 0.185 0.186 175.0 0.192 0.228 0.211 0.193 0.176 0.174 0.178 200.0 0.161 0.200 0.189 0.175 0.164 0.164 0.170 225.0 0.137 0.177 0.170 0.160 0.152 I 0.155 0.163 250.0 0.118 0.157 0.153 0.146 0.142 0.147 0.156 275.0 0.10J. 0.140 0.139 0.134 1 0.133 0.140 0.150 300.0 0.092 0.126 0.127 0.-12 0.125 I 0.132 o0.144 3 325.0 0.083 0.114 0.116 0.114 0.117 0.126 0.138 350.0 0.076 0.104 0.107 o0.106 0.111 0.120 0.133 Table 3.1 - Attack Rate Parameter as a Function of Methane and Carbon Monoxide Concentration for Gilsocarbon Graphite at 400 degrees C and 41 bar GRAPHITE TECHNOLOGY i .4) 6') ns .7 g\u0002 '-7 C? 4) .3 'I C) C, ''3 C, 5J .3) C, C-) I.\nC) (-) (11 I.) 0 C-) o *' C.) \u0002; C) 4. () C) ts I 4) I-.\nN a 'Ii (0 C*) if) C) ~... ....... if in1 '.3 4, \"If I '' I.. I) 43 7 .7 -.\n/ 7 / 7 7 1! cl 6, 61 4, -- fr*~ 3ill I I if :': :,;, / 7 / 7 .7.-, --'-7 .7, 4) 5; *r. I'. \"I 3, isCI 4) 6-I 63 * -... -C) \"I 4) 6. I:. 3, 43 :3 * 6'3 . I 11 illI1 I I I -ill I 1 1I 1 - 1 -C) 'Ii (\"I r\u0002) L U C.) 1111 I'I* I--I--\u0002-JII II I 1 'I *\u0002*I'- \u000211I II -51 6\u0002 U (\u00021 II) I-I It) g\u0002J A P4 \u0002' I U (I I () U I.,' bi) (\u0002j II) * C, U ti) bi) \u0002.i2 ''I -I (\u0002J (;. * Li. 'I C) \u0002 (3) * (CI I...\n;00.r * LI 4) r. '''I 6.s A Ij') .4 \"5 I I C) C) C.) III K-.\n(Ifi/fiii AIIAII:)I.';.)Il Lecture .3 VARLA'CN ) W!T CALE~ E~iVAL =iT EYATURE 0-5 ..~. .. . . . .\n.. \u0002ZUNA\u0002 / ________________\u0002\u0002.\u0002\u0002\u0002\u0002 W?\" E \u0002 L.A CLZL. D G SIa -ZM -K (42~2:ANS ::=NS-A-N7 x U-- C ===CIVE' 7 cN N;' \"CYt C.' I-I- -\u0002''\u0002\u0002N -\u0002 \u0002'\u0002: ____________________ I I -' 'C- I '\u0002.COO :ZCC :CE S 2 :C370 ________________________ S'Z 23S 2:~ :C.Ooo --'--C :S CC OC25 = Cs C7C . .\n. .. .\n-7 7 -i m CC 3CC I 3ICCC SCLE. UIVALE.47 -: Fizure 2. GRAPHITE TECHNNOLOGY 6 5 IO IS 25 IFigure 3. Section throucgh core of typical Civil (lose Ia;] within a 30'10 45 A .G.I(. showing variation of graphite init cell.\nG IRA III I IT'1\"\u2022 TIECA I N 0 1,) 0(G Y Lccturc 3 I I 'I)()'U IONO~ OI LI1!M 3D o tOII!UAI )I'NL .....\n: .... .. .\n.. .. . ................ -Y vp lt.: ... Ar ... ........ jyn~ lEJtfl... ~ t)'t\" ~ Pi &- .: .\na I *ut 1oil S .-.. 7 ::>:: 0 +:-\u00021 -9 C I e \u00020 0,\u0002w -\u0002\u00021'\u0002\u0002 .-... -. ____________________________________________________ / ________________________________ _____ _________ !\u0002- -:- :\u0002.--- T ~ T .\n:. ..-- -2... .- .I --77 \" ___ ..\n___ -___ ______ *1 *; \" .: _______________ .................. __________ .\n________ ______-_____________ .\n--.-.. ...... ... \u2022. .\n.. .......... -Lecture 3 vs. I *\u0002 Ft &'F 41 4.I; 4 X Figure 6. The Effect of CH 4 Concentration on G, for Mixtures Containing 1.0 - 1.5% CO GRAPHITE TECHNOLOGY ,XDO O1OC'HJDl 31IHclVID o021qd1D V'Od JO iNtSOJOd 012 uo SSO'-S 1 Z4 3!AX O11d - uI JO lTJJ9 1q.L tL ninz!Aq .\n...... .\n.\n. .. .. ... .\n... -.\n... ... .... .. .. ... .. .. .\n.... .-.. .\n.v: ......-.... ............ ..... ....... ........................ .. ... ..... -.\n-... .. ..... .... ... .\n...... ........... .- .\n.\n.- ..... ..... .\n.\n.~I ............. ... ... .. ............ ..... .\n.\n....... ...... .. ________________ .........\n___ ___ . .6.2--....... ..... -7 -i ..... ............. . . .\n.\n.\n...... .... .......1z.. .... ............... ... . .. . .\n.\n....... . .... . ... ........ : .. ..\u2022 ....:. :. .\n: . . . . .\n. . .. . . . . \u2022 .\n. - . : _ - .\n.\n. .\n. .\n.\n.\n.\n....\n: .. .+ ....-- : .;: ITI ... .\n..... \" ...... ..... ..... .. . . ... ........ -\u2022 \" - /. - .. ... -............ .... ... . .. . .... ..- ... . ....... : ..... .......... .\n- \u2022 \u2022 \" ' .. . j \u2022 _ \u2022 . .\" -\u2022 \u2022... .. . \u2022 : \u2022-2 \u2022 .~ ii\u2022::\u2022::: i :+?; III:: ;: \" ::: \u2022 ..... ? .. .......... ... .... ....... ... .-- : -.. .. .\n.\n.. ....... . .\n...\n\u2022 .... .. ...- .. .. ._ _... ...\nu... ..... ...... ...... -' ...\n: -............\n\" .. ... .\n.\n............ .\n.... -.. ..... .\n....... .. \" .\n.\n..... .\n.\n... .. \" .\n.. .\n......... ..~ ~ ~~..... ..... . ...... .... \" ...\u2022 ' .......\n: \" -.\n.: : \" : - : . :... . \" .v :: - -: . \" :;: .. , \" : - \" .\n\u2022 \u2022.2o\u2022.....\u2022-.............. ............\n\"\" ; \"': ' \u2022 -- \" '\u00b0 : \" \u2022 \u2022\u2022 T\u2022\u2022 \u2022 \u2022 \u2022 \u2022'- 7 '\u2022 ... ... .. .. .- .\n...... .... ..........- L . .\n: .\n.\n.. .. .\n.\n.\n.\n.\n.\n.\n..... ... .. . .- a:. .\n. .. . .\n.. .. ... .. ... ... . .. . +- ..... .\n.. ..... ... .. ... ...- ; :..... ::.. : : ..- .\n....\n; .\n; :..... : : .\n..\n:; . .:.: .\n......\n, . ..:....: .... .. . ..... . ... .\n,...:... .. .. . \u2022 , '. .\n. . . .: \u2022 . . . . \" \" ' : :\".. .. ..... . ... .... ... ..... .\n.: . .... :2 : \u2022 : -' -;. K -:i '-. ' .... \"..--. .\n....\n: .;..: . j : .' . . ...:. . .\n.\n.\n.\u2022. . . . . .. ...........\n\" .\n.- -...- ....\ni.:: .\n........ ...-.\n\u2022.- . -'\u2022 .\u2022. 7 \u2022 . .-... 7. Z7....\u2022:: !\u2022.7\u2022 : -. - .\u2022 \u2022 \u2022- . r -\u2022- : : .. .. .. .\n.... ..... .... .. .\n+ .. .\n.\n. .. .. ..... .. .\n,. . .... .\n.\n.. ... ..\n. ...\n\u2022-: .- -..- - - -- . .\n...-.......-..-.... .. .... .... .... .. ... o; 7 \u00a3. .\n... .\n.\n.\n.\n. .\n.\n.m .\n.\n.\n.\n..\n1. .\n.\n.\n.\n.\n\u00021 C; C a ) Lecture 3 go .\nINITIAL OXIDATION RATE AT TIME t.= 7e - EFFECTIVE INITIAL O.P.V. IN cn 1/cr.\nI JO q&o Figure 8. Effect of Increase in Porosity on Cumulative Weight Loss when Dose Rate to Gas is Constant .\n.-\u2022RAPHITE TECHNOLO GRAPHITE TECHNOLOGY 0 -J 0.,' O.\n(\".' Lecture 3 qc=INITIALOXIDATION RATE AT TIME = TT- EFFECTIVE INITIAL O.P.V. IN cm7/c,', ie 0 0 0 0 -0 400 v c 6 6o' o'6 6 ' ' , ,I I I?! \" V I // i I --____________i_!'__ I/ I I I 1~iII) I I -i If, I'i I I/i II ' i i /: /I ,,Ljjj : i: I ,, ' I h ' Il IZ I , r I 'I / I 10 Sct Figure 9. Effect of Increase in Porosity on Cumulative Weight Loss for a Reactor Operated at Constant Power \"\" TECHNOLOG Y GR.-P HITE TECHNOLOGY I00 I0 0 I-,, L) ..7 UJ / I 100 A ltctlure 3 00\" i 0.900 0 900 .2500 D ......\n7............ .................................. x 0.500 --o- 750110 S0.700 * 0.600 *- -..---- --- I.50 o 0.500 CIO 0.500- 75000 tit) 4)-- 0 0---- 10000 0.000__ -rT 20000 0.0 50.0 100.0 150.0 200.0 250.0 300.0 350.0 MethIlie CollceltIrtioll (vpmn) -0- 25000 i-igre 10. Attack Rate Parameter as a Function o\" Methane and Carbol Monoxide Concentration. for Gilsoc~arbon Graphite at 400 degrees C and 41. bar .. ................. .\n.......... .. .. ............ . .\n....... ..... ...\nGRAPHIT~E TfECHINOLOGY .\nE d -4> 4-c' RA --0 0 0* t '0 0 3 I \\ -\\ 0 Ii * A * 6 3 9 --J * I 6 I -9 S o -.* 0 .3 0* 0 S o 0 0 0 0 * ) ) I GRAPHITE TECHNOLOGY COURSE LECTURE 4 1. Moderator Structure Design in the UK The early designs of moderator structure in the UK such as BEPO and the Windscale Piles operated at atmospheric pressure and, using air as coolant, rejected the heat produced in the uranium fuel through a tall chimney. Thus the problem was one of shielding with overriding need for preservation of fuel and control rod alignment. Operational difficulties included the need to limit the build-up of stored energy in the graphite by regular annealing. With the higher power densities necessary for economic electric power production came the need for high pressure containment, increased coolant flows and significantly higher operating temperatures. The more severe operating conditions under which the graphite was required to operate necessitated major changes in moderator design aimed at mitigating against the dimensional and physical property changes suffered by the graphite under the combined effects of fast neutron damage and radiolytic oxidation in the CO2 coolant. These requirements intensified with the arrival of the Advanced Gas-cooled Reactors when not only improved designs, but also improved graphite became necessary. The following sections of this lecture trace the history of graphite structures used in the UK power reactors from Calder Hall onwards, outlining the development of moderator structure designs adopted to generate electricity safety and efficiently for the full design lifetime of the reactor. \") 2. The Calder Moderator Structure When the graphite moderator structure for the Calder hall reactors was designed, it was erroneously believed that fast neutron irradiation caused the graphite to grow in both the perpendicular and parallel directions to the extrusion axis of the bricks, and that the ratio of these growth rates was respectively about 6:1. In plan view the structure was to be a regular 24-sided polygon, 1100cm across comers by 800 cm high, with the fuel elements placed in 1696 vertical channels 10.6 cm diameter spaced on a square 20 cm lattice pitch. The predicted growth rates were such that if the core had been constructed as a solid structure comprising bricks 20 cm square by 80 cm high with their extrusion axes placed vertically, the diameter would have increased during irradiation by a maximum of about 16 cm and the height by a maximum of about 3 cm. The structure would have come barrel-shaped and resulted in unacceptable inter-channel leakage due to the tilting of bricks, and caused difficulties in inserting and removing control rods and fuel elements due to the distortion of channels. These difficulties were overcome by adopting the form of construction shown in Figures 4.1 and 4.2 in which the core comprised layers of bricks slightly smaller than 20 cm square having their extrusion axes placed vertically, interleaved with two layers of rectangular tiles having their extrusion axes placed horizontally and parallel with the long sides. It was \"argued that if all the tiles in the bottom layer were arranged with their long sides running I North to South forming parallel chords across the core face, and those in the top layer were arranged with their long sides running East to West, and garter restraints were placed around the periphery of the tile layers, the bricks would be located in their correct lattice position with a gap around their four vertical sides. Displacement of the channels during irradiation \"would thus occur at the lower axial rate. With this arrangement, however, the tiles become oval during irradiation and this would have caused them to interfere with fuel handling operations and made spigot location betweef bricks and tiles impracticable. To obviate these difficulties the tile bore was made slightly larger than the brick bore, and the bricks and tiles were located relative to one another by means of the cruciform keys shown in Figure 4.2. In addition, to minimise neutron streaming through the Wigner gaps, the bricks were rotated by 20 about their vertical axes relative to the lattice centre lines, each layer being twisted in a direction relative to those immediately above the below it. Control rod holes were formed by machining quadrant fillets on the long edges of four abutting bricks. The peripheral reflector was built solidly, ie without Wigner gaps as the growth in this region was estimated to be neglig!ible. Small gaps were, however, provided in the inner portions of the upper and lower reflectors, and the tile layers at the top and bottom of the core were made single tiles of much greater thickness than those in the intermediate layers. Circumferential garter restraints situated at each of the eleven tile layers were used to provide the necessary radial forces to hold the core structure together, and these restraints were suitably designed to accommodate the large tangential strains imposed upon them and to expand and contract thermally at the same rate as the graphite reflector. The core structure was supported on a diagrid over which a layer of abutting steel support plates was laid. Differential thermal expansion between the steel support system and the graphite structure was accommodated by supporting each column of bricks on a ball-bearing having flat tracks. Location of the core structure relative to the diagrid was effected by means of radial keys attached to the: bottom circumferential restraint engaging in keyways on the upper surface of the diagrid, and by spigotting the bottom bricks of the centre four columns into the core support plate. Steel castings, generally referred to as charge pans, were placed on the top surface of the C graphite structure to-form a robust funnel portion at the top of the coolant channel and hence facilitate fuel charging operations. The charge pans similarly guided the control rods, provided location for the bottom end of the charge chute, were a convenient point to attach burst cartridge detection piping and provide ballast to hold down the top reflector bricks. Although a separate charge pan could have been provided for each coolant channel, it was found most convenient to serve 115 channels with one charge pan. 3. Civil Magnox Reactor Designs The basic features of the Calder reactors were retained throughout the civil programme in particular the use of stacked columns of graphite bricks forming vertical fuel channels. Channel lattice pitching remained within the limits 19 - 20 cm with 9 - 10 cm diameter fuel channels utilising bricks approximately 80 cm high. Table 4.1 sets down the leading design particulars for each reactor in chronological order of building. The significant increase in 2 fuel channel numbers between Calder and the first two civil reactors will be noted. A further large increase in both fuel channels and output was achieved at Wylfa taking fullest advantage of the concrete pressure vessel. Changes to the graphite structure of quite fundamental nature were made during the design and construction stages as irradiation data at higher neutron doses (more approximating to. reactor life) became available from research reactors. These changes are described in the. following sections. 3.1 Dimensional Change In the Calder Hall design, end location of tile-to-tile -and tile-to-brick was obtained by keys and keyways placed on the 0\u00b0 and 900 axis of the channel. This form of keying was only retained at Berkeley. Later designs with tiles at Bradwell and Hinkley Point adopted off-set keys as shown in Figure 4.3. Larger ledges in the fuel channel occur on shrinkage but out of channel coolant leakage is reduced. The first Magnox stations planned to adopt a double tile layer as in Calder for lateral location of the graphite brick columns. However, as experimental irradiation data on dimensional change became available at higher doses, greater complexity of behaviour was identified. Shrinkage of graphite was shown to occur parallel to the direction of extrusion at temperatures between 150 and 450\"C and at temperatures above 300 0C in the direction perpendicular to extrusion with growth below 300'C in the latter case. Slackness in the structure following shrinkage became possible and the solution found was largely influenced by the stage reached in design and construction. For Berkeley, Bradwell and Hinkley Point 'A', which had reached an advanced stage in graphite machining, solutions using zirconium pins were adopted to maintain the channel lattice pitch as shown typically in Figure 4.3. The penalty of the low neutron absorption cross-section of zirconium is acceptable on performance considerations. The Trawsfynydd design was less advanced and a single tile at each brick junction utilizing radial keys on the 90 and 45 axes of the brick was adopted, located in a stable peripheral graphite arch, Figure 4.4. Later stations at Dungeness, Sizewell, Oldbury and Wylfa adopted side radial keying on the bricks. The peripheral graphite bricks were attached to steel structures which resulted in these bricks being moved radially as steel with change in temperature but the larger Wylfa reactor returned to the compressed arch type of construction to limit radial movement to that of graphite. With movement of the peripheral bricks as steel, those within form a complete mesh and, with imposed movement at the periphery, the lattice pitch of the fuel channel is opened and closed over the whole reactor. Absolute uniformity of lattice pitch in practice is not achieved as a working clearance must be provided in the keys and keyways requiring that allowance be made, in defining charge and control rod path, for the resulting brick freedom being taken up adversely. The high shrinkage along the axis parallel to extrusion had further consequences on the reactor design necessitating increased allowances for differential column shrinkage to cover spread of experimental data and variation of dose in adjacent brick columns. This had implications mainly on the charge pan design which surmounts the graphite structures 3 providing location for the fuel and control rod guide tubes. To avoid interference to fuel and control rod movement only small aLngular deviations from the horizontal can be tolerated. This is most readily achieved by a charge pan covering a large number of channels with well spaced three point support. However, a practical limit to size of a charge pan is imposed 'by the thermal movement between individual channels within the steel charge pan and accurate alignment of the fuelling chute to the channel. The charge pans carry the burst cartridge detection extraction points for individual fuel channels and may be used to support the sample piping on route out of the vessel. Berkeley, Bradwell and Hinkley Point 'A' usedc charge pans supported on the graphite with special feet to minimise tilt whilst Trawsfynydd used a system of individual channel charge pans with a surmounting upper plate to locate the guide tubes. Later stations suspended the charge pan from the pressure vessel standpipe, in certain cases with a permanent built-in chute for fuelling, allowing the graphite columns to shrink away with channel continuity maintained by flexibly mounted sleeves. Dimensional change data presented a further problem due to the variation of irradiation damage within individual moderator bricks. A brick at the core/reflector boundary or with a cross neutron flux gradient experiences higher shrinkage on one side than the other resulting in brick bow. As movement within the reactor structure is restricted at the tiled or keyed layers the columns cannot take up free standing positions and wedge shaped gaps result. From Hinkley 'A' onwards serviette ring type seals, made of magnox and located in grooves, have been incorporated at brick junctions to restrict out of channel leakage. (See Fig 4.5). Up to Trawsfynydd differential radial thermal movement between the diagrid/support plate and the graphite columns was allowed for by incorporating ball bearing support as at Calder. Access was possible below the diagrid for gag adjustment during commissioning, but with the required higher gas inlet temperatures movements were such that even with built-in cold pre-set, clearances against the diagrid member became inadequate. For Trawsfynydd the complete channel inlet assembly %as set above the levelling plate but it required that gag adjustment had to be carried out from charge pan level. Later reactors with inlet temperatures around 250\"C were built with side restraints moving radially as steel so that the lower reflector bricks in the moderator were spigotted directly into the levelling plates. On return to the rigid graphite reflector arch concept for Wylfa a system of rocking support pillars was introduced each carrying four graphite columns on a single levelling plate. This avoided use of ball bearings, which are limited by long term indentation and their low resistance to impact loading. 3.2 Stored Energy Build up of stored energy in the graphite increases with irradiation dose but anneals out faster at higher temperatures. Stored energy, therefore, influences design most in the bottom half of the reactor. Pessimistic extrapolation of data in the early design stages prior to Trawsfynydd suggested a severe problem and detailed studies were carried out to provide removable channel sleeving schemes. These introduced a stagnant gas barrier between the sleeves and the brick thereby proviiding thermal insulation from the coolant gas to raise the permanent graphite to a safe temperature. Removable sleeves were introduced into three of the later Chapelcross reactors, andi sleeves integral with the fuel elements were adopted at 4 Hunterston A and Tokai Mura. The need for this undesirable complication was removed when data improved, but the reactors chronologically later than Trawsfynydd all adopted gas inlet temperatures above 200'C to avoid the possibility of a self-sustaining release of stored .energy. 3.3 Fault Considerations At a very late stage in the design and manufacture of the restraint systems for Berkeley and Bradwell the possibility of failure of the main gas ducts was postulated as the worst credible accident. Assessment showed that pressure build-up in the Wigner gaps, resulting from escape of gas from voidage within the graphite under depressurisation, necessitated reinforcement of the restraint structures. The more onerous duty was met at Bradwell, Trawsfynydd and Dungeness 'A' by provision of a surrounding cage structure to restrain radial and overturning movements. Additional elastic garter restraints were provided at Berkeley and Hinkley'Point 'A'. At Berkeley the proximity of the cylindrical pressure vessel prevented gross movement and at Hinkley Point 'A' the problem of tilt was met by the provision of dowels at brick interfaces in the peripheral arch. The change to use of a concrete vessel for Oldbury and Wylfa removed the burst duct fracture as a criterion of restraint design and permitted use of a more restricted pressure circuit breach for gas pressure collapse. Features to meet earthquake conditions, eg stepping of the levels of bricks in adjacent columns as was adopted at Latina and Tokai Mura, have not been considered necessary for United Kingdom conditions. 4. AGR Moderator Structures The higher power densities experienced by the AGR cores necessitated the specification of an improved graphite (Gilsocarbon) and an upper moderator temperature of 500W C in order to contain irradiation induced dimensional changes at high dose within acceptable limits.( 4) The amount of coolant flow required to achieve this was too large to be taken in parallel with the fuel channel flow without causing unacceptable degradation of outlet gas temperature. \"The flow was therefore taken in series with the channel flow by directing about 50% of the circuit flow downwards through re-entrant passages formed around the moderator bricks and fuel sleeves to remove the 6% of fission heat which appears in the moderator. After passing through the moderator it mixes with the rest of the coolant which by-passes the core before flowing up the fuel channels. This feature was first introduced into the Windscale AGR prototype. This change in concept led to a significant increase in the diameter of the brick bore in relation to the lattice pitch, the ratio of these dimensions being approximately 2 to 3, as compared to the value of 1 to 2 used in the Magnox cores which reduced the material available for incorporating radial keys between the channel bores. Thus although the principle of radial keying the core components was retained, it was necessary to make significant changes in the arrangement of the bricks and keys.\n5 The design which evolved for the civil AGRs comprised large hollow 16-sided moderator bricks, one for each fuel channel, connected together by two types of key in a square lattice pitch arrangement as shown in Fig 4.6. Plain rectangular section keys (loose keys) were used between the abutting faces of adjacent bricks on the zero and 90\u00b0C axes, and cruciform keys integral with square interstitial bricks (interstitial keys) were used to connect adjacent bricks on the diagonal axes. A typical AGR moderator design is shown in Fig 4.7 which illustrates the Heysham II/Torness structure. The reactor core is a 16 sided stack of graphite bricks, connected at the periphery to a steel restraint tank. The core contains 332 fuel channels, 89 interstitial control channels and 163 interstitial secondary shutdown channels for the insertion of nitrogen and (for 32 of them) boron beads. The structure incorporates integral graphite and steel shields to reduce radiation levels so that personnel can enter the reactor pressure vessel for inspections. Columns of large polygonal bricks. 11 high , form the fuel channels whilst smaller square shaped bricks form either interstitial channels or act as filler bricks. In the radial reflector/shield the bricks are solid, but have small axial holes provided for cooling and insertion of steel shielding rods. The graphite bricks are inter-connected by graphite keys to provide stability of the assembly and to maintain the vertical channels on their correct lattice pitch. Direct access to each fuel channel is provided by charge tubes that pass through a steel pressure dome which segregates the cool high pressure re-entrant coolant from the hot low pressure channel outlet gas. The lower ends of these charge tubes spigot directly into the top of the core and it is therefore essential to match the thermal expansion of the graphite with that of the pressure dome. This is achieved by connecting the peripheral reflector bricks of the radially keyed core to a steel restraint tank as shown in Fig 4.7, and spigotting the bottom reflector bricks into the steel support plates which are, in turn, radially keyed to the diagrid. Cool re-entrant gas is directed over the inner surface of the pressure dome, the restraint tank and the diagrid thus ensuring near equal thermal expansion of these components. This arrangement enables the core to expand and contract \"as steel\" under the influence of thermal movements of the diagrid and restraint tank, and allows each brick the freedom to accommodate its own thermal and irradiation - induced dimensional changes without interference. The keying system is effective over approximately the centre third of each brick except in the top and bottom layers of the core where the keys are practically the full height of the bricks. This keying system extends to the core boundary where restraint rods connect the polygonal bricks to restraint beams at the inter-layer joints. There are 16 of these beams per inter-layer, each attached to the restraint tank by ball-ended restraint links. These links allow for vertical differential thermal movement between the steel tank and the graphite structure. The key/keyway clearances in the: top reflector are made much tighter than the rest of the core. This ensures that the top of each core channel and the bottom of each charge tube are accurately located throughout core life, thus avoiding undue distortion of the charge path. The tighter clearances are made possible by the much lower irradiation dose rates prevailing at the reflector level.\n6 The depth and width of the keys and keyways are optimized to provide maximum strength so that the structure can withstand anticipated core loads with an adequate safety margin when due allowance is made for loss of strength due to radiolytic oxidation. These loads .include static loads due to brick bowing, restraint movement and gas pressure differentials, and dynamic loads from a seismic disturbance. On top of the core is the upper neutron shield comprising overlapping square-section graphite and steel bricks . Below the core is the lower neutron shield comprising graphite bricks of similar design into which are inserted the steel fuel element support stands and their catchpots. The whole graphite structure is supported on, and located by, a series of large steel support plates which rest on a diagrid. Coolant flow through the core follows two routes. The larger re-entrant fraction flows downwards through the inter-brick gaps and sleeve/brick bore annulus to cool the graphite. The smaller bypass flow passes beneath the diagrid and then flows upwards through the lower neutron shield where it mixes with the re-entrant flow before entering the base of the fuel stringers. 4.1 Radiolytic Oxidation Radiolytic oxidation is an important consideration in reactor design because it reduces graphite strength, may change the dimensional behaviour of the graphite and, by removing moderator, may bring about the need for increased fuel enrichment. The addition of small concentrations of methane to the coolant gas is known to greatly reduce the rate of oxidation, but it was realised that methane destruction by radiolysis could bring about severe depletion of the inhibitor at positions within the moderator bricks remote from machined surfaces on account of the very low diffusivity of nuclear graphite. Calculations indicated that this could give rise to unacceptably high weight loss at the brick interior, and so several methods of improving methane access were investigated. Increasing methane concentration in the coolant was ruled out because this gave rise to unacceptable levels of carbonaceous deposition on the fuel cladding surface resulting in loss of heat transfer efficiency. Inducing a permeable flow of coolant gas through the pores proved impractical in the early AGRs since the required pressure drop across the brick wall was not available. The solution adopted was to drill a number of axial holes through the thickest parts of the bricks in order to reduce the length of the diffusion paths. By the time Heysham II and Torness were constructed it became possible to engineer a radial pressure drop of about 0.07 bar between the brick outer surface and the brick bore. This feature was therefore adopted as a means of inducing permeable flow of coolant through the graphite pores, although the drilling of axial holes was still retained as a safeguard. 5. AGR Fuel Sleeve Design Graphite was first used as a structural member in fuel elements when it was introduced into the design of fuel for Berkeley Magnox reactors in which each uranium fuel pin is supported between a pair of struts which engage in slots cut in the channel wall. At a later stage in Magnox reactor development the build-up of stored energy became important and graphite sleeves were introduced into the fuel element designs of the Hunterston \"A\" and Tokai Mura 7 reactors with the primary objective of raising moderator temperature at the cool inlet end of the core. The advent of the AGR with its higher gas outlet temperature and re-entrant flow arrangement has resulted in sleeves becoming an essential part of the fuel assembly. At the same time their duty has become somewhat more onerous on account of the more highly rated conditions under which they are required to operate. The requirements of the AGR fuel sleeve are: i. To provide support for a cluster of 36 fuel pins via the grid and braces of the fuel assembly. ii. To provide a thermrl barrier between the cool re-entrant and hot fuel channel gas flows. iii. To provide a secure pressure barrier between re-entrant and fuel channel coolant gas flows through which both joint leakage and permeable flow are minimised. (. iv. To retain dimensional stability under fast neutron irradiation. v. To possess adequate impact strength before and after in-pile exposure in order to permit safe on-load charge/discharge operations. A pitch coke graphite was selected as the sleeve material since this could be manufactured to a specification which most closely matched the above requirements. It possesses good impact properties, can be impregnated to give a low permeability, and although anisotropic is nevertheless reasonably stable under the required operating regime. Sleeve graphites having near isotropic properties have since been developed. An example of the graphite sleeve design first adopted for the Civil AGRs is shown in Fig 4.8 which illustrates the Hinkley Point \"B\" fuel element. The sleeve is of double-wall construction comprising one outer and two inner sleeves. The lower inner sleeve is screwed into the bottom end of the outer sleeve where it clamps the bottom fuel pin support grid. The upper inner sleeve, on the other hand, is retained by means of a screwed ring which also . clamps the upper fuel pin support brace. The free ends of the two inner sleeves have slots cut in them to provide location for the radial keys of the centre fuel pin support brace. An axial clearance is provided between free ends of the two inner sleeves to allow for differential axial shrinkage and thermal expansion, and axial creep of the outer sleeve under the influence of the static load imposed in the reactor by the weight of components above. For similar reasons, an axial clearance is provided between the ends of fuel pins on abutting elements. The inner sleeves are designed primarily as a thermal barrier. This is achieved by undercutting the wall of the inner sleeve to leave a location at either end and a radial clearance of 1.3mm over the major portion of the sleeve length, thus providing a stagnant gas space between inner and oute'r sleeves. A further function of the inner sleeves is to support and locate the stainless st:eel grid and braces which space the fuel pins, and to provide axial alignment of the pin assembly by keys registering in slots cut into the sleeves.\n8 The use of double-wall construction results in the outer sleeve being thinner and significantly weaker under transverse impact loading than would be the case if a single thick sleeve was used. Heat transfer calculations indicated that loss of heat through the wall of a single sleeve .would be tolerable, and so in view of the potential improvement in safety margin under impacts sustained during on-load refuelling it was decided to develop what became known as the Stage 2 AGR fuel sleeve. Figure 4.9 illustrates this design from which it may be seen that a much more robust sleeve is obtained in which the grid and braces are sprung and locked into three grooves machined in the sleeve bore. All new fuel elements loaded into the Civil AGRs now have Stage 2 graphite sleeves. 6. Assessment of Irradiation Effects The integrity of the graphite moderator structure is dependent upon the strength of its individual components. Thus brick shapes and key widths are optimised for maximum strength on the basis of tests conducted on representative slices of graphite bricks. Although a limited number of failures could probably be tolerated, the design criterion which has been adopted is that there should be no graphite failures. The condition of the graphite moderator bricks and keys will change significantly due to in pile exposure. The dimensions and shape will be modified, the physical properties and strength will be significantly altered, and the components will be subjected to both thermal and irradiation-induced internal stresses. Reliable assessment of these effects is an essential component of the safety case for continued operation of the reactor since failure to demonstrate the integrity of the moderator structure is potentially life-limiting. The various issues which must be addressed in the assessment of graphite moderator structures are briefly discussed below. 6.1 Brick Bowing The correct functioning of the core is critically dependent upon the clearances between keys and keyways. Thermal expansions and contractions of the steel restraint tank cause relative sliding of keys in keyways due to the lower expansion coefficient of graphite. Too small a clearance would cause jamming and heavy loads on the bricks and keys. Too large a clearance would allow excessive lateral movement of the brick columns and cause possible difficulties with fuel -cooling and control rod insertion.. Bricks will bow when one side shrinks axially more than the other due to the existence of a transverse damage flux gradient. Such gradients may exist due to the proximity of absorbers, empty channels or side reflector, but can also be the result of rating variations between fuel channels. Differential bowing between one brick and another to which it is keyed may be caused by differences in damage flux gradient, damage dose or material variability. It has the effect of reducing key/keyway clearance. This is allowed for in design calculations to determine key and keyway dimensions, but may require assessment in the later stages of core life when more reliable data on graphite behaviour become available. Another effect of brick bow is to bring about a small reduction in the effective bore size of the channels, although this is usually small enough to be unimportant. A more significant effect of brick bow is to cause a reduction in column stability due to the opening up of 9 wedge-shaped gaps between brick end faces. The resultant loads are not large, the main concern being whether or not alignment of unstable columns is maintained by the radial keying system. 6.2 Brick Barrelling Brick barrelling is a change in brick shape which is brought about by differences in shrinkage rate between the bore and outer region of the moderator brick which are, in turn, due to the radial damage flux gradient. The iradial fall-off in damage through the brick wall causes the bore region to shrink more than the outer region for the major part of reactor life. The net result of this differential shrinkage is to produce tensile stresses at the bore and compressive stresses near the periphery. In an infinitely long brick all plane sections would remain plane. However, in a moderator brick of finite length the lack of axial restraint at the brick ends causes the ends to \"barrel\" inwards so that the end faces become \"dished\". The effect is largely the result of the fall-off in axial stress towards the brick ends, and its main significance is in the effect it may have on brick-to-brick end sealing. 6.3 Brick Wheatsheafing ( Brick \"wheatsheafing\" may occur rather late in moderator life after the onset of shrinkage reversal if the growth rate at the btore is sufficiently greater than that at the brick exterior. This again is the result of the discontinuity at brick ends, but is opposite in sense to brick barrelling. The delay in shrinkage reversal brought about by radiolytic oxidation may well have the effect of significantly reducing or even preventing the occurrence of brick wheatsheafing. 6.4 Keyway Dovetailing Differential shrinkage across the moderator brick wall also causes hoop stresses as well as axial stresses. The outer region of the brick is in compression for the major part of life, whilst the bore region is in tension. However, the presence of keyways relieves the compressive stresses in the outer region causing the keyway sides to collapse inwards and become dovetail shaped. This change in keyway shape causes a reduction in key/keyway clearance which is allowed for in the design, but may not necessarily be adequate. In C. extreme cases the clearance between key and keyway can be lost altogether, and the key then becomes jammed in the keyway. This may give rise to additional stress being generated at the keyway root, usually referred to as \"key pinching stress\". It may also result in the generation of inter-column loads during thermal transients. 6.5 Thermal Stresses Temperature gradients will clearly be present due to heat conduction from the brick interior. These temperature gradients will increase due to irradiation-induced increases in graphite resistivity. The resultant differential thermal expansions will give rise to thermal stresses which will be alleviated by irradiation creep. However, when the reactor is shut down and the temperature gradients disappear, the thermal stresses will re-appear but in opposite sense, ie previously compressive stresses will become tensile, and vice versa. The shutdown thermal stresses may augment other stresses present in the bricks, and must be taken into 10 account in any assessment of brick integrity. The ABAQUS code is used to evaluate thermal stresses taking into account the various factors which influence their magnitude including irradiation-induced changes in thermal conductivity, coefficient of thermal expansion and .Young's modulus. The calculation routine may be modified to include the evaluation of key pinching stresses referred to in Section 6.4. 6.6 Shrinkage Stresses Significant variations in shrinkage rate will occur across the brick section due to the radial and circumferential variations in temperature and dose rate. These variations in shrinkage rate are largely accommodated by irradiation creep of the graphite, but nevertheless residual stresses and strains remain which may significantly reduce the ability of the brick to withstand external loading. The residual shrinkage stresses may well be augmented by thermal stresses during shutdown as described in Section 6.5. Calculation of shrinkage stresses is a complex procedure in which allowance must be made for the numerous effects of in-pile exposure and radiolytic oxidation on such properties as dimensional changes, irradiation creep and Young's modulus. The ABAQUS code is used for this type of calculation. The results of such calculations indicate that whilst stresses at the vulnerable keyway root are compressive during about the first half of reactor life, the subsequent onset of shrinkage reversal causes the stress pattern to change. When this happens, tensile stresses develop at the keyway root which is where the highest tensile stresses occur under external loading. From this it will be clear that shrinkage stresses are an important component in the assessment of moderator brick integrity. 7. Brick Loads Loads on the bricks and keys may arise from three main sources: i. Restraint movements. ii. Gas pressure forces. iii. Irradiation-induced brick distortions. iv. Seismic forces. An assessment is carried out to determine the capability of the bricks and keys to withstand these loads when it is necessary to make due allowance for changes in load-carrying capacity during life including: a. Reductions in strength due to radiolytic oxidation. b. Increases in strength due to fast neutron irradiation. c. The development of shrinkage stresses across the brick section. d. The development of thermal stresses during shutdown conditions.\n11 e. The development of key pinching stresses.\n7.1 Loads due to Restraint Movements In designs where the graphite core expands and contracts \"as steel\", relative movement between columns will occur due to the lower thermal expansion of graphite. Any tightness of the graphite radial keying system may limit free intercolumn movement and generate loads. The thermal inertia of the restraint tank and diagrid are also different, and so sudden changes in reactor operating conditions may cause these two components to respond thermally at different rates. The resultant differential radial thermal expansions (or contractions) will cause the graphite columns to tilt and thereby generate transverse loads. The number of columns affected, and hence the magnitude of the loads, will be dependent upon the difference in radial movement between one layer and another, and upon the amount of free movement permitted between adjacent graphite columns by the slackness of the graphite key/keyway system. The evaluation of loads due to boundary structure movement is done using classical statics, although computer codes can be used to solve the complex equations describing column interactions. 7.2 Loads due to Gas Pressure! Forces Although gas pressures in the re-entrant passages are nominally uniform, pressure variations do exist across the core due to zonal differences in cooling rate (eg between core reflector and side shield), the use of short or long keying, provision of cross flow headers and leakage of gas through boundary sealing keys. The pressure distribution is established from core flow network calculations which may be a 2D ring model network or a 3D octant model of the core. Pressure disturbances may also occur when fuel channels are empty during on-load refuelling operations. 7.3 Loads due to Irradiation-Induced Brick Distortions Loads from this source arise due to brick bowing and have already been mentioned in Section 6.1. The loads are not expected to be large because they arise due to pivoting of the bricks C about one edge (on the side where: axial shrinkage is lowest), and since the restoring forces required to maintain a \"pin-jointed\" column of bricks in equilibrium can be shown to be quite small. Much larger forces could arise if bricks in a given column were to bow in opposite directions due to some bricks having gone into shrinkage reversal. However, quite apart from the fact that this scenario is unlikely, it is considered that the transverse slackness of the radial keying system is such that the development of large bowing forces from this source is highly improbable. Evaluation of loads is by classical statics. 7.4 Loads due to Seismic Forces Transverse vibrations of the moderator structure may induce large inertia forces within the graphite core. These forces would be transmitted to the restraint tank via the restraint links and the graphite keys which connect each brick to its neighbours. The evaluation of seismic impact forces on the graphite keys and restraint links is extremely difficult due to the 12 complexity of performing a seismic response analysis on a structure comprising a large number of interconnecting components having several degrees of freedom. NCC have carried out such calculations for Heysham II/Torness using the computer code AGRCORE .at enormous expense. The results are difficult to validate. 8. Failure Criterion Experimental work has shown that when an AGR graphite brick is subjected to external loading via the keys and keyways it will fail in tension via a crack initiated at the keyway root. The calculated stress at failure, a,, is used as the basis for evaluating the structural integrity of graphite bricks following in-pile exposure after making due allowance for: i. Changes in graphite strength (ar,) due to fast neutron hardening and radiolytic oxidation ii. Shrinkage stress (aJ) iii. Thermal stress at shutdown (a) iv. Key pinching stress (ak) 8.1 Reserve Strength Factor Until fairly recently, combining the above stress mechanisms into a single failure criterion was achieved using the \"reserve strength factor\" (RSF) equation, defined as: RSF = Residual strength Stress due to applied load which, in terms of the mechanisms described above any be written: RSF = ac,- (a.+ a, + k)) -:\"- (1) O\"L where: CL is the load applied stress Thus, for an RSF - 2, the strength remaining in the brick is twice the load which must be borne. The RSF may thus be interpreted as a factor of safety for the component. 8.2 Fractional Remanent Strength Recent experimental work has indicated that there is a different value of critical stress associated with each of the above four stress mechanisms. Furthermore, it is considered necessary to apply separate correction factors to each critical stress when applying the results of slice tests to bricks and to allow for practical loading conditions. Three factors need to be defined: 13 Material Bulking Factor This reflects the changes in the strength of graphite as greater volumes are stressed. ii. Geometric Bulking Factor This reflects the difference in fracture load of a brick loaded along the full length and one loaded along only part length. iii. Key Tilt Factor This makes allowance for any tilting of the key in the keyway which serves to localise force transfer. It was considered that the RSF equation could not adequately represent the requirement for different critical stresses and correction factors, and that a more suitable approach was to use fractional remanent strength in which all the terms of the equation are ratios of predicted and ultimate stresses. The fractional remanent strength (A S) is defined asC2): aS = I- f .\u2022~hi'a2e stress 1 critical internal stress -[ shutdown thermal stress1 critical thermal stress -kev y pinching stress critical key pinching stress -[load applied stress 1 critical load applied stress The brick would be deemed to fail when AS is less than or equal to zero. The equation above may be rewritten as follows when including the various factors which need to be applied to the critical stress values: AS = 1 1 __ o + ] MCS, MIUM MtGOo cc, MLGLKoc,L (2) 14 where: C,, 0 1. CK and o. are the predicted values of shrinkage, thermal, key pinching and load applied stresses. cc,. ,a, a., and a,,, are the corresponding critical stresses M\", MO Mk and ML are the corresponding material bulking factors. GK and GL are the geometric bulking factors for key/pinching and load applied stress K is the key tilt factor. Values for the critical stresses and factors M, G and K must be determined by experiment and may be different for each type of graphite and brick geometry. References 1. Graphite Structures for Nuclear Reactors. Conference organised by the Nuclear Energy Group of the Institution of Mechanical Engineers. March 1972. 2. Davies M A and Judge R C B. Recommendations for Graphite Core Brick Failure Criterion. AEA-RS-5296, July 1992.\n15 TABLE 4.1 LEIADING PARTICULARS OF CEGB MAGINOX REACTORS Number Design Conditions Thermal Station of Fuel Pressure Peripheral Restraint Movement of Brick Location Interbrick Charge Pans Channels Containment Gauge T, T2 Peripheral Seals Pressure 6C 6C Bricks MN/im2 BERKELEY 3265 Steel 0.9584 160 345 Temperature As graphite Butting Tiles No Supported on Cylindrical Compensated Tie with graphite Bars Zirconium Pins BRADWELL 2624 Steel 1.0136 180 390 Temperature As graphite Butting Tiles No Supported on Spherical Compensated Tie with graphite Bars with oliter cage Zirconium Pins HINKLEY POINT 4500 Steel 1.3790 181 378 Temperature As graphite Butting Tiles Yes Supported on 'A' Spherical Compensated Tie with graphite Bars with dowelled Zirconium Pins reflector TRAWSFYNYDD 3740 Steel 1.7582 202 392 Temperature As graphite Radially Keyed Yes Supported on Spherical Compensated Tie Tiles with graphite Bars with outer cage cruciform Keys DUNGENESS 'A' 3932 Steel 1.9651 250 410 Steel cage with As steel Radial Keys Yes Suspended on Spherical puller rod control rod guide attachment to tube peripheral bricks SIZEWELL 'A' 3784 Steel 1.9306 214 401 Steel rings dowelled As steel Radial Keys Yes Suspended from Spherical to graphite and standpipes as radially keyed to a composite control/ restraint tank charge facility OLDBURY 'A' 3308 Concrete 2.5167 250 411 Steel rings hung As steel Radial Keys Yes Suspended from Cylindrical from boiler shield control rod guide wall with puller tubes rods attached to peripheral bricks WYLFA 6156 Concrete 2.7580 247 414 Temperature As graphite Radial Keys Yes Suspended from Spherical Compensated Tie standpipes as Bars composite control/ charge facility jiiiI~~ti~iY3Jijuva\u00b6I JJ(I1V JOMI-1110191J SAIM~71 flilIIa-.AP-fill! II!)jj ssoII 1t\u0002a\u0002 SU\u0002) PSS\u0002A ahIISS\u0002)I 1t II 9(1' flY \u0002 * IIIII!I \u0002sa)\u0002 \u0002P'.S tlqj, -SI lOd(I115 I inj 'fi.iIIIj\u0002; .... Xf .\n.. ..... . . .......\nLecture 4 THIS BRICK rwiSTED 2* CLOCKWISE -< a* INOICAT ES TMAT THIS DIMENSION IS IIIMINUS T\u00fdME WIGNER ALILOWANCE MADE A- -mAT PARTICULAR ZONE OFTM-- MODERATOR \"\\.-)ILE 2 INDICATES GRAIN A VI-W 01W ARROWA DIAGRAON SNOW 4G 71E A R~G-BRICK *-WzS**E 2' CLOCKWISE ExVLoost ISoUE-l ic 5,\u00fdOWING IRICK A04D TILE POtO~S.\nPLAN SNOWINGc BRICKS TrwiSTED ,MRO' 2- AL-ERNATE LAIYERS TWISFEZ CLO-CKWISil ANT I CLOCKWISE TILES 2C5MAIN sctUAAE TO AXES 7 NR-oUG)CUl TN.E S;tRuC.ZLIE.\nr.i C.- CLEP WORKS V00E=\u00fdA-,C S7;lUCTUP.E GRAPHITE TECHNOLOGY *sNol-D31910 Nolsnt:-,%\u00fd Nouv, Lh2s3 bci3.d VNM\u00fd3b-s .0g.r.a.\nt- ainjoa-I A. DOrlOKHDal alIHcIVUD Lecture 4 TR AWSFY NYDD (C) RADIALLY KEYED TILE LATINA (b) OCTAGONS.QUARE RADIALLY KEYED BRtCKS TOKA!\nMURA HEXAGON RADIALLY KEYED BRICKS Figure 4.\nEXAMPLES OF IN MAGN OX RADIAL KEYING USED REACTOR CORES -4 GRAPHITE TECHNOLOGY LAYER (c) --m r---; ..\n: .. t : , .\n, : .. ;--\" : : .\nRADIALLY K YED TILE RA DIA LLY HEXAGON IN MAGNOX ---- --------- ---------------/11 )\"' 0-4 C) H C-) C -I C) 4..\n'at0 ((A 4An ps A CD ino Or X1 fit 13 0 n r 3. -I 5 2 0 11 ID A) A 7% a 3. -U -I I -o il S0 -- 0 X < \u20ac III Ilt) Ifi o in --III __ HIl oil n. z III -I -v A)-o -I 3.I A) A) 3.iJ 3 Un z G1 0 2 P Lecture 4 Restraint ta.nk Fuel. -4\" brick .\n.:* . ?ig.6 . Part plan and sec:ion of care \" GR.APHITE TECILNOLOGY Lecture 4 FUEL FUEL IC HANNE u I / / CONTROL CHANNEL coo L! N C GAS CHANNEL Figure 7. Core Bricks and Keys GRAPHITE TECHNOLOGY SEAL RI NG BEY SUPDONT DLo EL K L.LE-; (\u0002b> Cr Lecture 4 F -x I -___~~~ ~~ r_ rIZTIlZZLZE U~lZ. LI~Z~LZ YI -TO BRACE f PE InNE SLEEVE /:tIlR E W BRAC E RETAIPING RIIG -OUTER SLEEVE -FOUE Pi\" Ass~eL4.y r x N. IN. N. N.. \\ \u0002 \\U\\I I -\u0002---\u0002I I -LOWIR INNER SLtEVi GAID GUIIDE lulf Figure 8. ARlRANGEMENT OF HINKU~.Y -POIW T A GA FUEL ELEMENT.\nSECTION X-1 .. .\n.\n.\n.\n.\n.. . . . .\nI/ Lecture 4 \u0002r*44.\nzm. fc. Figure 9. C~zm-e.rai :Gr? Stage 2, Iue* ~e\u00fd,eme zerbrm, ance and saeie7 zarcea ZY GRAPHITE TECHNOLOG~iY Lecture 1 - Workshop Fxercies Case 1 Calculate the radial variation of damage function, Calder equivalent rating, dose and tempera ture for the following example of a single rod of natural uranium in a single fuel channel sur rounded by graphite using the following parameters.\nFuel burnup = 45,000 MfWdft Operating life at constant fuel rating Fuel rod cross sectional area Channel bQre diameter Graphite density Graphite operating temperature Activation energy = 30 years at 85% load factor = 1.0 in2 = 4.0 in = 1.70 g/cmn 3 = 400 'C = 1.2eV Case 2 Calculate the DIDO nickel dose, dose rate and DIDO equivalent temperature at the channel bore position in Case 1. Case 3 Calculate the damage function at the bore of the central fuel channel of a 5 by 5 array of 4.0 inch fuel channels located on a 10 inch square lattice pitch, assuming all 25 channels have the same parameters as those in Case 1. The position of interest should be assumed to lie on a Line joining the central fuel channel to one of it's closest neighbours. What is the percentage of the total damage contributed by the five channels closest to the point of interest? ........... .\n...... .......... GRAPHITE TECHNOLOGY Lecture 1 - Workshop Cnse 4 If the fuel bumup on channel rows A,B,C,D and E are as follows, calculate the DIDO nickel dose at the point of interest A Burnup, MvVd/t 46,000 B 44,000 C 42,000 D 40,000 E 38,000 _.\nCase 5 How would you expect graphite weight loss and dimensional changes to affect calculated dam age function values? ..................................... .......................... .\n.\n~.... .. ..... . GRAPHITE TECHNOLOGY 4 cm = 10 inch A B C D E 3 _ _ o GRAPHITE TECHNOLOGY lccture 2 - NVorkshop ix rci \u00fdzcs Case 1 \"Using the parameters for Case 1 of Lecture 1 Workshop, estimate the percentage radial varia tion in dimensional change at the end of operating life in the parallel to extrusion direction of PGA graphite assuming a constant irradiation temperature of 200 'C. Limit the calculation to a 5 inch thickness of graphite. Case 2 A PGA graphite component having unirradiated CL (0-120) thermal expansion coefficient values of 3.0 x 10-6 OC(' perpendicular and 0.8 x 10-6 OC' parallel is irradiated at 300 0C to Calder equivalent dose of 40,000 MWD/T at a constant Calder equivalent rating of 4.38 MW/T. Estimate the post-irradiation CTE values in terms of: a) The mean GTE between 20 \"C and 120 'C, (X(20.120) b) The instantaneous CTE at 3000C c) The mean CTE between 20'C and 300'C, a (2o-300). Case 3 A PGA component is irradiated at 200'C to a Calder equivalent dose of 20,000 MWD/T whilst the reactor is operating continuously at constant power for a total of 25 years. During this time the component suffers 10% weight loss due to radiolytic oxidation. Calculate the resultant thermal conductivity at a temperature of 2000C assumning an unirradiated value of 100 W/m K at 300C. Case 4 What are the end of life values of total stored energy, E and the rate of release of stored energy, (dE/dT)max, for the component in Case 3? If (dE/dT)max occurs at 400'C, what is the fraction of specific heat represented by stored energy release at this temperature? Case 5 Produce the relationship which describes the dimensional change versus dose of Gilsocarbon graphite in terms of values read from the curves in Figure 4, if the graphite total opv is 0.11 cm 3jcm3 and the coolant gas composition is 1% CO and 300 vpm CH4. ..... .. ... ....... .. ........ .\n... R A I T E T E ... .GRAPHITE TECHNOLOGY Lecture 2 - Worlkshop Case 6 A Heysham U / Torness graphite component is irradiated at a temperature of 450'C to a total dose of 1.8 x 1022 n/cm2 whilst operating at constant power for 25 years. During this time the component suffers 20% radiolytic weight loss. Estimate the resultant thermal conductivity at 4500C assuming an irradiated value of 128 W/m0C at 30 0C. Case 7 A Gilsocarbon graphite component having an unirradiated coefficient of of thermal expansion, X(2\u2022-120), value of 4.5 x 10-6 'C-I is irradiated to an equivalent DIDO nickel dose of I x 1072 n/cm2 at a constant dose rate of 2 x 1013 n/cm2. s and at a temperature of 450'C. During this time the component suffers a tensile strain of 1%. Calculate the post irradiation values of ther mal expansion coefficient in terms of: a) The instantaneous CTE at 4500C b) The mean CTE between 20 and 450'C. ...\n* ... .\n.\n...........\nM GRAPHITE TECHNOLOGY Lecture 3 - Workshop Exercises Case I A graphite brick in a Magnox reactor operating at a temperature of 200\"C and a coolant gas pres sure of 120 psia is radiolytically oxidised in coolant at a G, value of 1.2. The virgin values of graphite density and effective open pore volume are 1.7g/cm 3and 0.22 cm3/cm 3, respectively. If the dose rate to graphite remains constant at 0. 1W/g, calculate the cumulative weight loss at 5 year intervals up to 25 years continuous operation. Check your answers using Figure 8. Case 2 If the dose rate increases with mean weight loss (i.e. inversely with graphite density change), esti mate new values for the cumulative weight loss every 5 years of operation assuming the following values for mean weight loss during each period.\nPeriod, years Mean Weight loss,% 0-5 1 5-10 3 10-15 5 15-20 7.5 20-25 10.5 Assume that got for any period 0 to t years is increased by the factor 1/(1-x), x is the time averaged fractional weight loss between 0 and t years. Thus for the period 0 to 10 years, x=(0.01+0.0 3)/2 =0.02. Hence the factor 1I(1-x)=1/0.98, and so on. Case 3 If the dose rate in Case I is assumed to increase in accordance with operation at constant power, calculate the cumulative weight loss after 25 years using equation 13. Check your answers using \"Figure 9.\nG T.E H..OO...Y .- -- ~GRAPHI~TE TECH:NOLOGY Lecture 3 - Workshop Exercise Case 4 The virgoin values of Young's Modulus and static strength for the PGA graphite component are: E. perpendicular E0 parallel a0 perpendicular ao parallel = 5.4 GPa = 11.7 GPa = 12.0 MPa = 19.0 MPa Using the weight loss results from Case 2 above, and the following damage dose values, draw up a table of giving the resultant values of Young's Modulus and strength allowing for the combined effects of radiolytic weight loss and fast neutron irradiation. Assume a DIDO equivalent tempera ture of 250\"C.\nTime, years EDND x10 2 1 r/cmn 2 5 0.435 10 0.871 15 1.306 20 1.742 25 2.177 \" .\n:. .. .\n.\n.~. ... :. .G:H.I.... ....................... TECHN.::O L \u2022\u2022 \u2022 ..............\nGRAPHITE TECHNOLOGY Lecture I - Workshop Answers Case I F 45, 000 = 4.83 MnW/t Fuel rating, P -30 x 0.85 x 365 Calder equivalent rating, Pe = 0.689 \u2022 A \u2022 df\" P = 0.689 x 1.0 x 4.83 x df = 3.33df MW/t Calder equivalent dose, D, = Pe x 30 x 0.85 x 365 = 9307 \u2022 5 P. Mwd/t + k 1 0 e loIe T , T. E'3.12 + 8.617x10- 1g 0 Pe -673 1.23.12 Re .....\n, Pe De R MW/t NfWd/t 1.0 0.50 1.665 15,497 1.2 0.40 1.332 12,398 1.15 0.288 0.959 8,926 1.025 0.205 0.683 6,570 0.74 0.106 0.353 3,286 0.29 0.024 0.080 745 0.12 0.007 0.020 186 0.04 0.002 0.007 65 0.02 0.01 0.03 28 Te Te K C 694 421 702 429 714 441 726 453 752 479 818 545 890 617 -954 681 1013 740 ;-:::.:::GRAPHITE TECHN::-OLOGY ins 0 2 3 5 10 15 20 25 30 Radius Rims 2 3 4 5 7 12 17 22 27 32 R, RgI.6 ins 0 1.06 2.13 3.19 5.31 10.63 15.94 21.25 26.56 31.58 Lecture 1 - Workshop Answers Case I Fuel rating., P 45,000 = 4.83 MNWIt 30 x 0.85 x 365 Calder equivalent rating, Pe = 0.689 \u2022 A \u2022 df\" P = 0.689 x 1.0 x 4.83 x df = 3.33df MW/t Calder equivalent dose, De = P. x 30 x 0.85 x 365 = 3 5 Pe Mwd/t 1 T\" TjE -3.12) I + 8.617xlO'5 P 1 673 1.2 3-6 Radius Rg Rg 1 .\n6 d Rg R e De Te Rins ins ins NfW/t MWd/t *K *C 2 0 0 1.0 0.50 1.665 15,497 694 421 3 1 1.06 1.2 0.40 1.332 12,398 702 429 4 2 2.13 1.15 0.288 0.959 8,926 714 441 5 3 3.19 1.025 0.205 0.683 6,570 726 453 7 5 5.31 0.74 0.106 0.353 3,286 752 479 12 10 10.63 0.29 0.024 0.080 745 818 545 17 15 15.94 0.12 0.007 0.020 186 890 617 22 20 21.25 0.04 0.002 0.007 65 \u2022 954 681 27 25 26.56 0.02 0.01 0.03 28 1013 740 32 30 31.58 .\n.... ............ .~. ..... . ... GRAPHITE TECHNNOLOGY Lecture I - Workshop Answers case I 45, 000 4.83 MYW/t Fuel rating, P 30 x 0.85 x 365 Calder equivalent rating, Pe = 0.689 \u2022 A \u2022 df\" P = 0.689 x 1.0 x 4.83 x df = 3.33df M'V/t Calder equivalent dose, De = Pe x 30 x 0.85 x 365 = 9307 -5 Pe Mwd/t 1 k1 klog,( P Te TA E -T .2 -14 8.617x10-5 loge (3.12 =67\"-3 1.2 3.1 4 bR, R 1 .\n6 \u20ac R- R ins 0 1.0 0.50 1.06 1.2 0.40 2.13 1.15 0.288 3.19 1.025 0.205 5.31 0.74 0.106 10.63 0.29 0.024 15.94 0.12 0.007 21.25 0.04 0.002 26.56 0.02 0.01 31.58 -pe De MNW/t NMWd/t 1.665 15,497 1.332 12,398 0.959 8,926 0.683 6,570 0.353 3,286 0.080 745 0.020 186 0.007 65 \u2022 0.03 28 0 0 K C 694 421 702 429 714 441 726 453 752 479 818 545 890 617 77954 681 1013 740 ~.~.'... .......\n~.......,... .. .\n... .\n. GRAPHITE TECHN-OLOGY Radius Rins 2 3 4 5 7 12 17 22 27 32 R, ins 0 1 2 3 5 10 15 20 30 Lecture 1 - WVorkshop Answers Case 2 EDND = 1.0887 x 1017 x De = 1.0887 x 1017 x 15,497 = 1.687 x 1021 n/cm 2 1.687 x 1021 Equivalent DIDO dose rate, \u20ac = 30 x 0.85 x 365 x 24 x 3600 - 0.21 x 1013 n/cm2 \u2022 S 8 Ti E 4x10t3 -~+ 8.617x10-5 In(0.1) 673 + 1.2 4\u2022\" e = 785 K = 512 0C GRAPHITE TECHNOLOGY Lecture 1 - \\orkslhop Answers Case 3 M.8-t Total Percentage = 9.7 + 6.0 + 59.0 + 9.1 = 83.8 % GRAPHITE TECHNOLOGY Lecture 1 - Workshop An \u00fd- .e rs Case 4 Channel B B df(from dfxB 42, 000 case 3) 42, 000 Al 46,000 1.095 0.003 0.003 B1 44,000 1.048 0.005 0.005 C1 42,000 1.000 0.013 0.013 Dl 40,000 0.952 0.005 0.005 El 38,000 0.905 0.003 0.003 A2 46,000 1.095 0.006 0.007 B2 44,000 1.048 0.036 0.038 C2 42,000 1.000 0.082 0.082 D2 40,000 0.952 0.041 0.039 E2 38,000 0.905 0.006 0.005 A3 46,000 1.095 0.009 0.010 B3 3 44,000 1.048 0.051 0.053 C3 42,000 1.000 0.500 0.500 D3 40,000 0.952 0.077 0.073 E3 38,000 0.905 0.011 0.010 EDND = 3.9942 x 10'5 x A x dfx B x Pu = 3.9942 x 1015 x 1.0 x 0.846 x 42,000 x 18.78 = 2.67 x 1021 n/cm2 GRAPHITE TECHNOLOGY lccturc I - Worklhop Case 5 Graphite weight loss will cause damage functions to increase. Dimensional changes in the moderator brick transverse direction will have little or no effect on damage functions. However, axial brick shrinkage will densify the core and cause calculated damage function values to be reduced. GR\u2022kPHITE TECHNOLOGX Lecture 2 - Workshop Answers Case 1 End of life bore dose, EDND = 1.687 x 1021 n/cm2-(See Lecture 1 workshop answers for cases 1 and 2). DIDO equivalent rating at bore = 0.21 x 1013 n/cm3. s At 2000C, Ti = 473 K 1 _ 1 kl1 (T 4\u00d7T\u2022 -T E ( 4x,10'13 I +78.617 x- 10-5I 2\"-' =526 K i.e. 2530 C -473 1.2~ 4 Draw up table using dose I dose rate ratios from Case 1. Answers of Lecture workshop SEDND T e 0Shrinkage I 1 n/cm 2 , s I n/cm2 1 K \u00b0C % 0 0.21 x 10' 3 1.687 x 1021 526 253 1.45 1 0.168 x 1013 1.350 x 102' 530 257 0.85 2 0.121 x 1013 0.972 x 1021 537 264 0.48 3 ! 0.086x1013 0.672 x 1021 545 272 0.26 5 _ 0.0445 x 1013 0.358 x 102i 559 286 0.10 Note: Shrinkages read by interpolation between 250 and 3000 C DIDO curves in Figure 2. GRA.. ............. .\nG ............ GRAPHITE TECHNNOLOGY Lecture 2 - Workshop AMnswers Case 2 EDND = 40, 000 MWD/t x 1.0887x1017 n/em2 IBVD/t = 4.355 x 1021 n/c2 Irradiation time = Calder equivalent dose. MWDft Calder equivalent rating, MW/t 40. 000 9132.4 days 4.38 Dose rate, 4.355 x 1021 0.552 x 1013 n/cm2 s s 3t 9132.4 x 24 x 3600 1 ,_ 1+ 8.617 x 10-I ( 0- 552) 37 573 1.2 4 0= 624K ie 351 0C From figures 2.6, 2.7, 2.8a and 2.8b:-a (20 - 120) a (20 - 120) =i (300) ci (300) cL(20 -300) cL(20 -300) perpendicular parallel perpendicular parallel perpendicular parallel = 5.09 x 10-6 = 1.53 x 10-6 = (5.09 + 0.83) 10 6 = 5.92 x 10 = (1.53 + 0.83) 10-= 2.36 x 10-= 5.43 x 10\"6 = 1.85 x 10 .......\nE .Ec .o.o Lecture 2 - Workshop Answers Case 3 EDND Dose rate,, = 20, 000 x 1.0887 x 10 7= 2.177 x 10 2n/cm2 2.177 x 1021 25 x 365 x 24 x 3600 0.276x10 n/cmn .s I= _L1, 8.617 x 10' rn ( )2 e 473 -1.2 4 6= 520Kie2470 C K (30) K FK\u00b0 +f\" 6 S,[- ox K(T) Ko(30)K T) JkLK 1 1 +53.5x0.84] 1.0xe(3.1-0.1) K(200) 10o00.78 K(200) 0 1.587W/imK K(200)-63.0 .. -.\n.\n.\n.\n...-. .- .GRAPHI E T.:.E.\" .\n:...::::.Y\" .:..::.: .\n.~GRAPHITE TECHLNOLOGY Lecture 2 - Workshop Answers Case 3 EDND = 20, 000 x 1.0887 x 10 17= 2.177 x 10 2n/c,2 Dose at 2.17 x - 0.276 x 1013/cm2. s 25 x 365 x 24 x 3600 1 1+ 8.617 x 10-5 In 0.27 6473- 1.2 --4 6= 520Kie2470 C x~K 3 Koo o K(T) F____01K,(7 1 _ 1 +053.5x0.84]1.0 xe 3 x01) K(200) 100 o.78 K(200)= 10= 1.587W/ImK 63.0 GRAPHITE TECHNOLOGY Lecture 2 - Workshop AnMwers Case 4 If f- 53.5 (from case 3 then: E= 31.1f= 31.1 x 53.5= 1664 J/g LE d. E. _ 1664 0.87= 0.87 J/g.K d'T a, 1910 1910 Specific heat at 4000C = 1.43 J/g.K 0.87 o-6 ...\nRatio .-. = 0:( .-. Rtio=1.43 ) .............. .... .. GRAPHITE TECHN'OLOGY Lecture 2 - Workshop Answers Case 4 If f= 53.5 (from case 3 then: E= 31.If= 31.1 x 53.5= 1664 J/g LdEI E 1664_ 0.8 7 = 0.87 J/g.K \"dT max 1910 1910 Specific heat at 4000 C = 1.43 J/g.K .Ratio= 0.87 1.43 GRAPHITE TECHNOLOGY Lecture 2 - Workshop _Answer5 Case 4 If f- 53.5 (from case 3 then: E= 31.1f= 31.1 x 53.5= 1664 Jig dE) 5 1664087 0.87 J/g.K 1~ dTma 1910 1910 Specific heat at 4000 C = 1.43 J/g.K .Ratio 0.87 0:61 \"1.43 GRAPHITE TEC.LNOLOG.\nLecture 2 - Workshop Case 5 Initial total open pore volume, VTOT= 0. 11ctn\"/c3 Initial reactive pore volume, VRp,= 0.15 x 10-8 x RR Now from table 2.1, RR = 0.124 x 10-8 ..VRPV 0.15 x 0.124= 0.0186 The ratio VRpV/ VTOT= 0.0186/0.11= 0.169 Dimensional change % =0.169 curve A + 0.831 curve B GRAPHITE TECHNOLOGY Lecture 2 - Workshop Answers Case 5 Initial total open pore volume, VTOT= 0. 11 cm3 /cm3 Initial reactive pore volume, VRPv= 0.15 x 10-8 x RR Now from table 2.1, RR = 0.124 x 10-8 ...VRPV= 0.15 x 0.124= 0.0186 Theratio VRpV/VToT= 0.0186/0.11= 0.169 Dimensional change% =0.169 curve A + 0.831 curve B GRAPHITE TECELNOLOGY Lecture 2 - Workshop Answers Case 5 Initial total open pore volume, VT -= 0. 11 cm3 / C in3 Initial reactive pore volume, VRpv- 0.15 x 10-8 x RR Now from table 2.1, RR = 0.124 x 10-8 .VRPV- 0.15 x 0.124= 0.0186 The ratio VRpVIV o=- 0.0186/0.11= 0.169 Dimensional change% =0.169 curve A + 0.831 curve B : ::::: : ====== ==== == ... ... : .. .. .. ..\n.. .:.: .*...\u2022.::\u2022....:::::::::... ..\n~ ... :...* .\n\u2022 .*. ... * .. ..... .. .\u2022.\u2022...:...:\u2022...**. ... .*. ..\n*. .\n5 :5.5 ..\n: :\u2022 .\n`.;. \u2022 .. ... ...\n: GRAPHITE TECHNOLOGY Lecture 2 - Workshop An-swers Case 6 EDND = 1.8xl0-- n/cmr 2 Dose rate = '8x102 = 2.283x101 3 n/cm 2.s 25 x 365 x 24 x 3600 (1.2eV) 1 T 1 -8.617x10\"5 In2.283 ( 23 1.2 4 e 745K ie 472\u00b0C 1_ 1 +8.617x10 2.283 (V 8 723 3 4 e = 732K ie 459\u00b0C 1 FKo (30) K,(30) LK,(T +f -T] Sk 0~ From Figs 2.14, 2.15, 2.17 and 2.18: Ko(30) = 128* ; f = 4.7 K, (30) 1 K,, (450) 0.825 6 T = 0.68 at 450\u00b0C; (o K ox 2.7 x 0.2 =e = 1.72, Sk = 1.1 1+ 4.7 x 0.68 \u20221.1 x 1.72 K(450) = 15.3 W/rK' * Given in quesdon. GRAPHITE TECHLNOLOGY I K(T) ( .-Lecture 2 - Workshop Answers Case 6 EDND = 1.8xi022 n/cm 2 Dose rate 1.8xi0 = 2.283xi 013 n/cm2 .s 25 x 365 x 24 x 3600 (1.2eV) 1 , I + 8.617x10- .\nIn2.283 o 723 1.2 4 0 = 745K ie 472'C (3eV) 8.617x10-5 In 2.283 6 723 3 4 E) = 732K ie 459\u00b0C 1 1 [Ko (30) K(7) K, o(30 KO) (7 +f' ST]. Sk\"I K9 From Figs 2.14, 2.15, 2.17 and 2.18: Ko(30) = 128*; f = 4.7; 1 1l 28 08-2-5 BT = 0.68 at 450\u00b0C; Sk = 1.1 + 4.7 x 0.68] ' 1.1 x 1.72 K(450) = 15.3 W/mK * Given in question.\nGRAPHITE TECHNOLOGY Ko(30) _ Ko (450) 0.825 2.7 x 0.2 =e = 1.72, 1 K (450) Lecture 2 - Workshop Answers Case 7 EDND = i x i0 n/cm2, Dose rate = 2x1013 n/cm2s 1 = 1 +8.617xi0 -(.1n 2) e 723 3.0 o = 734K ie 4610C From Figs 2.6, 2.7, 2.9a and 2.9b c (20 - 120) = 4.5xr10-6 Correction for ao = (4.5 -4.35)xlO-6 = 0.15x10-6 Correction for strain = -1.35xl0- 6 (Fig 2.9b) =370-6 (i .a Irradiated (20- 120) 10 (Fig 2.9a) Corrected a (2 0 -120) (3.7 + 0.15 - 1.35) xlO 6 = 2.5x10\"6 From Fig 2.6: Irradiatet axi( 4 5 0) = AOxi + Bi = I x 2.5x10-6 + 1.25x10-6 = 3.75x10-6 W/m.K From Fig 2.7: Irradiated a( 2 0 ., 4 50) = 3.2x10 6 W/m.K S...... .I.. GRAPHITE TECHNOLOGY Lecture 3 - Workshop Case I 145 -Ee\" G,-\" D'p go= T 0.22 e 1.7 3 3 Cm Cm = 0.1294 cm3 /g Cm g 145 x 0.1294 ; 1.2 x 0.1 x 120 = 0 %peryea \"\"Co= 473 0.7%erya Time gt C, ( 28.205[x got~i1 yrs % % 5 2.85 3.00 10 5.70 15 8.55 20 11.40 25 14.25 6.32 9.99 14.05 18.54 Ct A lexpkA 1l A 1007re _ 100 x 0.22 28.205 ( -ne) 0.78 .-. C= 2S.205[exp2 -10 .............. ..............\nGRAPHITE TECHNOLOGY Lecture 3 - Workshop .An swers Case I 145 \u2022-ee\" G-e'D ' p go= T 0.22 S = e 1.7 \"\"O0 3 3 Lm - = 0.1294 cm3/g 30 Cm -145 x 0.1294 x 1.2 x 0.1 x 120 = 0.57 % per year 473 a at Time cot C~ -=~..~eP 8 0 - I T~o =or 1 l.=07 218205=o yrs % % 5 2.85 3.00 10 5.70 6.32 15 8.55 9.99 20 11.40 14.05 25 14.25 18.54 Ct A[ exp A - 1] A 100te 100 x 0.22 28.205 (1 -ie) 0.78 ... Ct -28.205[expgot 0 5 1] GRAPH.ITE TEC5NOLOG Lecture 3 - Workshop Answr Time t Mean got New got New Ct Weight Loss yrs to time t, % const. dose % % 5 1 2.85 2.879 3.03 10 2 5.70 5.816 6.46 15 3 8.55 8.814 10.34 20 4.12 11.40 11.89 14.79 25 5.4 14.25 15.06 19.90 New Ct = 2 8. 2 0 5 [exp g 2 0 1] .GRAP..TE.TECHY\u2022 .\n.\u2022\u2022....OG GRAPHITE TECHNOLOGY Lecture 3 - Workshop Answers Case 2 Time t Mean got New got New Ct Weight Loss yrs to Wime t, % const. dose % % 5 1 2.85 2.879 3.03 10 2 5.70 5.816 6.46 15 3 8.55 8.814 10.34 20 4.12 11.40 11.89 14.79 25 5.4 14.25 15.06 19.90 New C,: 2 8 .20 5 [expg I] GRAPHITE TECHNOLOGY Lecture 3 - Workshop Answems Case 2 Time t Mean got New got New Ct Weight Loss yrs to time t, % const. dose % % 5 1 2.85. 2.879 3.03 10 2 5.70 5.816 6.46 15 3 8.55 8.814 10.34 20 4.12 25 5.4 11.40 11.89 14.25 15.06 14.79 19.90 NewC t = 28.205 [exp .-g 5 o 1 ... .. -...... GRAPHITE TECHNOLOGY Lecture 3 - Workshop Answers Case 2 Time t Mean got New got New C, Weight Loss yrs to time t, % const, dose % % 5 1 2.85 2.879 3.03 10 2 5.70 5.816 6.46 15 3 8.55 8.814 10.34 20 4.12 11.40 11.89 14.79 25 5.4 14.25 15.06 19.90 New C = 28.205[exp2gt 0 5 t ] :.\u2022 : \u2022 .. :::.\u2022\u2022 .. \u2022 :;::.:..-...: .+:\u2022:.:.::::::::::::::::::::::::::::::::::::: :::::::::::::::::::::::::::::::::::::::::::::::::::::::::=============_==_====____________________ ==: ,-::\u2022 .\n::\u2022; g\u2022:::::.-:::::-,.\u2022:...:::,..:\u2022:. ..\nGRAPHITE TECHNOLOG Y F !\nLecture 3 - Workshop Answers Case 3 For operation at constant power (i.e. increasing dose rate) the following equation must be satisfied A 2nl~ C _ AC l n e+ -! 1- t = g o t Nwfo - 1. A 10025 hn Now ifA =28.205, e ~0.22, got = 14.25, then: 28.2052 In( 1 + 100 x 0.22 128.205 S.Q t = 14.25 100 36.16 x In 1 + 28.\"5 - 0.282C, = 14.25 TryC= 20 36.16 x In (1.7091) - 5.64 = 14.25 19.38 - 5.64 = 13.74 * 14.25 Try Ct = 21 36.16 x In (1.7445) - 5.92 = 14.25 20.12 - 5.92 = 14.20 = 14.25 From figure 9, Ct = 21% GRAPHITE TECHNOLOGY .................\n_______________________________________________________________ P Lecture 3 - Workshop Answers Case 3 For operation at constant power (i.e. increasing dose rate) the following equation must be satisfied A1 2 ln 1 4CE_ A C = got Now if A = 28.205, ne -0.22, got = 14.25, then: 28.2052 .\n1+ C. 14.25 100 x 0.22 1 28.205, 100 36.16 x In ( + 28.205 - 0.282C, = 14.25 Try Ct = 20 36.16 x in (1.7091) - 5.64 = 14.25 19.38 - 5.64 = 13.74 * 14.25 Try C= 21 36.16 x In (1.7445) - 5.92 = 14.25 20.12 - 5.92 = 14.20 - 14.25 From figure 9, Ct = 21% GRAP..... TY......................-.EC:NOLOGY GRAPHITE TECHNOLOGY Lecture 3 - Workshop Case 4 Perpendicular Young's Modulus, E Time, t EDND EWt Loss E yrS x 102' n/cm2 EO) i % ox GPa 0 0 1 0 5.4 I 5 5 0.435 2.19 3.03 0.8646 1.8935 10.22 10 0.871 2.07 6.46 0.7334 1.518 8.20 15 1.306 2.07 10.34 0.6088 1.260 6.81 20 1.742 2.07 14.79 0.4917 1.018 5.50 25 2.177 2.07 19.90 0.3847 0.7963 4.30 E)from Figure 2.23, (E e 48 Perpendicular Strength, a T i me ,t E D ND W t L o ss U yrs xl& 1 0'n/cm- %O L 0 GPa 0 0 1 0 1 1 12.00 5 0.435 1.48 3.03 0.8542 1.264 15.17 10 0.871 1.44 6.46 0.7147 1.029 12.35 15 1.306 1.44 10.34 0.5841 0.841 10.09 20 1.742 1.44 14.79 0.4634 0.667 8.01 25 ] 2.177 1.44 19.90 0.3553 0.512 6.14 Er = E from above, a.\u2022 E ( C, -5.x .. . .\n. .\n. .. .. .. . .\n.\n.. .... * .'....- .. ..\n*... .\n.\n.:..:-::...\u2022{::\u2022: \u2022 ::\u2022\u2022:\u2022i \u00a2 GRAPHITE TECHNOLOGY Lecture 3 - Workshop Answers C\u2022ase 4 (Continued) Parallel Young's Modulus, E Time, t EDND WtLoss yrs x 102 ' n/cm2 E %IO ox 0 0 1 0 1 5 0.435 2.12 3.03 0.8646 10 0.871 1.87 6.46 0.7334 15 1.306 1.73 10.34 0.6088 20 1. 742 1.61 14.79 0.4917 25 2.177 1.-61 19.90 0.3847 L --------L -Parallel Strength, f Tune, t EDND Wt Loss yrs x 102 1 ncm2 %GPa 0 0 1 0 1 1 19.00 5 0.435 1.456 3.03 0.8542 1.244 23.63 10 0.871 1.368 6.46 0.17 0.9771 18.57 15 1.306 1.315 10.34 0.5841 0.7683 14.60 20 1.742 1.269 14.79 0.4634 0.5880 11.17 25 2.177 1.269 19.90 0.3553 0.4508 8.57 (- frtom above, ) = 52 ....... ...\n..\n\" . \"GRAPHITE TECHNOLOGY E GPa 1 11.70 1.8313 21.45 1.372 16.05 1.053 12.32 0.7916 9.26 E0\u00fd. 6j1 94 7.25 Lecture 3 - Workshop Par-lwelYu Case t (Continue-d parallel Young,*s Modulus, E E i\u00fd fromxFigure2.22, E e 01i(f -Parallel Strength. u ( I- = from above,1Ei = e. Gi TH ..... ...\n::\u2022 \u2022\u2022\u2022:\u2022:::::\u2022?.\u2022 ..\n??::::\"\u2022''...\n\u2022;\u2022:\u2022 \u2022.....RAPHtITE TECHINOLOGY"
+        },
+        {
+            "title": "Nuclear Graphite Research Needs in the 21st Century",
+            "url": "http://large.stanford.edu/courses/2017/ph241/sarkisian2/",
+            "content": "Nuclear power reactors that use graphite moderators, such as those based on the RBMK design, encase the nuclear fuel rods in chambers with graphite walls.",
+            "raw_content": "# Nuclear Graphite Research Needs in the 21st Century\n\n## Dylan Sarkisian May 15, 2017\n\n### Submitted as coursework for [PH241](..), Stanford University, Winter 2017\n\n## Introduction\n\n|  |\n| **Fig. 1:** Schematic of a Reaktor Bolshoy Moshchnosti Kanalnyy (\u0420\u0435\u0430\u043a\u0442\u043e\u0440 \u0411\u043e\u043b\u044c\u0448\u043e\u0439 \u041c\u043e\u0449\u043d\u043e\u0441\u0442\u0438 \u041a\u0430\u043d\u0430\u043b\u044c\u043d\u044b\u0439), or \"High Power Channel-type Reactor,\" a class of graphite-moderated nuclear reactor designed in the Soviet Union. The RBMK is the oldest commercial reactor design still in operation. (Source: [Wikimedia Commons](https://commons.wikimedia.org/wiki/File:RBMK_reactor_schematic.svg)) |\n\n![](images/f1.png)\n\nThe neutron moderator is one of the core elements of\na nuclear power reactor, responsible for slowing the neutrons that are\nejected from the nuclear fuel rods during fission. When the nuclear\nfuel, such as Uranium-235, undergoes fission, it ejects fast-moving\nneutrons. However, a nucleus must be hit by a slow-moving free neutron,\nor thermal neutron, in order to be incited to fission. Thus in order to\ncreate a nuclear chain reaction, whereby the neutrons ejected from one\nnucleus incite fusion in other nuclei, the fast-moving neutrons ejected\nduring fusion must be slowed down by a moderator.\n\nAs shown in Fig. 1, nuclear power reactors that use\ngraphite moderators, such as those based on the RBMK design, encase the\nnuclear fuel rods in chambers with graphite walls so that the\nfast-moving neutrons emitted from one rod are slowed by the graphite\nmoderator before reaching other rods, allowing a chain reaction to\nspread throughout all the rods. The heat generated from fission then is\nused to heat water either directly (such as in the RBMK design) or\nindirectly, by heating gas that travels elsewhere to heat water.\n\n## Current Limitations of Graphite Moderators\n\nThree different types of materials are commonly used\nas moderators in nuclear reactors today today: light (regular) water,\nheavy water (deuterium oxide), and solid graphite. Only around 20% of\ntoday's nuclear reactors use graphite as a moderator. [1] The popularity\nof graphite as a moderator has declined significantly in the late 20th\nand 21st centuries, so it is predicted that the percentage of\ngraphite-moderated nuclear reactors will further decline in the next\nfifty years. [2] There are several reasons for this decline in\npopularity. One is residual public fear of graphite-moderated nuclear\nreactors due to the fact that the Chernobyl nuclear explosion occurred\nin an RBMK. [3] This fear is grounded in a key distinction between\ngraphite-moderated reactors and light-water-moderated reactors. If a\nwater-moderated reactor has a loss-of-coolant event, the reactor stops\nfunctioning because the water moderator evaporates away, thus ceasing\nthe nuclear chain reaction. In graphite-moderated reactors, however,\nthe moderator has an extremely high heat of sublimation and thus it\nremains in place through loss-of-coolant events, allowing the nuclear\nreaction to continue in potentially catastrophic circumstances. For\ninstance, the Three- Mile Island and Fukushima nuclear disasters both\noccurred in light water-moderated nuclear reactors, so their moderators\nevaporated quickly after their coolant systems failed. Conversely, the\nChernobyl nuclear disaster occurred in a RBMK graphite-moderated nuclear\nreactor and thus it was able to progress all the way to a full nuclear\nrunaway explosion. [3]\n\nAnother important factor contributing to the decline\nof graphite-moderated reactors is the fact that in nuclear reactors with\ngraphite moderators, the graphite moderator is almost always the\nlife-limiting component of the reactor. [4] Irradiation-induced crack\npropagation (irradiation creep) and stress from extreme heat fluctuation\nlead to significant loss of structural integrity in the graphite\nmoderator over the lifetime of the nuclear reactor. [5] In comparison, a\nwater moderator appears more promising because liquid water is not\nsusceptible to the same structural failures as solid graphite.\n\nAnother unique problem associated with graphite\nmoderators is the fact that spent fuel contains a large quantity of\ngraphite, which makes these reactors have different disposal needs than\nwater-moderated reactors. [6] This requires countries to diversify an\nalready complicated and controversial waste disposal process. In order\nto avoid this complication, many energy providers prefer to use a single\nkind of moderator. [6]\n\nThough graphite moderators currently pose more\nchallenges for plant operators than water moderators, this does not\nrender graphite moderators obsolete. Since a nuclear moderator is\nsupposed to only slow neutrons but not absorb them, a nuclear moderator\nmaterial's effectiveness increases with its ability to scatter neutrons\nand decreases with its ability to absorb neutrons. Mathematically, this\ncan be expressed as the moderating efficiency, which is given by the\nmaterial's neutron scattering cross section divided by its neutron\nabsorption cross section. [7] Graphite's moderating efficiency (1343) is\nless than heavy water's (8154), but two orders of magnitude greater than\nlight water's (74.24), so graphite theoretically has a high potential to\nbe a much better moderator than light water. [7] (The majority of\ntoday's nuclear reactors use light water as a moderator. [1]) Solving\nthe problems of graphite moderator degradation would therefore once\nagain make graphite an appealing moderator material that could make\nreactors much more efficient than light water reactors.\n\n## Research Needs for Nuclear Graphite in the 21st Century\n\nSeveral opportunities exist to improve graphite's\nutility as a moderator in modern nuclear reactors. One is increasing\noxidation resistance by either diffusing interstitials throughout the\ngraphite structure or treating the outside surface of each section\nof graphite in the moderator with oxidation-resistant coatings. [4,5]\nAnother is making the graphite sections more structurally resilient to\nhalt dislocations caused by thermal stress and irradiation creep. [5]\nFinally, additional areas for fruitful research can be found in creating\na simpler system for handling graphite-moderated reactor waste and\ntesting more dependable ways to prevent nuclear meltdowns in\ngraphite-moderated nuclear reactors. [6] Thus nuclear graphite would\nbenefit from both highly focused materials science research and also\nbroad environmental health and safety research.\n\n## Conclusion\n\nRecently, graphite has fallen out of favor as a\nmoderator material for nuclear power reactors. Although graphite is in\ntheory a much better moderator than light water, light water is\ncurrently more appealing because graphite moderators are susceptible to\ndegradation, are capable of causing nuclear runaway explosions in\nloss-of-coolant events, and require more complicated waste management\nthan light water reactors. These challenges present a variety of\nresearch opportunities that could allow modern nuclear reactors to\nutilize graphite's excellent moderating efficiency without suffering the\nsame problems as graphite-moderated nuclear reactors in the past.\n\n\u00a9 Dylan Sarkisian. The author grants permission\nto copy, distribute and display this work in unaltered form, with\nattribution to the author, for noncommercial purposes only. All other\nrights, including commercial rights, are reserved to the author.\n\n## References\n\n[1] G. T. Miller and S. Spoonman,  *Living in the\nEnvironment: Concepts, Connections, and Solutions, 16th Ed.* (Brooks\nCole, 2009).\n\n[2] B. Raj, *et al*, \"Challenges in Materials\nResearch for Sustainable Nuclear Energy,\" MRS Bull. **33**, 327\n(2008).\n\n[3] A. Strupczewski, \"Accident Risks in Nuclear-Power\nPlants,\" Appl. Energy **75**, 79 (2003).\n\n[4] C. Tang and J. Guan, \"Improvement in Oxidation\nResistance of the Nuclear Graphite By Reaction-Coated SiC Coating,\" J.\nNucl. Mater. **224**, 103 (1995).\n\n[5] D. R. De Halas, \"Radiation Effects in Graphite,\"\nin *Nuclear Graphite*, ed. by R. E. Nightengale 7: Theory of\nRadiation Effects in Graphite,\" (Academic Press, 1962).\n\n[6] \"[Technology Roadmap\nUpdate for Generation IV Nuclear Energy Systems](docs/gif-jan14.pdf),\" Gen IV\nInternational Forum, January 2014.\n\n[7] J. C. Bryan, *Introduction to Nuclear\nScience* (CRC Press, 2008), pp. 150-170."
+        },
+        {
+            "title": "Nuclear Power Reactors",
+            "url": "https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors",
+            "content": "In a nuclear power reactor, the energy released is used as heat to make steam to generate electricity. The main design is the pressurized water reactor (PWR) which has water at over 300\u00b0C under pressure in its primary cooling/heat transfer circuit, and generates steam in a secondary circuit. Part of the cooling system of pressurised water reactors (PWR & PHWR) where the high-pressure primary coolant bringing heat from the reactor is used to make steam for the turbine, in a secondary circuit. Magnox reactors were also graphite moderated and CO2 cooled, used natural uranium fuel in metal form, and water as secondary coolant.",
+            "raw_content": "Nuclear Power Reactors - World Nuclear Association\n=============== \n\n[JOIN US](https://world-nuclear.org/our-association/join-us)\n\n[SHOP](https://world-nuclear.org/shop)\n\n[MEMBERS LOGIN](https://world-nuclear.org/member-dashboard)\n\n \n\n[![Image 1: World Nuclear Association](https://world-nuclear.org/images/wna-logo.png)](https://world-nuclear.org/)\n\n*   [NUCLEAR INFORMATION](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#)\n    \n    [Information Library A library of resources related to the Nuclear industry.](https://world-nuclear.org/information-library)\n    \n    [Reactor Database Information on nuclear reactors from around the globe.](https://world-nuclear.org/nuclear-reactor-database/summary)\n    \n    [Essentials Nuclear technology, radiation,and uranium.](https://world-nuclear.org/nuclear-essentials)\n    \n    [Climate Change Nuclear Energy, Climate Change and COP27.](https://world-nuclear.org/climate-change-and-nuclear-energy)\n    \n    [Publications Reports, guides and books from the World Nuclear Association.](https://world-nuclear.org/our-association/publications)\n    \n    [Ukraine Ukraine conflict and nuclear energy.](https://world-nuclear.org/ukraine-information)\n    \n    Subscribe to receive our  \n    enewsletter and updates\n    \n    Please select the mailing  \n    you wish to subscribe to:\n    \n    WNN Daily\n    \n    WNN Weekly\n    \n    Events\n    \n    Press\n    \n*   [News and Media](https://world-nuclear.org/news-and-media)\n*   [ABOUT US](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#)\n    \n    [Our Mission Information on the leadership, organisation and secretariat of the World Nuclear Association.](https://world-nuclear.org/our-association/who-we-are)\n    \n    [Leadership World Nuclear Association Members appoint the Director General and elect a Board of Management.](https://world-nuclear.org/our-association/leadership)\n    \n    [Membership Which companies are members of the World Nuclear Association, and what could you gain from membership.](https://world-nuclear.org/our-association/membership)\n    \n    [At Work At Work is an annual report of World Nuclear Association's activities.](https://world-nuclear.org/our-association/publications/annual-reports/at-work-2024-report)\n    \n    [Vacancies Join us and contribute to the Association\u2019s goals and mission.](https://world-nuclear.org/our-association/vacancies)\n    \n    [World Nuclear University Inspiring, empowering and connecting generations of nuclear leaders](https://www.world-nuclear-university.org/)\n    \n    Subscribe to receive our  \n    enewsletter and updates\n    \n    Please select the mailing  \n    you wish to subscribe to:\n    \n    WNN Daily\n    \n    WNN Weekly\n    \n    Events\n    \n    Press\n    \n*   [Working Groups](https://world-nuclear.org/our-association/about-working-groups)\n*   [EVENTS](https://world-nuclear.org/events)\n*   [SHOP](https://world-nuclear.org/shop)\n*   [MEMBERS LOGIN](https://world-nuclear.org/member-dashboard)\n\n[HOME](https://world-nuclear.org/) / [Information Library](https://world-nuclear.org/information-library) / [nuclear fuel cycle](https://world-nuclear.org/information-library/nuclear-fuel-cycle) / [nuclear-power-reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors) / Nuclear Power Reactors\n\nnuclear fuel cycle\n\nNuclear Power Reactors\n======================\n\nUpdated Friday, 21 March 2025\n\n*   **Nuclear reactors work by using the heat energy released from splitting atoms of certain elements to generate electricity.**\n*   **Most nuclear electricity is generated using just two kinds of reactor which were developed in the 1950s and improved since.**\n*   **The first generation of these reactors have all been retired, and most of those operating are second-generation.**\n*   **New designs are coming forward, both large and small.**\n*   **About 9% of the world's electricity is produced from nuclear energy.**\n\n_This page is about the main conventional types of nuclear reactor. See also pages on\u00a0[Advanced Nuclear Power Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/advanced-nuclear-power-reactors \"Advanced Nuclear Power Reactors\"),\u00a0[Small Nuclear Power Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/small-nuclear-power-reactors \"Small Nuclear Power Reactors\"), [Fast Neutron Reactors](https://world-nuclear.org/information-library/current-and-future-generation/fast-neutron-reactors) and\u00a0[Generation IV Nuclear Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/generation-iv-nuclear-reactors \"Generation IV Nuclear Reactors\")._\n\nHow does a nuclear reactor work?\n--------------------------------\n\nA nuclear reactor produces and controls the release of energy from splitting the atoms of certain elements. In a nuclear power reactor, the energy released is used as heat to make steam to generate electricity. (In a research reactor the main purpose is to utilize the actual neutrons produced in the core. In most naval reactors, steam drives a turbine directly for propulsion.)\n\nThe principles for using nuclear power to produce electricity are the same for most types of reactor. The energy released from continuous fission of the atoms of the fuel is harnessed as heat in either a gas or water, and is used to produce steam. The steam is used to drive the turbines which produce electricity (as in most fossil fuel plants).\n\nThe world's first nuclear reactors 'operated' naturally in a uranium deposit about two billion years ago. These were in rich uranium orebodies and moderated by percolating rainwater. The 17 known at Oklo in west Africa, each less than 100 kW thermal, together consumed about six tonnes of uranium. It is assumed that these were not unique worldwide.\n\nToday, reactors derived from designs originally developed for propelling submarines and large naval ships generate about 85% of the world's nuclear electricity. The main design is the pressurized water reactor (PWR) which has water at over 300\u00b0C under pressure in its primary cooling/heat transfer circuit, and generates steam in a secondary circuit. The less numerous boiling water reactor (BWR) makes steam in the primary circuit above the reactor core, at similar temperatures and pressure. Both types use water as both coolant and moderator, to slow neutrons. Since water normally boils at 100\u00b0C, they have robust steel pressure vessels or tubes to enable the higher operating temperature. (Another type uses heavy water, with deuterium atoms, as moderator. Hence the term \u2018light water\u2019 is used to differentiate.)\n\nComponents of a nuclear reactor\n-------------------------------\n\nThere are several components common to most types of reactor:\n\n**Fuel**  \n[Uranium](https://world-nuclear.org/information-library/nuclear-fuel-cycle/uranium-resources/supply-of-uranium \"Supply of Uranium\") is the basic fuel. Usually pellets of uranium oxide (UO2) are arranged in tubes to form fuel rods. The rods are arranged into fuel assemblies in the reactor core.\\*\u00a0In a 1000 MWe class PWR there might be 51,000 fuel rods with over 18 million pellets.\n\n\\* In a new reactor with new fuel a neutron source is needed to get the reaction going. Usually this is beryllium mixed with polonium, radium or other alpha-emitter. Alpha particles from the decay cause a release of neutrons from the beryllium as it turns to carbon-12. Restarting a reactor with some used fuel may not require this, as there may be enough neutrons to achieve criticality when control rods are removed.\n\n**Moderator**  \nMaterial in the core which slows down the neutrons released from fission so that they cause more fission. It is usually water, but may be heavy water or graphite.\n\n**Control rods or blades**  \nThese are made with neutron-absorbing material such as cadmium, hafnium or boron, and are inserted or withdrawn from the core to control the rate of reaction, or to halt it.\\* In some PWR reactors, special control rods are used to enable the core to sustain a low level of power efficiently.\u00a0(Secondary control systems involve other neutron absorbers, usually boron in the coolant \u2013 its concentration can be adjusted over time as the fuel burns up.) PWR control rods are inserted from the top, BWR cruciform blades from the bottom of the core.\n\n\\* In fission, most of the neutrons are released promptly, but some are delayed. These are crucial in enabling a chain reacting system (or reactor) to be controllable and to be able to be held precisely critical.\n\n**Coolant**  \nA fluid circulating through the core so as to transfer the heat from it. In light water reactors the water moderator functions also as primary coolant. Except in BWRs, there is secondary coolant circuit where the water becomes steam. (See also later section on primary coolant characteristics.)\u00a0A PWR has two to four primary coolant loops with pumps, driven either by steam or electricity \u2013 China\u2019s Hualong One design has three, each driven by a 6.6 MW electric motor, with each pump set weighing 110 tonnes.\n\n**Pressure vessel or pressure tubes**  \nUsually a robust steel vessel containing the reactor core and moderator/coolant, but it may be a series of tubes holding the fuel and conveying the coolant through the surrounding moderator.\n\n**Steam generator**  \nPart of the cooling system of pressurised water reactors (PWR & PHWR) where the high-pressure primary coolant bringing heat from the reactor is used to make steam for the turbine, in a secondary circuit. Essentially a heat exchanger like a motor car radiator.\\* Reactors have up to six 'loops', each with a steam generator.\u00a0Since 1980 over 110 PWR reactors have had their steam generators replaced after 20-30 years service, over half of these in the USA.\n\n\\* These are large heat exchangers for transferring heat from one fluid to another \u2013 here from high-pressure primary circuit in PWR to secondary circuit where water turns to steam. Each structure weighs up to 800 tonnes and contains from 300 to 16,000 tubes about 2 cm diameter for the primary coolant, which is radioactive due to nitrogen-16 (N-16, formed by neutron bombardment of oxygen, with half-life of 7 seconds). The secondary water must flow through the support structures for the tubes. The whole thing needs to be designed so that the tubes don't vibrate and fret, operated so that deposits do not build up to impede the flow, and maintained chemically to avoid corrosion. Tubes which fail and leak are plugged, and surplus capacity is designed to allow for this. Leaks can be detected by monitoring N-16 levels in the steam as it leaves the steam generator.\n\n**Containment**  \nThe structure around the reactor and associated steam generators which is designed to protect it from outside intrusion and to protect those outside from the effects of radiation in case of any serious malfunction inside. It is typically a metre-thick concrete and steel structure.\n\nNewer Russian and some other reactors install core melt localization devices or 'core catchers' under the pressure vessel to catch any melted core material in the event of a major accident.\n\nThere are several different types of reactor as indicated in the following table.\n\nFuelling a nuclear reactor\n--------------------------\n\nMost reactors need to be shut down for refuelling, so that the reactor vessel can be opened up. In this case refuelling is at intervals of 12, 18 or 24 months, when a quarter to a third of the fuel assemblies are replaced with fresh ones. The CANDU and RBMK types have pressure tubes (rather than a pressure vessel enclosing the reactor core) and can be refuelled under load by disconnecting individual pressure tubes. The AGR is also designed for refuelling on-load.\n\nIf graphite or heavy water is used as moderator, it is possible to run a power reactor on natural instead of enriched uranium. Natural uranium has the same elemental composition as when it was mined (0.7% U-235, over 99.2% U-238), enriched uranium has had the proportion of the fissile isotope (U-235) increased by a process called enrichment, commonly to 3.5-5.0%. In this case the moderator can be ordinary water, and such reactors are collectively called light water reactors. Because the light water absorbs neutrons as well as slowing them, it is less efficient as a moderator than heavy water or graphite. Some new small reactor designs require high-assay low-enriched uranium fuel, enriched to near 20% U-235.\n\nDuring operation, some of the U-238 is changed to [plutonium](https://world-nuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/plutonium \"Plutonium\"), and Pu-239 ends up providing about one-third of the energy from the fuel.\n\nIn most reactors the fuel is ceramic uranium oxide (UO2 with a melting point of 2800\u00b0C) and most is enriched. The fuel pellets (usually about 1 cm diameter and 1.5 cm long) are typically arranged in a long zirconium alloy (zircaloy) tube to form a fuel rod, the zirconium being hard, corrosion-resistant and transparent to neutrons.\\* Numerous rods form a fuel assembly, which is an open lattice and can be lifted into and out of the reactor core. In the most common reactors these are about 4 metres long.\u00a0A BWR fuel assembly may be about 320 kg, a PWR one 655 kg, in which case they hold 183 kg uranium and 460 kgU respectively. In both, about 100 kg of zircaloy is involved.\n\n\\* Zirconium is an important mineral for nuclear power, where it finds its main use. It is therefore subject to controls on trading. It is normally contaminated with hafnium, a neutron absorber, so very pure 'nuclear grade' Zr is used to make the zircaloy, which is about 98% Zr plus about 1.5% tin, also iron, chromium and sometimes nickel to enhance its strength.\n\nA significant industry initiative is to develop accident-tolerant fuels which are more resistant to melting under conditions such as those in the Fukushima accident, and with the cladding being more resistant to oxidation with hydrogen formation at very high temperatures under such conditions.\n\nBurnable poisons are often used in fuel or coolant to even out the performance of the reactor over time from fresh fuel being loaded to refuelling. These are neutron absorbers which decay under neutron exposure, compensating for the progressive build up of neutron absorbers in the fuel as it is burned, and hence allowing higher fuel burn-up (in terms of GW days per tonne of U)\\*. The best known is gadolinium, which is a vital ingredient of fuel in naval reactors where installing fresh fuel is very inconvenient, so reactors are designed to run more than a decade between refuellings (full power equivalent\u00a0\u2013 in practice they are not run continuously). Gadolinium is incorporated in the ceramic fuel pellets. An alternative is zirconium diboride integral fuel burnable absorber (IFBA) as a thin coating on normal pellets.\n\n\\* Average burn-up of fuel used in US reactors has increase to nearly 50 GWd/t, from half that in the 1980s.\n\nGadolinium, mostly at up to 3g oxide per kilogram of fuel, requires slightly higher fuel enrichment to compensate for it, and also after burn-up of about 17 GWd/t it retains about 4% of its absorbtive effect and does not decrease further. The ZrB2 IFBA burns away more steadily and completely, and has no impact on fuel pellet properties. It is now used in most US reactors and a few in Asia. China has the technology for AP1000 reactors.\n\nMain types of nuclear reactor\n-----------------------------\n\n#### Pressurized water reactor (PWR)\n\nThis is the most common type, with about 300 operable reactors for power generation and several hundred more employed for naval propulsion. The design of PWRs originated as a [submarine power plant](https://world-nuclear.org/information-library/non-power-nuclear-applications/transport/nuclear-powered-ships \"Nuclear Powered Ships\"). PWRs use ordinary water as both coolant and moderator. The design is distinguished by having a primary cooling circuit which flows through the core of the reactor under very high pressure, and a secondary circuit in which steam is generated to drive the turbine. In Russia these are known as VVER types\u00a0\u2013\u00a0water-moderated and -cooled.\n\n**A pressurized water reactor (PWR)**\n\n![Image 2: A Pressurized Water Reactor (PWR) main features and components](https://wna.origindigital.co/images/articles/6582d2ca-fa9b-4db9-b972-1ca9dbe924e3.png)\n\nSource: World Nuclear Association\n\nA PWR has fuel assemblies of 200-300 rods each, arranged vertically in the core, and a large reactor would have about 150-250 fuel assemblies with 80-100 tonnes of uranium.\n\nWater in the reactor core reaches about 325\u00b0C, hence it must be kept under about 150 times atmospheric pressure to prevent it boiling. Pressure is maintained by steam in a pressuriser (see diagram). In the primary cooling circuit the water is also the moderator, and if any of it turned to steam the fission reaction would slow down. This negative feedback effect is one of the safety features of the type. The secondary shutdown system involves adding boron to the primary circuit.\n\nThe secondary circuit is under less pressure and the water here boils in the heat exchangers which are thus steam generators. The steam drives the turbine to produce electricity, and is then condensed and returned to the heat exchangers in contact with the primary circuit.\n\n#### Boiling water reactor (BWR)\n\nThis type of reactor has many similarities to the PWR, except that there is only a single circuit in which the water is at lower pressure (about 75 times atmospheric pressure) so that it boils in the core at about 285\u00b0C. The reactor is designed to operate with 12-15% of the water in the top part of the core as steam, and hence with less moderating effect and thus efficiency there. BWR units can operate in load-following mode more readily than PWRs.\n\nThe steam passes through drier plates (steam separators) above the core and then directly to the turbines, which are thus part of the reactor circuit. Since the water around the core of a reactor is always contaminated with traces of radionuclides, it means that the turbine must be shielded and radiological protection provided during maintenance. The cost of this tends to balance the savings due to the simpler design. Most of the radioactivity in the water is very short-lived\\*, so the turbine hall can be entered soon after the reactor is shut down.\n\n\\* mostly N-16, with a 7 second half-life\n\nA BWR fuel assembly comprises 90-100 fuel rods, and there are up to 750 assemblies in a reactor core, holding up to 140 tonnes of uranium. The secondary control system involves restricting water flow through the core so that more steam in the top part reduces moderation.\n\n**A boiling water reactor (PWR)**\n\n![Image 3: A Boiling Water Reactor (BWR) main features and components](https://wna.origindigital.co/images/articles/7d87142d-620d-4835-a968-8ac11e10f6cc.png)\n\nSource: World Nuclear Association\n\nPressurized heavy water reactor (PHWR)\n\nThe PHWR reactor has been developed since the 1950s in Canada as the CANDU, and from 1980s also in India. PHWRs generally use natural uranium (0.7% U-235) oxide as fuel, hence needs a more efficient moderator, in this case heavy water (D2O).\\*\\*\u00a0The PHWR produces more energy per kilogram of mined uranium than other designs,\u00a0but also produces a much larger amount of used fuel per unit output.\n\n\\*\\* with the CANDU system, the moderator is enriched (_i.e._ water) rather than the fuel \u2013 a cost trade-off.\n\nThe moderator is in a large tank called a calandria, penetrated by several hundred horizontal pressure tubes which form channels for the fuel, cooled by a flow of heavy water under high pressure (about 100 times atmospheric pressure) in the primary cooling circuit, typically reaching 290\u00b0C. As in the PWR, the primary coolant generates steam in a secondary circuit to drive the turbines. The pressure tube design means that the reactor can be refuelled progressively without shutting down, by isolating individual pressure tubes from the cooling circuit.\u00a0It is also less costly to build than designs with a large pressure vessel, but the tubes have not proved as durable.\n\n**A pressurized heavy water reactor (PWR)**\n\n![Image 4: A Pressurised Heavy Water Reactor (PHWR/Candu) main features and components](https://wna.origindigital.co/images/articles/d8eddf50-7b52-4ddb-a48f-6eefd0229423.png)\n\nSource: World Nuclear Association\n\nA CANDU fuel assembly consists of a bundle of 37 half metre long fuel rods (ceramic fuel pellets in zircaloy tubes) plus a support structure, with 12 bundles lying end to end in a fuel channel. Control rods penetrate the calandria vertically, and a secondary shutdown system involves adding gadolinium to the moderator. The heavy water moderator circulating through the body of the calandria vessel also yields some heat (though this circuit is not shown on the diagram above).\n\nNewer PHWR designs such as the Advanced Candu Reactor (ACR) have light water cooling and slightly-enriched fuel.\n\nCANDU reactors can accept a variety of fuels. They may be run on recycled uranium from reprocessing LWR used fuel, or a blend of this and depleted uranium left over from enrichment plants. About 4000 MWe of PWR might then fuel 1000 MWe of CANDU capacity, with addition of depleted uranium. Thorium may also be used in fuel.\n\n#### Advanced gas-cooled reactor (AGR)\n\nThese are the second generation of British gas-cooled reactors, using graphite moderator and carbon dioxide as primary coolant. The fuel is uranium oxide pellets, enriched to 2.5\u00a0\\-\u00a03.5%, in stainless steel tubes. The carbon dioxide circulates through the core, reaching 650\u00b0C and then past steam generator tubes outside it, but still inside the concrete and steel pressure vessel (hence 'integral' design). Control rods penetrate the moderator and a secondary shutdown system involves injecting nitrogen to the coolant. The high temperature gives it a high thermal efficiency\u00a0\u2013\u00a0about 41%. Refuelling can be on-load.\n\n**An advanced gas-cooled reactor (AGR)**\n\n![Image 5: An Advanced Gas-Cooled Reactor (AGR) main features and components](https://wna.origindigital.co/images/articles/f944ec7d-43d0-44af-a16e-07713678a922.png)\n\nSource: World Nuclear Association\n\nThe AGR was developed from the Magnox reactor. Magnox reactors were also graphite moderated and CO2 cooled, used natural uranium fuel in metal form, and water as secondary coolant.\u00a0The UK's last Magnox reactor closed at the end of 2015.\n\n#### Light water graphite-moderated reactor (LWGR)\n\nThe main LWGR design is the RBMK, a Soviet design, developed from plutonium production reactors. It employs long (7 metre) vertical pressure tubes running through graphite moderator, and is cooled by water, which is allowed to boil in the core at 290\u00b0C and at about 6.9 MPa, much as in a BWR. Fuel is low-enriched uranium oxide made up into fuel assemblies 3.5 metres long. With moderation largely due to the fixed graphite, excess boiling simply reduces the cooling and neutron absorbtion without inhibiting the fission reaction, and a positive feedback problem can arise, which is why they have never been built outside the Soviet Union. See appendix on [RBMK Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/appendices/rbmk-reactors \"RBMK Reactors\") for further information.\n\n#### Fast neutron reactor (FNR)\n\nSome reactors do not have a moderator and utilise fast neutrons, generating power from plutonium while making more of it from the U-238 isotope in or around the fuel. While they get more than 60 times as much energy from the original uranium compared with normal reactors, they are expensive to build. Further development of them is likely in the next decade, and the main designs expected to be built in two decades are FNRs. If they are configured to produce more fissile material (plutonium) than they consume they are called fast breeder reactors (FBR). See also pages on [Fast Neutron Reactors](https://world-nuclear.org/information-library/current-and-future-generation/fast-neutron-reactors \"Fast Neutron Reactors\")\u00a0and\u00a0[Small Nuclear Power Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/small-nuclear-power-reactors \"Small Nuclear Power Reactors\")\u00a0papers.\n\n#### Operable nuclear power plants\n\n| Reactor type | Main countries | Number | GWe | Fuel | Coolant | Moderator |\n| --- | --- | --- | --- | --- | --- | --- |\n| Pressurized water reactor (PWR) | USA, France, Japan, Russia, China, South Korea | 312 | 300.4 | enriched UO2\n | water | water |\n| Boiling water reactor (BWR) | USA, Japan, Sweden | 60 | 61.0 | enriched UO2 | water | water |\n| Pressurized heavy water reactor (PHWR) | Canada, India | 47 | 24.0 | natural UO2 | heavy water | heavy water |\n| Light water graphite reactor (LWGR) | Russia | 10 | 6.5 | enriched UO2 | water | graphite |\n| Advanced gas-cooled reactor (AGR) | UK | 8 | 4.7 | natural U (metal), enriched UO2 | CO2 | graphite |\n| Fast neutron reactor (FNR) | Russia | 2 | 1.4 | PuO2 and UO2 | liquid sodium | none |\n| High temperature gas-cooled reactor (HTGR) | China | 1 | 0.2 | enriched UO\n\n | helium | graphite |\n\nFor reactors under construction, see information page on\u00a0[Plans for New Reactors Worldwide](https://world-nuclear.org/information-library/current-and-future-generation/plans-for-new-reactors-worldwide \"Plans for New Nuclear Reactors Worldwide\").\n\nAdvanced reactors\n-----------------\n\nSeveral generations of reactors are commonly distinguished. Generation I reactors were developed in the 1950-60s and the last one (Wylfa 1 in the UK) shut down at the end of 2015. They mostly used natural uranium fuel and used graphite as moderator. Generation II reactors are typified by the present US fleet and most in operation elsewhere. They typically use enriched uranium fuel and are mostly cooled and moderated by water. Generation III are the advanced reactors evolved from these, the first few of which are in operation in Japan, China, Russia and the UAE. Others are under construction and ready to be ordered. They are developments of the second generation with enhanced safety. There is no clear distinction between Generation II and Generation III.\n\nGeneration IV designs are still on the drawing board. They will tend to have closed fuel cycles and burn the long-lived actinides now forming part of spent fuel, so that fission products are the only high-level waste. Of seven designs under development with international collaboration, four or five are fast neutron reactors. Four would use fluoride or liquid metal coolants, hence operate at low pressure. Two will be gas-cooled. Most are expected to run at much higher temperatures than today\u2019s water-cooled reactors. See [Generation IV Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/generation-iv-nuclear-reactors \"Generation IV Nuclear Reactors\") paper.\n\nMore than a dozen (Generation III) [advanced reactor](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/advanced-nuclear-power-reactors \"Advanced Nuclear Power Reactors\") designs are in various stages of development. Some are evolutionary from the PWR, BWR and CANDU designs above, some are a more radical departures. The former include the Advanced Boiling Water Reactor, a few of which are now operating with others under construction. Advanced PWRs operate in China, Russia and the UAE, with more under construction. The best-known radical new design has the fuel as large 'pebbles' and uses helium as coolant, at very high temperature, possibly to drive a turbine directly.\n\nConsidering the closed fuel cycle, Generation I-III reactors could recycle plutonium (and possibly uranium), while Generation IV are expected to have full actinide recycle.\n\nMany advanced reactor designs are for small units\u00a0\u2013 under 300 MWe\u00a0\u2013 and in the category of small modular reactors (SMRs), since several of them together may comprise a large power plant, maybe built progressively. Apart from the normal oxide fuels, other fuel types are metal, TRISO\\*, carbide, nitride, or liquid salt.\n\n\\* TRISO (tristructural-isotropic) particles less than a millimetre in diameter. Each has a kernel (_c_. 0.5 mm) of uranium oxycarbide (or uranium dioxide), with the uranium enriched up to 20% U-235. This kernel is surrounded by layers of carbon and silicon carbide, giving a containment for fission products which is stable to over 1600 \u00b0C.\n\nFloating nuclear power plants\n-----------------------------\n\nApart from over 200 nuclear reactors powering various kinds of ships, Rosatom in Russia has set up a subsidiary to supply floating nuclear power plants ranging in size from 70 to 600 MWe. These will be mounted in pairs on a large barge, which will be permanently moored where it is needed to supply power and possibly some desalination to a shore settlement or industrial complex. The first has two 40 MWe reactors based on those in icebreakers and operates at a remote site in Siberia. Electricity cost is expected to be much lower than from present alternatives. For more information see page on [Nuclear Power in Russia](https://world-nuclear.org/information-library/country-profiles/countries-o-s/russia-nuclear-power \"Nuclear Power in Russia\").\n\nThe Russian KLT-40S is a reactor well proven in icebreakers. Here a 150 MWt unit produces 35 MWe (gross) as well as up to 35 MW of heat for desalination or district heating. These are designed to run 3-4 years between refuelling and it is envisaged that they will be operated in pairs to allow for outages, with on-board refuelling capability and used fuel storage. At the end of a 12-year operating cycle the whole plant is taken to a central facility for two-year overhaul and removal of used fuel, before being returned to service.\n\nSecond generation Russian FNPPs will have two 175 MWt, 50 MWe RITM-200M reactor units, each about 1500 tonnes lighter but more powerful than KLT-40S, and thus on a much smaller barge \u2013 about 12,000 rather than 21,000 tonnes displacement. Refuelling will be every 10-12 years. Very similar RITM-200 reactors power the latest Russian icebreakers.\u00a0In December 2022 Rosatom announced that it had developed nuclear fuel for the RITM-200s.\n\nPower rating of a nuclear reactor\n---------------------------------\n\nNuclear plant reactor power outputs are quoted in three ways:\n\n*   Thermal MWt, which depends on the design of the actual nuclear reactor itself, and relates to the quantity and quality of the steam it produces.\n*   Gross electrical MWe, which indicates the power produced by the attached steam turbine and generator, and also takes into account the ambient temperature for the condenser circuit (cooler means more electric power, warmer means less). Rated gross power assumes certain conditions with both.\n*   Net electrical MWe, which is the power available to be sent out from the plant to the grid, after deducting the electrical power needed to run the reactor (cooling and feedwater pumps,\u00a0_etc._) and the rest of the plant.\\*\n\n\\* Net electrical MWe and\u00a0gross MWe vary slightly from summer to winter, so normally the lower summer figure, or an average figure, is used. If the summer figure is quoted, plants may show a capacity factor greater than 100% in cooler times. Watts Bar PWR in Tennessee is reported to run at about 1125 MWe in summer and about 1165 MWe net in winter, due to different condenser cooling water temperatures. Some design options, such as powering the main large feedwater pumps with electric motors (as in EPR or Hualong One) rather than steam turbines (taking steam before it gets to the main turbine-generator), explains some gross to net differences between different reactor types. The EPR has a relatively large drop from gross to net MWe for this reason, and as noted above, the Hualong One needs 20 MWe to run its primary pumps.\n\n**Gross, thermal and net power generation from nuclear power plants**\n\n![Image 6: Gross, thermal and net power generation explanation](https://wna.origindigital.co/images/articles/46ee1eee-438e-42a7-a23b-1d266e75bbad.png)\n\nSource: World Nuclear Association\n\nThe relationship between these is expressed in two ways:\n\n*   Thermal efficiency %, the ratio of gross MWe to MWt. This relates to the difference in temperature between the steam from the reactor and the cooling water. It is often 33-37% in light water reactors, reaching 38% in the latest PWRs.\n*   Net efficiency %, the ratio of net MWe achieved to MWt. This is a little lower, and allows for plant usage.\n\nIn World Nuclear Association information pages and figures and World Nuclear News items, generally net MWe is used for operating plants, and gross MWe for those under construction or planned/proposed.\n\nLifetime of nuclear reactors\n----------------------------\n\nMost of today's nuclear plants which were originally designed for 30 or 40-year operating lives. However, with major investments in systems, structures and components operating lifetimes can be extended, and in several countries there are active programmes to extend operation. In the USA nearly all of the almost 100 reactors have been granted operating licence extensions from 40 to 60 years. This justifies significant capital expenditure in upgrading systems and components, including building in extra performance margins. Some will operate for 80 years or more.\n\nSome components simply wear out, corrode or degrade to a low level of efficiency. These need to be replaced. Steam generators are the most prominent and expensive of these, and many have been replaced after about 30 years where the reactor otherwise has the prospect of running for 60 or more years. This is essentially an economic decision. Lesser components are more straightforward to replace as they age. In Candu reactors, pressure tube replacement has been undertaken on some plants after about 30 years of operation.\n\nA second issue is that of obsolescence. For instance, older reactors have analogue instrument and control systems. Some have been replaced with digital systems. Thirdly, the properties of materials may degrade with age, particularly with heat and neutron irradiation. In respect to all these aspects, investment is needed to maintain reliability and safety. Also, periodic safety reviews are undertaken on older plants in line with international safety conventions and principles to ensure that safety margins are maintained.\n\nAnother important issue is knowledge management over the full lifecycle from design, through construction and operation to decommissioning for reactors and other facilities. This may span a century and involve several countries, and involve a succession of companies. The plant lifespan will cover several generations of engineers. Data needs to be transferable across several generations of software and IT hardware, as well as being shared with other operators of similar plants.\\* Significant modifications may be made to the design over the life of the plant, so original documentation is not sufficient, and loss of design base knowledge can have huge implications (_e.g._ Pickering A and Bruce A in Ontario). Knowledge management is often a shared responsibility and is essential for effective decision-making and the achievement of plant safety and economics.\n\n\\* ISO15926 covers portability and interoperability for lifecycle open data standard. Also EPRI in 2013 published _Advanced Nuclear Technology: New Nuclear Power Plant Information Handover Guide_.\n\nSee also section on _Ageing_, in\u00a0[Safety of Plants](https://world-nuclear.org/information-library/safety-and-security/safety-of-plants/safety-of-nuclear-power-reactors \"Safety of Plants\")\u00a0paper.\n\nPrimary coolants\n----------------\n\nThe advent of some of the designs mentioned above provides opportunity to review the various primary heat transfer fluids used in nuclear reactors. There is a wide variety \u2013 gas, water, light metal, heavy metal and salt:\n\n**Water or heavy water** must be maintained at very high pressure (1000-2200 psi, 7-15 MPa, 150 atmospheres) to enable it to function well above 100\u00b0C, up to 345\u00b0C, as in present reactors. This has a major influence on reactor engineering. However, supercritical water around 25 MPa can give 45% thermal efficiency \u2013 as at some fossil-fuel power plants today with outlet temperatures of 600\u00b0C, and at ultra supercritical levels (30+ MPa) 50% may be attained.\n\nWater cooling of steam condensers is fairly standard in power plants because it works very well, is relatively inexpensive, and there is a huge experience base. Water (at 75 atm pressure) has good heat capacity\u00a0\u2013\u00a0about 4000 kJ/m3 \u2013 so is a lot more effective than gas for removing heat, though its thermal conductivity is less than liquid alternatives.\n\nA possible variation on this is having a high proportion of heavy water in the coolant early in the fuel cycle so that more Pu-239 is bred from U-238, thereby extending the cycle and improving uranium utilization. This is known as spectral shift control.\n\n**Helium** must be used at similar pressure (1000-2000 psi, 7-14 MPa) to maintain sufficient density for efficient operation. However, even at 75 atm pressure its heat capacity is only about 20 kJ/m3. Again, there are engineering implications from the high pressure required, but it can be used in the Brayton cycle to drive a turbine directly.\n\n**Carbon dioxide** was used in early British reactors, and their current AGRs, which operate at much higher temperatures than light water reactors. It is denser than helium and thus likely to give better thermal conversion efficiency. It also leaks less readily than helium. But at very high temperatures \u2013 such as in HTRs \u2013 it breaks down, hence the focus on helium. There is now interest in supercritical CO2 for the Brayton cycle.\n\n**Sodium**, as normally used in fast neutron reactors at around 550\u00baC, melts at 98\u00b0C and boils at 883\u00b0C at atmospheric pressure, so despite the need to keep it dry the engineering required to contain it is relatively modest. It has high thermal conductivity and high heat capacity \u2013 about 1000 kJ/m3 at 2 atm pressure. However, normally water/steam is used in the secondary circuit to drive a turbine (Rankine cycle) at lower thermal efficiency than the Brayton cycle. In some designs sodium is in a secondary circuit to steam generators. Sodium does not corrode the metals used in the fuel cladding or primary circuit, nor the fuel itself if there is cladding damage, but it is very reactive generally.\u00a0In particular it reacts exothermically with water or steam to liberate hydrogen. It burns in air, but much less vigorously. Sodium has a low neutron capture cross-section, but it is enough for some Na-23 to become Na-24, which is a beta-emitter and very gamma-active with 15-hour half-life, so some shielding is required. In a large reactor, with about 5000 t sodium per GWe, Na-24 activity reaches an equilibrium level of nearly 1 TBq/kg \u2013 a large radioactive inventory. If a reactor needs to be shut down frequently, NaK eutectic which is liquid at room temperature (about 13\u00b0C) may be used as coolant, but the potassium is pyrophoric, which increases the hazard.\u00a0Sodium is about six times more transparent to neutrons than lead.\n\n**Lead or lead-bismuth eutectic** in fast neutron reactors are capable of higher temperature operation at atmospheric pressure. They are transparent to neutrons, aiding efficiency due to greater spacing between fuel pins which then allows coolant flow by convection for decay heat removal, and since they do not react with water the heat exchanger interface is safer. They do not burn when exposed to air. However, they are corrosive of fuel cladding and steels, which originally limited temperatures to 550\u00b0C (boiling point of lead is 1750\u00b0C). With today's materials 650\u00b0C can be reached, and in future 800\u00b0C is envisaged with the second stage of Generation IV development, using oxide dispersion-strengthened steels. Lead and Pb-Bi have much higher thermal conductivity than water, but lower than sodium. Rosatom is building a demonstration 300 MWe BREST lead-cooled fast neutron reactor in Russia. Westinghouse is developing a lead-cooled fast reactor concept and LeadCold in Canada is developing one also, using novel aluminium-steel alloys that are highly corrosion-resistant to 450\u00b0C.\u00a0The compound Ti3SiC2 (titanium silicon carbide) is suggested for primary circuits, resisting corrosion.\n\nWhile lead has limited activation from neutrons, a problem with Pb-Bi is that it yields toxic polonium (Po-210) activation product, an alpha-emitter with a half-life of 138 days. Pb-Bi melts at a relatively low 125\u00b0C (hence eutectic) and boils at 1670\u00b0C, Pb melts at 327\u00b0C and boils at 1737\u00b0C but is very much more abundant and cheaper to produce than bismuth, hence is envisaged for large-scale use in the future, though freezing must be prevented. The development of nuclear power based on Pb-Bi cooled fast neutron reactors is likely to be limited to a total of 50-100 GWe, basically for small reactors in remote places. In 1998 Russia declassified a lot of research information derived from its experience with submarine reactors, and US interest in using Pb generally or Pb-Bi for small reactors has increased subsequently. The Gen4 Module (Hyperion) reactor will use lead-bismuth eutectic which is 45% Pb, 55% Bi.\u00a0A secondary circuit generating steam is likely.\n\nFor details of lead-bismuth eutectic coolants, see the 2013 IAEA report in References.\n\n**SALT: Fluoride salts** boil at around 1400\u00b0C at atmospheric pressure, so allow several options for use of the heat, including using helium in a secondary Brayton cycle circuit with thermal efficiencies of 48% at 750\u00b0C to 59% at 1000\u00b0C, for manufacture of hydrogen. Fluoride salts have a very high boiling temperature, very low vapour pressure even at red heat, very high volumetric heat capacity (4670 kJ/m3 for FLiBe, higher than water at 75 atm pressure), good heat transfer properties, low neutron absorbtion, good neutron moderation capability, are not damaged by radiation, are chemically very stable so absorb all fission products well and do not react violently with air or water, are compatible with graphite, and some are also inert to some common structural metals.\u00a0Some gamma-active F-20 is formed by neutron capture, but has very short half-life (11 seconds).\n\nLithium-beryllium fluoride Li2BeF4 (FLiBe) salt is a eutectic version of LiF (2LiF + BeF2) which solidifies at 459\u00b0C and boils at 1430\u00b0C. It is favoured in MSR and AHTR/FHR primary cooling and when uncontaminated has a low corrosion effect. LiF without the toxic beryllium solidifies at about 500\u00b0C and boils at about 1200\u00b0C. FLiNaK (LiF-NaF-KF) is also eutectic and solidifies at 454\u00b0C and boils at 1570\u00b0C. It has a higher neutron cross-section than FLiBe or LiF but can be used intermediate cooling loops.\n\nFor details of molten salt coolants, both as coolant only and as fuel-carriers, see the 2013 IAEA report on [Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors \u2013 Report of the Collaborative Project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)](https://www-pub.iaea.org/MTCD/Publications/PDF/TE_1696_web.pdf \"Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors\").\n\n**Chloride salts** have advantages in fast-spectrum molten salt reactors, having higher solubility for actinides than fluorides. While NaCl has good nuclear, chemical and physical properties its high melting point means it needs to be blended with MgCl2 or CaCl2, the former being preferred in eutectic, and allowing the addition of actinide trichlorides. The major isotope of chlorine, Cl-35 gives rise to Cl-36 as an activation product \u2013 a long-lived energetic beta source, so Cl-37 is much preferable in a reactor. In thermal reactors, chlorides are only candidates for secondary cooling loops.\n\n**All low-pressure liquid coolants** allow all their heat to be delivered at high temperatures, since the temperature drop in heat exchangers is less than with gas coolants. Also, with a good margin between operating and boiling temperatures, passive cooling for decay heat is readily achieved.\u00a0Since heat exchangers do leak to some small extent, having incompatible primary and secondary coolants can be a problem. The less pressure difference across the heat exchanger, the less is the problem.\n\nThe removal of **passive decay heat** is a vital feature of primary cooling systems, beyond heat transfer to do work. When the fission process stops, fission product decay continues and a substantial amount of heat is added to the core. At the moment of shutdown, this is about 6.5% of the full power level, but after an hour it drops to about 1.5% as the short-lived fission products decay. After a day, the decay heat falls to 0.4%, and after a week it will be only 0.2%. This heat could melt the core of a light water reactor unless it is reliably dissipated, as shown in the March 2011 accident at [Fukushima Daiichi](https://world-nuclear.org/information-library/safety-and-security/safety-of-plants/fukushima-daiichi-accident \"Fukushima Daiichi Accident information page\"), where about 1.5% of the heat was being generated when the tsunami disabled the cooling. In passive systems, some kind of convection flow is relied upon. Decay heat removal is more of a problem in gas-cooled reactors due to low thermal inertia, and this has limited the size of individual units.\n\n**Heat transfer for different primary coolants used in nuclear power reactors**\n\n![Image 7: heat transfer for different primary coolants used in nuclear power reactors](https://wna.origindigital.co/images/articles/2d209f25-f1df-481a-b463-5b6030e8aca0.png)\n\nNote: low pressure liquid coolants allow more heat to be delivered\u00a0at higher temperatures\n\nSource: Forsberg[1](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#References \"See Reference 1\")\n\nSee also information page on [Cooling Power Plants](https://world-nuclear.org/information-library/current-and-future-generation/cooling-power-plants \"Cooling Power Plants.\").\n\nThere is some radioactivity in the cooling water flowing through the core of a water-cooled reactor, due mainly to the activation product nitrogen-16, formed by neutron capture from oxygen. N-16 has a half-life of only 7 seconds but produces high-energy gamma radiation during decay. It is the reason that access to a BWR turbine hall is restricted during actual operation.\n\nLoad-following capability\n-------------------------\n\nNuclear power plants are best run continuously at high capacity to meet base-load demand in a grid system. If their power output is ramped up and down on a daily and weekly basis, efficiency is compromised, and in this respect they are similar to most coal-fired plants. (It is also uneconomic to run them at less than full capacity, since they are expensive to build but cheap to run.) However, in some situations it is necessary to vary the output according to daily and weekly load cycles on a regular basis, for instance in France, where there is a very high reliance on nuclear power. Areva has developed its Advanced Load-Following Control System for PWRs that automatically adjusts the plant's electrical output according to the needs of the grid operator. It involves a software upgrade of the reactor control system which varies the plant's output between 50% and 100% of its installed capacity without intervention of the operator. Since 2008, Areva NP has installed the technology at four German nuclear power units, Philippsburg 2 (now shutdown), Isar 2, Brokdorf, and Grohnde, as well as Goesgen in Switzerland.\n\nBWRs can be made to follow loads reasonably easily without burning the core unevenly, by changing the coolant flow rate. Load following is not as readily achieved in a PWR, but especially in France since 1981, so-called 'grey' control rods are used. The ability of a PWR to run at less than full power for much of the time depends on whether it is in the early part of its 18 to 24-month refuelling cycle or late in it, and whether it is designed with special control rods which diminish power levels throughout the core without shutting it down. Thus, though the ability on any individual PWR reactor to run on a sustained basis at low power decreases markedly as it progresses through the refuelling cycle, there is considerable scope for running a fleet of reactors in load-following mode. European Utility Requirements (EUR) since 2001 specify that new reactor designs must be capable of load-following between 50 and 100% of capacity with a rate of change of electric output of 3-5% per minute. The economic consequences are mainly due to diminished load factor of a capital-intensive plant. Further information in the\u00a0[Nuclear Power in France](https://world-nuclear.org/information-library/country-profiles/countries-a-f/france \"Nuclear Power in France\")\u00a0information page and the 2011 Nuclear Energy Agency report,\u00a0[Technical and Economic Aspects of Load Following with Nuclear Power Plants](https://www.oecd-nea.org/ndd/reports/2011/load-following-npp.pdf \"Technical and Economic Aspects of Load Following with Nuclear Power Plants\").\n\nAs fast neutron reactors become established in future years, their ability to load-follow will be a benefit.\n\nNuclear reactors for process heat\n---------------------------------\n\nProducing steam to drive a turbine and generator is relatively easy, and a light water reactor running at 350\u00b0C does this readily. As the above section and Figure show, other types of reactor are required for higher temperatures. A 2010 US Department of Energy document quotes 500\u00b0C for a liquid metal cooled reactor (FNR), 860\u00b0C for a molten salt reactor (MSR), and 950\u00b0C for a high temperature gas-cooled reactor (HTR). Lower-temperature reactors can be used with supplemental gas heating to reach higher temperatures, though employing an LWR would not be practical or economic. The DOE said that high reactor outlet temperatures in the range 750 to 950\u00b0C were required to satisfy all end user requirements evaluated to date for the Next Generation Nuclear Plant.\n\nFor more information see page on [Nuclear Process Heat for Industry](https://world-nuclear.org/information-library/non-power-nuclear-applications/industry/nuclear-process-heat-for-industry \"Nuclear Process Heat for Industry\").\n\nPrimitive reactors\n------------------\n\nThe world's oldest known nuclear reactors operated at what is now Oklo in Gabon, West Africa. About 2 billion years ago, at least 16 natural nuclear reactors achieved criticality in a high-grade deposit of uranium ore (a 17th was in the Bangombe deposit 30 km away). Each operated intermittently at about 20 kW thermal, the reaction ceasing whenever the water turned to steam so that it ceased to function as moderator. At that time the concentration of U-235 in all natural uranium was about 3.6% instead of 0.7% as at present. (U-235 decays much faster than U-238, whose half-life is about the same as the age of the Earth. When the Earth was formed U-235 was about 30% of uranium.) These natural chain reactions started spontaneously and continued overall for one or two million years before finally dying away.\u00a0It appears that each reactor operated in pulses of about 30 minutes. It is estimated that about 130 TWh of heat was produced. (The reactors were discovered when assays of mined uranium showed only 0.717% U-235 instead of 0.720% as everywhere else on the planet. Further investigation identified particular reactor zones with U-235 levels down to 0.44%. There were also significant concentrations of decay nuclides from fission products of\u00a0both uranium and plutonium.)\n\nDuring this long reaction period about 5.4 tonnes of fission products as well as up to two tonnes of plutonium together with other transuranic elements were generated in the orebody. The initial radioactive products have long since decayed into stable elements but close study of the amount and location of these has shown that there was little movement of radioactive wastes during and after the nuclear reactions. Plutonium and the other transuranics remained immobile.\n\n* * *\n\nReferences & notes\n------------------\n\n### General sources\n\nWilson, P.D., The Nuclear Fuel Cycle, OUP (1996)  \nAlex P. Meshik, [The Workings of an Ancient Nuclear Reactor](https://www.scientificamerican.com/article/ancient-nuclear-reactor/ \"The Workings of an Ancient Nuclear Reactor\"), Scientific American (26 January 2009; originally published in the October 2005 edition of _Scientific American_)  \nEvelyn Mervine, [Nature's Nuclear Reactors: The 2-Billion-Year-Old Natural Fission Reactors in Gabon, Western Africa](https://blogs.scientificamerican.com/guest-blog/natures-nuclear-reactors-the-2-billion-year-old-natural-fission-reactors-in-gabon-western-africa/ \"Nature's Nuclear Reactors: The 2-Billion-Year-Old Natural Fission Reactors in Gabon, Western Africa, Scientific American (13 July 2011)\"), Scientific American (13 July 2011)  \n[Technical and Economic Aspects of Load Following with Nuclear Power Plants](https://www.oecd-nea.org/ndd/reports/2011/load-following-npp.pdf \"Technical and Economic Aspects of Load Following with Nuclear Power Plants\"), OECD Nuclear Energy Agency (June 2011)  \nInternational Atomic Energy Agency, [Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors \u2013 Report of the Collaborative Project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)](https://www-pub.iaea.org/MTCD/Publications/PDF/TE_1696_web.pdf \"Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors\"), IAEA-TECDOC-1696 (May 2013)\n\n### References\n\n1.\u00a0C. W. Forsberg, P. F. Peterson and P. S. Pickard, [Molten-Salt-Cooled Advanced High-Temperature Reactor for Production of Hydrogen and Electricity](http://fhr.nuc.berkeley.edu/wp-content/uploads/2014/09/AHTR.Nuclear.Technology.Article.May20.2003.pdf \"C. W. Forsberg et al, Molten-Salt-Cooled Advanced High-Temperature Reactor for Production of Hydrogen and Electricity\"), _Nuclear Technology_, American Nuclear Society (May 2003) \\[[Back](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#1 \"Back\")\\]\n\nAppendices\n----------\n\n[RBMK Reactors](https://world-nuclear.org/information-library/appendices/rbmk-reactors)  \n  \n\nRelated information\n-------------------\n\n[Advanced Nuclear Power Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/advanced-nuclear-power-reactors)  \n[Cooling Power Plants](https://world-nuclear.org/information-library/current-and-future-generation/cooling-power-plants)  \n[Fast Neutron Reactors](https://world-nuclear.org/information-library/current-and-future-generation/fast-neutron-reactors)  \n[Generation IV Nuclear Reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/generation-iv-nuclear-reactors)  \n[France](https://world-nuclear.org/information-library/country-profiles/countries-a-f/france)  \n[Plans For New Reactors Worldwide](https://world-nuclear.org/information-library/current-and-future-generation/plans-for-new-reactors-worldwide)  \n[Physics of Nuclear Energy](https://world-nuclear.org/information-library/nuclear-fuel-cycle/introduction/physics-of-nuclear-energy)  \n[Financing Nuclear Energy](https://world-nuclear.org/information-library/economic-aspects/financing-nuclear-energy)  \n\nContents\n\n* * *\n\n[How does a nuclear reactor work?](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#how-does-a-nuclear-reactor-work) [Components of a nuclear reactor](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#components-of-a-nuclear-reactor) [Fuelling a nuclear reactor](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#fuelling-a-nuclear-reactor) [Main types of nuclear reactor](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#main-types-of-nuclear-reactor) [Advanced reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#advanced-reactors) [Floating nuclear power plants](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#floating-nuclear-power-plants) [Power rating of a nuclear reactor](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#power-rating-of-a-nuclear-reactor) [Lifetime of nuclear reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#lifetime-of-nuclear-reactors) [Primary coolants](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#primary-coolants) [Load-following capability](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#load-following-capability) [Nuclear reactors for process heat](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#nuclear-reactors-for-process-heat) [Primitive reactors](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#primitive-reactors) [References & notes](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#references-amp-notes) [Related Information](https://world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/nuclear-power-reactors#related-information)\n\n![Image 8: World Nuclear Association](https://world-nuclear.org/images/wna-logo_footer.png)\n\nOUR ASSOCIATION\n\n*   [Our Mission](https://world-nuclear.org/our-association/who-we-are)\n*   [Leadership](https://world-nuclear.org/our-association/leadership)\n*   [Our Members](https://world-nuclear.org/our-association/membership)\n*   [Vacancies](https://world-nuclear.org/our-association/vacancies)\n*   [Contact Us](https://world-nuclear.org/our-association/contact-us)\n\nADDRESS\n\nYork House,  \n23 Kingsway,  \nLondon,  \nWC2B 6UJ,  \nUnited Kingdom\n\nGENERAL ENQUIRIES\n\nt: [+44 (0)20 7451 1520](tel:+44 (0)20 7451 1520)  \nf: [+44 (0)20 7839 1501](tel:+44 (0)20 7839 1501)  \ne: [info@world-nuclear.org](mailto:info@world-nuclear.org)\n\nMEMBERSHIP ENQUIRIES\n\nMember support  \n[members@world-nuclear.org](mailto:members@world-nuclear.org) 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+        },
+        {
+            "title": "Graphite in Nuclear Reactors | Jinsun Carbon",
+            "url": "https://jinsuncarbon.com/graphite-in-nuclear-reactors/",
+            "content": "Graphite is commonly used in nuclear reactors as a moderator to slow down neutrons produced during fission.",
+            "raw_content": "![jinsun carbon logo](data:image/svg+xml,%3Csvg%20xmlns='http://www.w3.org/2000/svg'%20viewBox='0%200%20207%2080'%3E%3C/svg%3E)\n![jinsun carbon logo](https://jinsuncarbon.com/wp-content/uploads/2022/12/LOGO-1.png)\n\n### [Have Any Question](tel:+8613131040125)\n\n+8613131040125\n\n### [Send Your Mail](mailto:info@jinsuncarbon.com)\n\ninfo@jinsuncarbon.com\n\n![ondustry.png](data:image/svg+xml,%3Csvg%20xmlns='http://www.w3.org/2000/svg'%20viewBox='0%200%20517%20109'%3E%3C/svg%3E \"ondustry.png\")\n![ondustry.png](https://jinsuncarbon.com/wp-content/uploads/2022/12/ondustry.png \"ondustry.png\")\n\n# A Complete Guide to Understand Graphite in Nuclear Reactors\n\njinsuncarbon\n\nJanuary 4, 2025\n\nGraphite plays an important role in a number of nuclear reactors, especially those which are at high temperatures or blow natural uranium as fuel. Graphite is commonly used in nuclear reactors as a moderator to slow down neutrons produced during fission. Graphite\u2019s role in slowing down these neutrons allows for a much greater probability of causing further induced fission events, thus continuing the chain reaction.\n\nTable of Contents\n\n## Why is Graphite Used in Nuclear Reactors?\n\nNeutron Moderation: The main draw of graphite in [nuclear reactors](https://en.wikipedia.org/wiki/Nuclear_reactor) is its capability to slow down fast neutrons. Neutrons are ejected at speeds much higher than after a fission reaction. The neutrons emitted from the fission processes must be slowed down so that they are more likely to cause further fission reactions in the reactor\u2019s fuel. Graphite serves as a very good neutron moderator, and will not absorb neutrons too much.\n\nHigh Temperature Resistance: Graphite can resist very high temperatures, a critical property in reactors that are intended to function at high temperatures. Graphite is the only material available which maintains its structural integrity even above 1,000\u00b0C, making it well suited for uses in high-temperature gas\u2013cooled reactors (HTGR) and several advanced reactor types.\n\nGraphs are \u201cTransparent\u201d for Neutron: Graphite is a \u201ctransparent\u201d material according to neutron absorption, meaning that it does not absorb a significant number of the neutrons it moderates. This property helps guarantee that enough neutrons stay around to keep the chain reaction going.\n\nProvides Structural Stability: Graphite is a relatively stable and durable material when exposed to extreme conditions; therefore, it provides a structural framework that ensures the proper operation of the reactor. It can also be shaped in its original state to adapt to the reactor shape, which opens up applicability to a large variety of options.\n\n![Graphite in Nuclear Reactors](data:image/svg+xml,%3Csvg%20xmlns='http://www.w3.org/2000/svg'%20viewBox='0%200%20656%20436'%3E%3C/svg%3E)![Graphite in Nuclear Reactors](https://jinsuncarbon.com/wp-content/uploads/2025/01/Graphite-in-Nuclear-Reactors.jpg)\n\n![Graphite in Nuclear Reactors](data:image/svg+xml,%3Csvg%20xmlns='http://www.w3.org/2000/svg'%20viewBox='0%200%20656%20436'%3E%3C/svg%3E)\n![Graphite in Nuclear Reactors](https://jinsuncarbon.com/wp-content/uploads/2025/01/Graphite-in-Nuclear-Reactors.jpg)\n\n## How Does Graphite Work in Nuclear Reactors?\n\nFission Reaction: Fission of uranium or another fissile material creates fast neutrons.\n\nSlower Neutrons for More Fission: The slowed neutrons, now known as thermal neutrons, are more likely to induce more fission when they collide with the uranium fuel. This moderation of the neutron cycle allows for the nuclear chain reaction to be controlled and maintained.\n\n## Graphite as a Moderator in Nuclear Reactors\n\nInelastic Scattering: Neutrons lose energy due to inelastic scattering with the graphite atoms, which leads to their slowing down. This process is very efficient with Graphite\u2019s atomic structure, such that the neutrons losing speed still have enough energetic collision to loss neutron energy to fission.\n\nAvailability and cost: Graphite is abundant in nature and less expensive than other materials that can perform the same function, such as heavy water. This trend enhances the economic feasibility of graphite reactors, especially in high-energy output.\n\n## Functions\n\nNeutron Reflection: Graphite not only slows down neutrons, it also reflects them back into the reactor core. This is an important quality because it helps to constrain the neutrons in the core where they are required, increasing the effectiveness of the reactor.\n\nHeat Control: Although graphite is a very strong conductor of heat, used to carry heat from fission reactions from one area throughout the reactor. This is especially beneficial for reactors that can operate at higher temperatures, such as the HTGRs, because they need to be effective at dissipating heat to avoid overheating.\n\nStructural Functionality: Another role that graphite serves is structural in nature within the reactor core. What are the properties that make it more suited and more stable than the materials that will be needed in reactors that will have complex geometries and that will need precision under extreme conditions?\n\n## Graphite\u2019s Role in Reactor Efficiency\n\nUse of Natural Uranium: A major advantage of [graphite](https://jinsuncarbon.com/the-uses-of-graphite/) as a moderator is it enables reactors to use natural uranium as fuel. Most other reactors require what is known as \u201cenriched\u201d uranium, which is far more expensive than natural uranium, so reactors that do not require the more expensive enriched uranium also operation at a lower operating cost.\n\nEnhanced Operational Temperatures: Reactor cores can also operate at higher temperatures due to graphite. Graphite enables better thermal efficiency in reactors such as the HTGR because it is capable of enduring the considerable heat generated during the fission reaction.\n\n## Safety Considerations\n\nImprovements of Graphite: Through years of exposure to radiation and elevated temperatures, graphite is prone to decompose. This can affect its mildying properties and if this trend persists, possibly lead to structurally compromise. Therefore, its longevity inside reactors necessitates regular inspection and maintenance.\n\nFlammability: Graphite is combustible particularly with oxygen high-temperature condition. This was a major problem in the Chernobyl disaster, where graphite fires exacerbated the nature of the disaster. If graphite gets hot enough, it can catch fire, so extra care must be taken to prevent that from happening if the reactor malfunctions.\n\nRadiation Damage: Long-term exposure to radiation can lead to [physical property](https://jinsuncarbon.com/is-graphite-a-metal-what-are-the-physical-properties-of-graphite/) changes in graphite, including embrittlement or cracking. This can result in poorer performance and higher maintenance requirements.\n\n## The Future of Graphite in Nuclear Reactors\n\nNext-Gen Reactors: Graphite is being studied for use in next-gen nuclear reactors, including small modular reactors (SMRs) and high-temperature gas-cooled reactors (HTGRs). These reactors are smaller, safer, and more efficient, and they continue to have graphite as a key design component.\n\nNew Beginnings: From developing advanced materials like new forms of graphite or composite materials that can withstand even higher levels of radiation and higher temperatures to improve safety and efficiency of the reactor.\n\nSpace Applications: Graphite is also under consideration for use in nuclear reactors meant for out-of-Earth applications, where the need for heat resistance and the capability of neutronic moderation make it an attractive option in space reactors.\n\n## Is graphite radioactive?\n\nPure graphite itself is not radioactive. It is a stable substance composed of carbon. It can be safely used for daily purposes, such as pencils, batteries, industrial lubrication, etc. If graphite is artificially irradiated in the nuclear industry or certain experiments, or comes into contact with radioactive materials (such as uranium, thorium, etc.), it may carry radioactive contamination.\n\n## Conclusion\n\nGraphite was an essential part of nuclear reactor design for many years; it served as a [moderator](https://en.wikipedia.org/wiki/Moderator), structural material, and heat conductor. Its neutrons-lowering capacity combined with thermal stability and minimal neutron absorption makes it essential in reactors operating at natural uranium level and high temperature. Hence, research continues to tackle these major safety challenges while graphite based reactors are being improved upon. As the nuclear energy sector develops over the next decades, graphite could stay an important part of the energy mix for many years.\n\n## Quick Links\n\n## Products\n\n## Contact Info\n\nCopyright \u00a9 2018. 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+        },
+        {
+            "title": "Pebble-bed reactor - Wikipedia",
+            "url": "https://en.wikipedia.org/wiki/Pebble-bed_reactor",
+            "content": "The pebble-bed reactor (PBR) is a design for a graphite-moderated, gas-coolednuclear reactor. The pebble design is relatively simple, with each sphere consisting of the nuclear fuel, fission product barrier, and moderator (which in a traditional water reactor would all be different parts). The primary criticism of pebble-bed reactors is that encasing the fuel in graphite poses a hazard. Main article: Pebble bed modular reactor X-energy is a private American nuclear reactor and fuel design engineering company. It is developing a Generation IV high-temperature gas-cooled pebble-bednuclear reactor design. \"A future for nuclear energy: pebble bed reactors, Int. J. *   Conceptual Design of a Very High Temperature Pebble-Bed Reactor 2003",
+            "raw_content": "Published Time: 2002-11-02T20:35:43Z\n\nPebble-bed reactor - Wikipedia\n===============\n[Jump to content](https://en.wikipedia.org/wiki/Pebble-bed_reactor#bodyContent)\n\n- [x] Main menu \n\nMain menu\n\nmove to sidebar hide\n\n Navigation \n\n*   [Main page](https://en.wikipedia.org/wiki/Main_Page \"Visit the main page [alt-shift-z]\")\n*   [Contents](https://en.wikipedia.org/wiki/Wikipedia:Contents \"Guides to browsing Wikipedia\")\n*   [Current events](https://en.wikipedia.org/wiki/Portal:Current_events \"Articles related to current events\")\n*   [Random article](https://en.wikipedia.org/wiki/Special:Random \"Visit a randomly selected article [alt-shift-x]\")\n*   [About Wikipedia](https://en.wikipedia.org/wiki/Wikipedia:About \"Learn about Wikipedia and how it works\")\n*   [Contact us](https://en.wikipedia.org/wiki/Wikipedia:Contact_us \"How to contact Wikipedia\")\n\n Contribute \n\n*   [Help](https://en.wikipedia.org/wiki/Help:Contents \"Guidance on how to use and edit Wikipedia\")\n*   [Learn to edit](https://en.wikipedia.org/wiki/Help:Introduction \"Learn how to edit Wikipedia\")\n*   [Community portal](https://en.wikipedia.org/wiki/Wikipedia:Community_portal \"The hub for editors\")\n*   [Recent changes](https://en.wikipedia.org/wiki/Special:RecentChanges \"A list of recent changes to Wikipedia [alt-shift-r]\")\n*   [Upload file](https://en.wikipedia.org/wiki/Wikipedia:File_upload_wizard \"Add images or other media for use on Wikipedia\")\n*   [Special pages](https://en.wikipedia.org/wiki/Special:SpecialPages)\n\n[![Image 1](https://en.wikipedia.org/static/images/icons/wikipedia.png)![Image 2: Wikipedia](https://en.wikipedia.org/static/images/mobile/copyright/wikipedia-wordmark-en.svg)![Image 3: The Free Encyclopedia](https://en.wikipedia.org/static/images/mobile/copyright/wikipedia-tagline-en.svg)](https://en.wikipedia.org/wiki/Main_Page)\n\n[Search](https://en.wikipedia.org/wiki/Special:Search \"Search Wikipedia [alt-shift-f]\")\n\nSearch\n\n- [x] Appearance \n\nAppearance\n\nmove to sidebar hide\n\nText\n\n*   Small  Standard  Large   \n\nThis page always uses small font size\n\nWidth\n\n*   Standard  Wide   \n\nThe content is as wide as possible for your browser window.\n\nColor (beta)\n\n*   Automatic  Light  Dark   \n\nThis page is always in light mode.\n\n*   [Donate](https://donate.wikimedia.org/?wmf_source=donate&wmf_medium=sidebar&wmf_campaign=en.wikipedia.org&uselang=en)\n*   [Create account](https://en.wikipedia.org/w/index.php?title=Special:CreateAccount&returnto=Pebble-bed+reactor \"You are encouraged to create an account and log in; however, it is not mandatory\")\n*   [Log in](https://en.wikipedia.org/w/index.php?title=Special:UserLogin&returnto=Pebble-bed+reactor \"You're encouraged to log in; however, it's not mandatory. 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[alt-shift-o]\")\n\n Pages for logged out editors [learn more](https://en.wikipedia.org/wiki/Help:Introduction)\n\n*   [Contributions](https://en.wikipedia.org/wiki/Special:MyContributions \"A list of edits made from this IP address [alt-shift-y]\")\n*   [Talk](https://en.wikipedia.org/wiki/Special:MyTalk \"Discussion about edits from this IP address [alt-shift-n]\")\n\n- [x] Toggle the table of contents \n\nContents\n--------\n\nmove to sidebar hide\n\n*   [(Top)](https://en.wikipedia.org/wiki/Pebble-bed_reactor#)\n*   [1 Design](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Design)\n\n*   [2 Safety](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Safety)Toggle Safety subsection\n    *   [2.1 Containment](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Containment)\n\n    *   [2.2 Fuel production](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Fuel_production)\n\n*   [3 Design criticisms](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Design_criticisms)Toggle Design criticisms subsection\n    *   [3.1 Graphite combustion](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Graphite_combustion)\n\n    *   [3.2 Containment building](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Containment_building)\n\n    *   [3.3 Waste handling](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Waste_handling)\n\n    *   [3.4 2008 report](https://en.wikipedia.org/wiki/Pebble-bed_reactor#2008_report)\n\n*   [4 History](https://en.wikipedia.org/wiki/Pebble-bed_reactor#History)Toggle History subsection\n    *   [4.1 AVR](https://en.wikipedia.org/wiki/Pebble-bed_reactor#AVR)\n        *   [4.1.1 Decommissioning](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Decommissioning)\n\n        *   [4.1.2 Thorium high-temperature reactor](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Thorium_high-temperature_reactor)\n\n    *   [4.2 China](https://en.wikipedia.org/wiki/Pebble-bed_reactor#China)\n\n*   [5 Other designs](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Other_designs)Toggle Other designs subsection\n    *   [5.1 South Africa](https://en.wikipedia.org/wiki/Pebble-bed_reactor#South_Africa)\n\n    *   [5.2 Adams Atomic Engines](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Adams_Atomic_Engines)\n\n    *   [5.3 X-Energy](https://en.wikipedia.org/wiki/Pebble-bed_reactor#X-Energy)\n\n*   [6 See also](https://en.wikipedia.org/wiki/Pebble-bed_reactor#See_also)\n\n*   [7 References](https://en.wikipedia.org/wiki/Pebble-bed_reactor#References)\n\n*   [8 External links](https://en.wikipedia.org/wiki/Pebble-bed_reactor#External_links)\n\nPebble-bed reactor\n==================\n\n- [x] 16 languages \n\n*   [\u0627\u0644\u0639\u0631\u0628\u064a\u0629](https://ar.wikipedia.org/wiki/%D9%85%D9%81%D8%A7%D8%B9%D9%84_%D8%B0%D9%88_%D9%82%D8%A7%D8%B9%D8%AF%D8%A9_%D8%AD%D8%AC%D8%B1%D9%8A%D8%A9 \"\u0645\u0641\u0627\u0639\u0644 \u0630\u0648 \u0642\u0627\u0639\u062f\u0629 \u062d\u062c\u0631\u064a\u0629 \u2013 Arabic\")\n*   [Deutsch](https://de.wikipedia.org/wiki/Kugelhaufenreaktor \"Kugelhaufenreaktor \u2013 German\")\n*   [Espa\u00f1ol](https://es.wikipedia.org/wiki/Reactor_modular_de_lecho_de_bolas \"Reactor modular de lecho de bolas \u2013 Spanish\")\n*   [Fran\u00e7ais](https://fr.wikipedia.org/wiki/R%C3%A9acteur_%C3%A0_lit_de_boulets \"R\u00e9acteur \u00e0 lit de boulets \u2013 French\")\n*   [\ud55c\uad6d\uc5b4](https://ko.wikipedia.org/wiki/%ED%8E%98%EB%B8%94%EB%B2%A0%EB%93%9C_%EC%9B%90%EC%9E%90%EB%A1%9C \"\ud398\ube14\ubca0\ub4dc \uc6d0\uc790\ub85c \u2013 Korean\")\n*   [Bahasa Indonesia](https://id.wikipedia.org/wiki/Reaktor_bola_kerikil \"Reaktor bola kerikil \u2013 Indonesian\")\n*   [Italiano](https://it.wikipedia.org/wiki/Reattore_nucleare_modulare_pebble_bed \"Reattore nucleare modulare pebble bed \u2013 Italian\")\n*   [Bahasa Melayu](https://ms.wikipedia.org/wiki/Reaktor_lapisan_kelikir \"Reaktor lapisan kelikir \u2013 Malay\")\n*   [Nederlands](https://nl.wikipedia.org/wiki/Hogetemperatuurreactor \"Hogetemperatuurreactor \u2013 Dutch\")\n*   [Portugu\u00eas](https://pt.wikipedia.org/wiki/Reator_de_leito_de_esferas \"Reator de leito de esferas \u2013 Portuguese\")\n*   [Sloven\u0161\u010dina](https://sl.wikipedia.org/wiki/Reaktor_na_krogli%C4%8Dno_gorivo \"Reaktor na krogli\u010dno gorivo \u2013 Slovenian\")\n*   [Suomi](https://fi.wikipedia.org/wiki/Kuulareaktori \"Kuulareaktori \u2013 Finnish\")\n*   [Svenska](https://sv.wikipedia.org/wiki/Pebble_bed-reaktor \"Pebble bed-reaktor \u2013 Swedish\")\n*   [\u0e44\u0e17\u0e22](https://th.wikipedia.org/wiki/%E0%B9%80%E0%B8%84%E0%B8%A3%E0%B8%B7%E0%B9%88%E0%B8%AD%E0%B8%87%E0%B8%9B%E0%B8%8F%E0%B8%B4%E0%B8%81%E0%B8%A3%E0%B8%93%E0%B9%8C%E0%B8%99%E0%B8%B4%E0%B8%A7%E0%B9%80%E0%B8%84%E0%B8%A5%E0%B8%B5%E0%B8%A2%E0%B8%A3%E0%B9%8C%E0%B9%81%E0%B8%9A%E0%B8%9A%E0%B8%96%E0%B8%B1%E0%B8%87%E0%B8%81%E0%B8%A3%E0%B8%A7%E0%B8%94 \"\u0e40\u0e04\u0e23\u0e37\u0e48\u0e2d\u0e07\u0e1b\u0e0f\u0e34\u0e01\u0e23\u0e13\u0e4c\u0e19\u0e34\u0e27\u0e40\u0e04\u0e25\u0e35\u0e22\u0e23\u0e4c\u0e41\u0e1a\u0e1a\u0e16\u0e31\u0e07\u0e01\u0e23\u0e27\u0e14 \u2013 Thai\")\n*   [\u0423\u043a\u0440\u0430\u0457\u043d\u0441\u044c\u043a\u0430](https://uk.wikipedia.org/wiki/%D0%A0%D0%B5%D0%B0%D0%BA%D1%82%D0%BE%D1%80_%D0%BD%D0%B0_%D0%B3%D1%80%D0%B0%D0%BD%D1%83%D0%BB%D1%8C%D0%BE%D0%B2%D0%B0%D0%BD%D0%BE%D0%BC%D1%83_%D0%BF%D0%B0%D0%BB%D0%B8%D0%B2%D1%96 \"\u0420\u0435\u0430\u043a\u0442\u043e\u0440 \u043d\u0430 \u0433\u0440\u0430\u043d\u0443\u043b\u044c\u043e\u0432\u0430\u043d\u043e\u043c\u0443 \u043f\u0430\u043b\u0438\u0432\u0456 \u2013 Ukrainian\")\n*   [\u4e2d\u6587](https://zh.wikipedia.org/wiki/%E7%90%83%E5%BA%8A%E5%8F%8D%E6%87%89%E5%A0%86 \"\u7403\u5e8a\u53cd\u61c9\u5806 \u2013 Chinese\")\n\n[Edit links](https://www.wikidata.org/wiki/Special:EntityPage/Q1622414#sitelinks-wikipedia \"Edit interlanguage links\")\n\n*   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\"Structured data on this page hosted by Wikidata [alt-shift-g]\")\n\nFrom Wikipedia, the free encyclopedia\n\nType of very-high-temperature reactor\n\n[![Image 4](https://upload.wikimedia.org/wikipedia/en/thumb/9/99/Question_book-new.svg/60px-Question_book-new.svg.png)](https://en.wikipedia.org/wiki/File:Question_book-new.svg)This article **needs additional citations for [verification](https://en.wikipedia.org/wiki/Wikipedia:Verifiability \"Wikipedia:Verifiability\")**. Please help [improve this article](https://en.wikipedia.org/wiki/Special:EditPage/Pebble-bed_reactor \"Special:EditPage/Pebble-bed reactor\") by [adding citations to reliable sources](https://en.wikipedia.org/wiki/Help:Referencing_for_beginners \"Help:Referencing for beginners\"). Unsourced material may be challenged and removed.\n\n_Find sources:_[\"Pebble-bed reactor\"](https://www.google.com/search?as_eq=wikipedia&q=%22Pebble-bed+reactor%22)\u2013[news](https://www.google.com/search?tbm=nws&q=%22Pebble-bed+reactor%22+-wikipedia&tbs=ar:1)**\u00b7**[newspapers](https://www.google.com/search?&q=%22Pebble-bed+reactor%22&tbs=bkt:s&tbm=bks)**\u00b7**[books](https://www.google.com/search?tbs=bks:1&q=%22Pebble-bed+reactor%22+-wikipedia)**\u00b7**[scholar](https://scholar.google.com/scholar?q=%22Pebble-bed+reactor%22)**\u00b7**[JSTOR](https://www.jstor.org/action/doBasicSearch?Query=%22Pebble-bed+reactor%22&acc=on&wc=on)_(September 2013)_ _([Learn how and when to remove this message](https://en.wikipedia.org/wiki/Help:Maintenance\\_template\\_removal \"Help:Maintenance template removal\"))_\n[![Image 5](https://upload.wikimedia.org/wikipedia/commons/thumb/0/0f/Pebble_bed_reactor_scheme_%28English%29.svg/250px-Pebble_bed_reactor_scheme_%28English%29.svg.png)](https://en.wikipedia.org/wiki/File:Pebble_bed_reactor_scheme_(English).svg)\n\nSketch of a pebble-bed reactor.\n\n**The pebble-bed reactor** (**PBR**) is a design for a graphite-[moderated](https://en.wikipedia.org/wiki/Neutron_moderator \"Neutron moderator\"), [gas-cooled](https://en.wikipedia.org/wiki/Gas_cooled_reactor \"Gas cooled reactor\")[nuclear reactor](https://en.wikipedia.org/wiki/Nuclear_reactor \"Nuclear reactor\"). It is a type of [very-high-temperature reactor](https://en.wikipedia.org/wiki/Very-high-temperature_reactor \"Very-high-temperature reactor\") (VHTR), one of the six classes of nuclear reactors in the [Generation IV initiative](https://en.wikipedia.org/wiki/Generation_IV_reactor \"Generation IV reactor\").\n\n[![Image 6](https://upload.wikimedia.org/wikipedia/commons/thumb/f/f4/Graphitkugel_fuer_Hochtemperaturreaktor.JPG/250px-Graphitkugel_fuer_Hochtemperaturreaktor.JPG)](https://en.wikipedia.org/wiki/File:Graphitkugel_fuer_Hochtemperaturreaktor.JPG)\n\nGraphite pebble for reactor\n\nThe basic design features spherical fuel elements called pebbles. These [tennis ball](https://en.wikipedia.org/wiki/Tennis_ball \"Tennis ball\")-sized elements (approx. 6.7 cm or 2.6 in in diameter) are made of [pyrolytic graphite](https://en.wikipedia.org/wiki/Pyrolytic_carbon \"Pyrolytic carbon\") (which acts as the moderator), and contain thousands of fuel particles called [tristructural-isotropic](https://en.wikipedia.org/wiki/Nuclear_fuel#TRISO_fuel \"Nuclear fuel\") (TRISO) particles. These TRISO particles consist of a fissile material (such as [235 U](https://en.wikipedia.org/wiki/Uranium-235 \"Uranium-235\")) surrounded by a ceramic coating of [silicon carbide](https://en.wikipedia.org/wiki/Silicon_carbide \"Silicon carbide\") for structural integrity and fission product containment. Thousands of pebbles are amassed to create a [reactor core](https://en.wikipedia.org/wiki/Reactor_core \"Reactor core\"). The core is cooled by a gas that does not react chemically with the fuel elements, such as [helium](https://en.wikipedia.org/wiki/Helium \"Helium\"), [nitrogen](https://en.wikipedia.org/wiki/Nitrogen \"Nitrogen\") or [carbon dioxide](https://en.wikipedia.org/wiki/Carbon_dioxide \"Carbon dioxide\"). Other coolants such as [FLiBe](https://en.wikipedia.org/wiki/FLiBe \"FLiBe\") (molten Li(BeF 4))[[1]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-1) have been suggested.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_] The pebble bed design is [passively safe](https://en.wikipedia.org/wiki/Passive_nuclear_safety \"Passive nuclear safety\").[[2]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-MIT_Kadak-2)\n\nBecause the reactor is designed to handle high temperatures, it can cool by natural circulation and survive accident scenarios, which may raise the temperature of the reactor to 1,600 \u00b0C (2,910 \u00b0F). Such high temperatures allow higher thermal efficiencies than possible in traditional nuclear power plants (up to 50%). Additionally, the gases do not dissolve contaminants or absorb neutrons as water does, resulting in fewer radioactive fluids in the core.\n\nThe concept was first suggested by [Farrington Daniels](https://en.wikipedia.org/wiki/Farrington_Daniels \"Farrington Daniels\") in the 1940s, inspired by the innovative design of the [Benghazi burner](https://en.wikipedia.org/wiki/Benghazi_burner \"Benghazi burner\") by British desert troops in WWII. Commercial development came in the 1960s via the [West German](https://en.wikipedia.org/wiki/West_Germany \"West Germany\")[AVR reactor](https://en.wikipedia.org/wiki/AVR_reactor \"AVR reactor\") designed by [Rudolf Schulten](https://en.wikipedia.org/wiki/Rudolf_Schulten \"Rudolf Schulten\").[[3]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-3) This system was plagued with problems and the technology was abandoned.[[4]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-4) The AVR design was licensed to [South Africa](https://en.wikipedia.org/wiki/South_Africa \"South Africa\") as the [PBMR](https://en.wikipedia.org/wiki/PBMR \"PBMR\") and [China](https://en.wikipedia.org/wiki/China \"China\") as the [HTR-10](https://en.wikipedia.org/wiki/HTR-10 \"HTR-10\"). The HTR-10 prototype was developed into China's [HTR-PM](https://en.wikipedia.org/wiki/HTR-PM \"HTR-PM\") demonstration plant, which connects two reactors to a single turbine producing 210 MW e, operating commercially since 2023. Other designs are under development by [MIT](https://en.wikipedia.org/wiki/Massachusetts_Institute_of_Technology \"Massachusetts Institute of Technology\"), [University of California at Berkeley](https://en.wikipedia.org/wiki/University_of_California_at_Berkeley \"University of California at Berkeley\"), [General Atomics](https://en.wikipedia.org/wiki/General_Atomics \"General Atomics\") (U.S.), [Dutch](https://en.wikipedia.org/wiki/Netherlands \"Netherlands\") company Romawa B.V., [Adams Atomic Engines](https://en.wikipedia.org/wiki/Adams_Atomic_Engines \"Adams Atomic Engines\"), [Idaho National Laboratory](https://en.wikipedia.org/wiki/Idaho_National_Laboratory \"Idaho National Laboratory\"), [X-energy](https://en.wikipedia.org/wiki/X-energy \"X-energy\") and Kairos Power.\n\nDesign\n------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=1 \"Edit section: Design\")]\n\n[![Image 7](https://upload.wikimedia.org/wikipedia/en/thumb/9/99/Question_book-new.svg/60px-Question_book-new.svg.png)](https://en.wikipedia.org/wiki/File:Question_book-new.svg)This section **needs additional citations for [verification](https://en.wikipedia.org/wiki/Wikipedia:Verifiability \"Wikipedia:Verifiability\")**. Please help [improve this article](https://en.wikipedia.org/wiki/Special:EditPage/Pebble-bed_reactor \"Special:EditPage/Pebble-bed reactor\") by [adding citations to reliable sources](https://en.wikipedia.org/wiki/Help:Referencing_for_beginners \"Help:Referencing for beginners\") in this section. Unsourced material may be challenged and removed._(January 2021)_ _([Learn how and when to remove this message](https://en.wikipedia.org/wiki/Help:Maintenance\\_template\\_removal \"Help:Maintenance template removal\"))_\n\nA pebble-bed power plant combines a gas-cooled core[[5]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-5) and a novel fuel packaging.[[6]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-6)\n\nThe [uranium](https://en.wikipedia.org/wiki/Uranium \"Uranium\"), [thorium](https://en.wikipedia.org/wiki/Thorium \"Thorium\") or [plutonium](https://en.wikipedia.org/wiki/Plutonium \"Plutonium\")[nuclear fuels](https://en.wikipedia.org/wiki/Nuclear_fuel \"Nuclear fuel\") are in the form of a [ceramic](https://en.wikipedia.org/wiki/Ceramic \"Ceramic\") (usually [oxides](https://en.wikipedia.org/wiki/Oxide \"Oxide\") or [carbides](https://en.wikipedia.org/wiki/Carbide \"Carbide\")) contained within spherical pebbles a little smaller than the size of a tennis ball and made of pyrolytic graphite, which acts as the primary [neutron moderator](https://en.wikipedia.org/wiki/Neutron_moderator \"Neutron moderator\"). The pebble design is relatively simple, with each sphere consisting of the nuclear fuel, fission product barrier, and moderator (which in a traditional water reactor would all be different parts). Grouping sufficient pebbles in the correct geometry creates [criticality](https://en.wikipedia.org/wiki/Critical_mass \"Critical mass\").\n\nThe pebbles are held in a vessel, and an [inert gas](https://en.wikipedia.org/wiki/Inert_gas \"Inert gas\") (such as helium, nitrogen or carbon dioxide) circulates through the spaces between the fuel pebbles to carry heat away from the reactor. Pebble-bed reactors must keep the pebbles' [graphite](https://en.wikipedia.org/wiki/Graphite \"Graphite\") from burning in the presence of air if the reactor wall is breached (the flammability of the pebbles is [disputed](https://en.wikipedia.org/wiki/Pebble-bed_reactor#Containment)). The heated gas is run directly through a [turbine](https://en.wikipedia.org/wiki/Turbine \"Turbine\"). However, if the gas from the primary [coolant](https://en.wikipedia.org/wiki/Coolant \"Coolant\") can be made radioactive by the [neutrons](https://en.wikipedia.org/wiki/Neutron \"Neutron\") in the reactor, or a fuel defect could contaminate the power production equipment, it may be brought instead to a [heat exchanger](https://en.wikipedia.org/wiki/Heat_exchanger \"Heat exchanger\") where it heats another gas or produces steam. The turbine exhaust is warm and may be used to heat buildings or in other applications.\n\nPebble-bed reactors are gas-cooled, sometimes at low pressures. The spaces between the pebbles replace the piping in conventional reactors. Since there is no actual piping in the core and the coolant contains no hydrogen, embrittlement is not a failure concern. The preferred gas, helium, does not easily absorb neutrons or impurities. Therefore, compared to water, it is both more efficient and less likely to become radioactive.\n\nMuch of the cost of a [conventional, water-cooled nuclear power plant](https://en.wikipedia.org/wiki/Pressurized_water_reactor \"Pressurized water reactor\") is due to cooling system complexity, which is not a factor in PBRs. Conventional plants require extensive safety systems and redundant backups. Their reactor cores are dwarfed by cooling systems. Further, the core irradiates the water with neutrons causing the water and impurities dissolved in it to become radioactive. The high-pressure piping in the primary side eventually becomes [embrittled](https://en.wikipedia.org/wiki/Hydrogen_embrittlement \"Hydrogen embrittlement\") and requires inspection and replacement.\n\nSome designs are throttled by temperature rather than [control rods](https://en.wikipedia.org/wiki/Control_rod \"Control rod\"). Such reactors do not need to operate well at the varying neutron profiles caused by partially withdrawn control rods.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\nPBRs can use fuel pebbles made from various fuels in the same design (though perhaps not simultaneously). Proponents claim that pebble-bed reactors can use thorium, plutonium and natural unenriched uranium, as well as [enriched uranium](https://en.wikipedia.org/wiki/Enriched_uranium \"Enriched uranium\").\n\nIn most stationary designs, fuel replacement is continuous. Pebbles are placed in a bin-shaped reactor. Pebbles travel from the bottom to the top about ten times over a period of years, and are tested after each pass. Expended pebbles are removed to the nuclear-waste area, replaced by a new pebble.\n\nSafety\n------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=2 \"Edit section: Safety\")]\n\nWhen the reactor temperature rises, the atoms in the fuel move rapidly, causing [Doppler broadening](https://en.wikipedia.org/wiki/Doppler_broadening \"Doppler broadening\"). The fuel then experiences a wider range of neutron speeds. [Uranium-238](https://en.wikipedia.org/wiki/Uranium-238 \"Uranium-238\"), which forms the bulk of the uranium, is much more likely to absorb fast or [epithermal neutrons](https://en.wikipedia.org/wiki/Neutron_temperature \"Neutron temperature\") at higher temperatures. This reduces the number of neutrons available to cause fission, and reduces power. Doppler broadening therefore creates a negative feedback: as fuel temperature increases, reactor power decreases. All reactors have reactivity feedback mechanisms. The pebble-bed reactor is designed so that this effect is relatively strong, inherent to the design, and does not depend on moving parts. This negative feedback creates passive control of the reaction process.\n\nThus PBRs passively reduce to a safe power-level in an accident scenario. This is the design's main passive safety feature.\n\nThe reactor is cooled by an inert, fireproof gas, which has no phase transitions\u2014it is always in the gaseous phase. The moderator is solid carbon; it does not act as a coolant, or move, or change phase.\n\nConvection of the gas, driven by the heat of the pebbles, ensures that the pebbles are passively cooled.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\nEven in the event that all supporting machinery fails, the reactor will not crack, melt, explode or spew hazardous wastes. It heats to a designed \"idle\" temperature, and stays there. At idle, the reactor vessel radiates heat, but the vessel and fuel spheres remain intact and undamaged. The machinery can be repaired or the fuel can be removed.\n\nIn a safety test using the German AVR reactor, all the control rods were removed, and coolant flow was halted. The fuel remained undamaged.[[7]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-ISR-7)\n\nPBRs are intentionally operated above the 250\u00b0C (482\u00b0F) [annealing](https://en.wikipedia.org/wiki/Annealing_(metallurgy) \"Annealing (metallurgy)\") temperature of graphite, so that [Wigner energy](https://en.wikipedia.org/wiki/Wigner_energy \"Wigner energy\") does not accumulate. This solves a problem discovered in the [Windscale fire](https://en.wikipedia.org/wiki/Windscale_fire \"Windscale fire\"). One reactor (not a PBR) caught fire because of the release of energy stored as crystalline dislocations (Wigner energy) in the graphite. The dislocations are caused by neutron passage through the graphite. Windscale regularly annealed the graphite to release accumulated Wigner energy. However, the effect was not anticipated, and since the reactor was cooled by ambient air in an open cycle, the process could not be reliably controlled, and led to a fire.\n\nBerkeley professor [Richard A. Muller](https://en.wikipedia.org/wiki/Richard_A._Muller \"Richard A. Muller\") described PBRs as \"in every way ... safer than the present nuclear reactors\".[[8]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-8)\n\n### Containment\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=3 \"Edit section: Containment\")]\n\nMost PBR designs include multiple reinforcing levels of containment to prevent contact between the radioactive materials and the biosphere:\n\n*   Most reactors are enclosed in a [containment building](https://en.wikipedia.org/wiki/Containment_building \"Containment building\") designed to resist aircraft crashes and earthquakes.\n*   The reactor is usually in a room with two-meter-thick walls with doors that can be closed, and cooling [plenums](https://en.wikipedia.org/wiki/Plenum_space \"Plenum space\") that can be filled with water.\n*   The reactor vessel is typically sealed.\n*   Each pebble, within the vessel, is a 60 millimetres (2.4 in) hollow sphere of pyrolytic graphite, wrapped in fireproof silicon carbide.\n*   Low density porous pyrolytic carbon, high density nonporous pyrolytic carbon\n*   The fission fuel is in the form of metal oxides or carbides.\n\nPyrolytic graphite is the main structural material in pebbles. It sublimates at 4,000\u00b0C (7,230\u00b0F), more than double the design temperature of most reactors. It slows neutrons effectively, is strong, inexpensive, and has a long history of use in reactors and other high temperature applications. For example, pyrolytic graphite is also used, unreinforced, to construct missile reentry nose-cones and large solid rocket nozzles.[[9]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-9) Its strength and hardness come from its anisotropic crystals.\n\nPyrolytic carbon can burn in air when the reaction is catalyzed by a hydroxyl radical (e.g., from water).[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_] Infamous examples include the [accidents](https://en.wikipedia.org/wiki/List_of_nuclear_accidents \"List of nuclear accidents\")at Windscale[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_] and Chernobyl\u2014both graphite-moderated reactors. However, PBRs are cooled by inert gases to prevent fire. All designs have at least one layer of silicon carbide that serves as a fire break and seal.\n\n### Fuel production\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=4 \"Edit section: Fuel production\")]\n\nAll kernels are precipitated from a [sol-gel](https://en.wikipedia.org/wiki/Sol-gel \"Sol-gel\"), then washed, dried and calcined. U.S. kernels use [uranium carbide](https://en.wikipedia.org/wiki/Uranium_carbide \"Uranium carbide\"), while German (AVR) kernels use [uranium dioxide](https://en.wikipedia.org/wiki/Uranium_dioxide \"Uranium dioxide\"). German-produced fuel-pebbles release about 1000 times less radioactive gas than the U.S. equivalents, due to that construction method.[[10]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-Key-diff-2002-10)[[11]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-Key-diff-2003-11)\n\n[![Image 8: [icon]](https://upload.wikimedia.org/wikipedia/commons/thumb/1/1c/Wiki_letter_w_cropped.svg/20px-Wiki_letter_w_cropped.svg.png)](https://en.wikipedia.org/wiki/File:Wiki_letter_w_cropped.svg)This section **needs expansion**. You can help by [adding to it](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=). _(October 2021)_\n\nDesign criticisms\n-----------------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=5 \"Edit section: Design criticisms\")]\n\n### Graphite combustion\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=6 \"Edit section: Graphite combustion\")]\n\nThe primary criticism of pebble-bed reactors is that encasing the fuel in graphite poses a hazard. Graphite can burn in the presence of air, which could happen if the reactor vessel is compromised. Fire could vaporize the fuel, which could then be released to the surroundings. Fuel kernels are coated with a layer of silicon carbide to isolate the graphite. While silicon carbide is strong in abrasion and [compression](https://en.wikipedia.org/wiki/Compression_(physical) \"Compression (physical)\") applications, it has less resistance to expansion and shear forces. Some [fission](https://en.wikipedia.org/wiki/Nuclear_fission \"Nuclear fission\") products such as 133\n\nXe have limited absorbance in carbon, so some fuel kernels could accumulate enough gas to rupture the silicon carbide.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\n### Containment building\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=7 \"Edit section: Containment building\")]\n\nSome designs do not include a containment building, leaving reactors more vulnerable to attack. However, most are surrounded by a reinforced concrete containment structure.[[12]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-12)\n\n### Waste handling\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=8 \"Edit section: Waste handling\")]\n\nPBR waste volumes are much greater, but have similar [radioactivity](https://en.wikipedia.org/wiki/Radioactivity \"Radioactivity\") measured in [becquerels](https://en.wikipedia.org/wiki/Becquerel \"Becquerel\") per [kilowatt-hour](https://en.wikipedia.org/wiki/Kilowatt-hour \"Kilowatt-hour\"). The waste tends to be less hazardous and simpler to handle.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_] Current US [legislation](https://en.wikipedia.org/wiki/Legislation \"Legislation\") requires all waste to be safely contained, requiring waste storage facilities. Pebble defects may complicate storage. Graphite pebbles are more difficult to reprocess due to their construction.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\n### 2008 report\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=9 \"Edit section: 2008 report\")]\n\nIn 2008, a report[[13]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-13)[[14]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-14) about safety aspects of Germany's [AVR reactor](https://en.wikipedia.org/wiki/AVR_reactor \"AVR reactor\") and general PBR features drew attention. The claims are contested.[[15]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-15) The report cited:\n\n*   Impossible to place standard measurement equipment in the reactor core[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n*   The cooling circuit can be contaminated with metallic fission products ([90 Sr](https://en.wikipedia.org/wiki/Strontium-90 \"Strontium-90\"), [137 Cs](https://en.wikipedia.org/wiki/Caesium-137 \"Caesium-137\")) due to limited pebble retention capabilities for metallic fission products. The report claimed that even modern fuel elements do not sufficiently retain [strontium](https://en.wikipedia.org/wiki/Strontium \"Strontium\") and [caesium](https://en.wikipedia.org/wiki/Caesium \"Caesium\").\n*   Elevated core temperatures (>200\u00b0C or 360\u00b0F above calculated values)\n*   Dust formation from pebble friction under pebble breach (Dust acts as a mobile fission product carrier, if fission products escape the fuel particles.)\n\nReport author [Rainer Moormann](https://en.wikipedia.org/wiki/Rainer_Moormann \"Rainer Moormann\"), recommended that average hot helium temperatures be limited to 800\u00b0C (1,470\u00b0F) minus the uncertainty of the core temperatures (about 200\u00b0C or 360\u00b0F).\n\nHistory\n-------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=10 \"Edit section: History\")]\n\n[Farrington Daniels](https://en.wikipedia.org/wiki/Farrington_Daniels \"Farrington Daniels\") originated the concept and the name in 1947 at Oak Ridge.[[16]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-16)[Rudolf Schulten](https://en.wikipedia.org/wiki/Rudolf_Schulten \"Rudolf Schulten\") advanced the idea in the 1950s. The crucial insight was to combine fuel, structure, containment, and neutron moderator in a small, strong sphere. The concept depended on the availability of engineered forms of silicon carbide and pyrolytic carbon that were strong.\n\n### AVR\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=11 \"Edit section: AVR\")]\n\n![Image 9](https://upload.wikimedia.org/wikipedia/en/thumb/b/b4/Ambox_important.svg/40px-Ambox_important.svg.png)hide**This article has multiple issues.** Please help **[improve it](https://en.wikipedia.org/wiki/Special:EditPage/Pebble-bed_reactor \"Special:EditPage/Pebble-bed reactor\")** or discuss these issues on the **[talk page](https://en.wikipedia.org/wiki/Talk:Pebble-bed_reactor \"Talk:Pebble-bed reactor\")**. _([Learn how and when to remove these messages](https://en.wikipedia.org/wiki/Help:Maintenance\\_template\\_removal \"Help:Maintenance template removal\"))_ ![Image 10](https://upload.wikimedia.org/wikipedia/commons/thumb/9/98/Ambox_current_red.svg/60px-Ambox_current_red.svg.png)This section needs to be **updated**. Please help update this article to reflect recent events or newly available information._(December 2023)_ [![Image 11](https://upload.wikimedia.org/wikipedia/en/thumb/9/99/Question_book-new.svg/60px-Question_book-new.svg.png)](https://en.wikipedia.org/wiki/File:Question_book-new.svg)This section **needs additional citations for [verification](https://en.wikipedia.org/wiki/Wikipedia:Verifiability \"Wikipedia:Verifiability\")**. Please help [improve this article](https://en.wikipedia.org/wiki/Special:EditPage/Pebble-bed_reactor \"Special:EditPage/Pebble-bed reactor\") by [adding citations to reliable sources](https://en.wikipedia.org/wiki/Help:Referencing_for_beginners \"Help:Referencing for beginners\") in this section. Unsourced material may be challenged and removed._(December 2023)_ _([Learn how and when to remove this message](https://en.wikipedia.org/wiki/Help:Maintenance\\_template\\_removal \"Help:Maintenance template removal\"))_ _([Learn how and when to remove this message](https://en.wikipedia.org/wiki/Help:Maintenance\\_template\\_removal \"Help:Maintenance template removal\"))_\n\nMain article: [AVR reactor](https://en.wikipedia.org/wiki/AVR_reactor \"AVR reactor\")\n\n[![Image 12](https://upload.wikimedia.org/wikipedia/commons/thumb/c/c8/Hogetemperatuurreactor.JPG/250px-Hogetemperatuurreactor.JPG)](https://en.wikipedia.org/wiki/File:Hogetemperatuurreactor.JPG)\n\nAVR in Germany.\n\nA 15 [MW e](https://en.wikipedia.org/wiki/MWe#MWe,_MWt \"MWe\") demonstration reactor, Arbeitsgemeinschaft Versuchsreaktor (_experimental reactor consortium_), was built at the [J\u00fclich Research Centre](https://en.wikipedia.org/wiki/J%C3%BClich_Research_Centre \"J\u00fclich Research Centre\") in [J\u00fclich](https://en.wikipedia.org/wiki/J%C3%BClich \"J\u00fclich\"), [West Germany](https://en.wikipedia.org/wiki/West_Germany \"West Germany\"). The goal was to gain operational experience with a high-temperature gas-cooled reactor. Construction costs of AVR were 115 million Deutschmark (1966), corresponding to a 2010 value of 180 million \u20ac. The unit's first criticality was on August 26, 1966. The facility ran successfully for 21 years.\n\nIn 1978, the AVR suffered from a water/steam ingress accident of 30 metric tons (30 long tons; 33 short tons), which led to contamination of soil and groundwater by strontium-90 and by tritium.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_] The leak in the steam generator leading to this accident was probably caused by high core temperatures (see criticism section). A re-examination of this accident was announced by the local government in July 2010.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\nThe AVR was originally designed to breed [uranium-233](https://en.wikipedia.org/wiki/Uranium-233 \"Uranium-233\") from [thorium-232](https://en.wikipedia.org/wiki/Thorium-232 \"Thorium-232\"). A practical thorium [breeder reactor](https://en.wikipedia.org/wiki/Breeder_reactor \"Breeder reactor\") was considered valuable technology. However, the AVR's fuel design contained the fuel so well that the transmuted fuels were uneconomic to extract\u2014it was cheaper to use mined and purified uranium.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\nThe AVR used helium [coolant](https://en.wikipedia.org/wiki/Coolant \"Coolant\"), has a low [neutron cross-section](https://en.wikipedia.org/wiki/Neutron_cross-section \"Neutron cross-section\"). Since few neutrons are absorbed, the coolant remains less radioactive. It is practical to route the primary coolant directly to power generation turbines. Even though the power generation used primary coolant, it was reported that the AVR exposed its personnel to less than 1/5 as much radiation as a typical light water reactor.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\n#### Decommissioning\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=12 \"Edit section: Decommissioning\")]\n\nIt was decommissioned on December 1, 1988, in the wake of the [Chernobyl disaster](https://en.wikipedia.org/wiki/Chernobyl_accident \"Chernobyl accident\") and operational problems. During removal of the fuel elements it became apparent that the neutron reflector under the pebble-bed core had cracked during operation. Some hundred fuel elements remained stuck in the crack. During this examination it was revealed that the AVR was the world's most heavily beta-contaminated ([strontium-90](https://en.wikipedia.org/wiki/Strontium-90 \"Strontium-90\")) nuclear installation and that this contamination was present as dust (the worst form).[[17]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-17)\n\nLocalized fuel temperature instabilities resulted in heavy vessel contamination by [Cs-137](https://en.wikipedia.org/wiki/Cs-137 \"Cs-137\") and [Sr-90](https://en.wikipedia.org/wiki/Sr-90 \"Sr-90\"). The reactor vessel was filled with light concrete in order to fix the radioactive dust and in 2012 the reactor vessel of 2,100 metric tons (2,100 long tons; 2,300 short tons) was to be moved to intermediate storage until a permanent solution is devised. The reactor buildings were to be dismantled and soil and groundwater decontaminated. AVR dismantling costs were expected to far exceed its construction costs. In August 2010, the German government estimated costs for AVR dismantling without consideration of the vessel dismantling at 600 million \u20ac ( $750 million, which corresponded to 0.4 \u20ac ($0.55) per kWh of electricity generated by the AVR. A separate containment was erected for dismantling purposes, as seen in the AVR-picture.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\n#### Thorium high-temperature reactor\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=13 \"Edit section: Thorium high-temperature reactor\")]\n\nMain article: [THTR-300](https://en.wikipedia.org/wiki/THTR-300 \"THTR-300\")\n\nFollowing the experience with the AVR, Germany constructed a full scale power station (the thorium high-temperature reactor or [THTR-300](https://en.wikipedia.org/wiki/THTR-300 \"THTR-300\") rated at 300 MW), using thorium as the fuel. THTR-300 suffered technical difficulties, and owing to these and political events in Germany, was closed after four years of operation. An incident on 4 May 1986, only a few days after the Chernobyl disaster, allowed a release of part of the radioactive inventory into the environment. Although the radiological impact was small, it had a disproportionate impact. The release was caused by a human error during a blockage of pebbles in a pipe. Trying to restart the pebbles' movement by increasing gas flow stirred up dust, always present in PBRs, which was then released, unfiltered, into the environment due to an erroneously open valve.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\nIn spite of the limited amount of radioactivity released (0.1 GBq [60 Co](https://en.wikipedia.org/wiki/Cobalt \"Cobalt\"), [137 Cs](https://en.wikipedia.org/wiki/Caesium \"Caesium\"), [233 Pa](https://en.wikipedia.org/wiki/Protactinium \"Protactinium\")), a commission of inquiry was appointed. The radioactivity in the vicinity of the THTR-300 was finally found to result 25% from Chernobyl and 75% from THTR-300. The handling of this minor accident severely damaged the credibility of the German pebble-bed community, which lost support in Germany.[[18]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-18)\n\nThe overly complex design of the reactor, which is contrary to the general concept of self-moderated thorium reactors designed in the U.S., also suffered from the unplanned high destruction rate of pebbles during the test series and the resulting higher contamination of the containment structure. Pebble debris and graphite dust blocked some of the coolant channels in the bottom reflector, as was discovered during fuel removal after final shut-down. A failure of insulation required frequent reactor shut-downs for inspection, because the insulation could not be repaired. Metallic components in the hot gas duct failed in September 1988, probably due to thermal fatigue induced by unexpected hot gas currents.[[19]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-19) This failure led to a long shut-down for inspections. In August, 1989, the THTR company almost went bankrupt, but was rescued by the government. The unexpected high costs of THTR operation and the accident ended interest in THTR reactors. The government decided to terminate the THTR operation at the end of September, 1989. This particular reactor was built despite criticism at the design phase. Most of those design critiques by German physicists, and by American physicists at the National Laboratory level, went ignored until shutdown. Nearly every problem encountered by the THTR 300 reactor was predicted by the physicists who criticized it as \"overly complex\".[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\n### China\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=14 \"Edit section: China\")]\n\nIn 2004 China licensed the AVR technology and developed a reactor for power generation.[[20]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-20) The 10 megawatt prototype is called the [HTR-10](https://en.wikipedia.org/wiki/HTR-10 \"HTR-10\"). It is a conventional helium-cooled, helium-turbine design. In 2021 the Chinese then built a 211 MW e gross unit [HTR-PM](https://en.wikipedia.org/wiki/HTR-PM \"HTR-PM\"), which incorporates two 250 MW t reactors.[[21]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-HTR-PM-2021-21) As of 2021[[update]](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit), four sites were being considered for a 6-reactor successor, the HTR-PM600.[[21]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-HTR-PM-2021-21) The reactor entered service in December 2023.[[22]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-22)\n\nOther designs\n-------------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=15 \"Edit section: Other designs\")]\n\n### South Africa\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=16 \"Edit section: South Africa\")]\n\nMain article: [Pebble bed modular reactor](https://en.wikipedia.org/wiki/Pebble_bed_modular_reactor \"Pebble bed modular reactor\")\n\nIn June 2004, it was announced that a new PBMR would be built at [Koeberg](https://en.wikipedia.org/wiki/Koeberg \"Koeberg\"), [South Africa](https://en.wikipedia.org/wiki/South_Africa \"South Africa\") by [Eskom](https://en.wikipedia.org/wiki/Eskom \"Eskom\"), the government-owned electrical utility to operate at 940\u00b0C (1,720\u00b0F).[[23]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-23) The PBMR project was opposed by groups such as [Koeberg Alert](https://en.wikipedia.org/wiki/Koeberg_Alert \"Koeberg Alert\") and [Earthlife Africa](https://en.wikipedia.org/wiki/Earthlife_Africa \"Earthlife Africa\"), the latter of which sued Eskom.[[24]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-24) The reactor was never completed and the testing facility was decommissioned and placed in a \"care and maintenance mode\" to protect the IP and the assets.[[25]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-25)\n\nA Pretoria-based company, Stratek Global, created a variant of the PBMR reactor. The Stratek HTMR-100 reactor functions at 750\u00b0C (1,380\u00b0F). It directs the heat into water to create steam and is helium-cooled. The HTMR-100 reactor produces output of 35 MWe.[[26]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-26)\n\n### Adams Atomic Engines\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=17 \"Edit section: Adams Atomic Engines\")]\n\nAdams Atomic Engines (AAE) design was self-contained so it could be adapted to extreme environments such as space, polar and underwater environments. Their design was for a nitrogen coolant passing directly though a conventional low-pressure gas turbine,[[27]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-Adams1994-27) and due to the rapid ability of the turbine to change speeds, it can be used in applications where instead of the turbine's output being converted to electricity, the turbine itself could directly drive a mechanical device, for instance, a propeller aboard a ship.\n\nLike all high temperature designs, the AAE engine would have been inherently safe, as the engine naturally shuts down due to [Doppler broadening](https://en.wikipedia.org/wiki/Doppler_broadening \"Doppler broadening\"), stopping heat generation if the fuel in the engine gets too hot in the event of a loss of coolant or a loss of coolant flow.[_[citation needed](https://en.wikipedia.org/wiki/Wikipedia:Citation\\_needed \"Wikipedia:Citation needed\")_]\n\nThe company went out of business in December 2010.[[28]](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_note-28)\n\n### X-Energy\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=18 \"Edit section: X-Energy\")]\n\nThis section is an excerpt from [X-energy](https://en.wikipedia.org/wiki/X-energy \"X-energy\").[[edit](https://en.wikipedia.org/w/index.php?title=X-energy&action=edit)]\n\n[X-energy](https://en.wikipedia.org/wiki/X-energy \"X-energy\") is a private American nuclear reactor and fuel design engineering company. It is developing a [Generation IV](https://en.wikipedia.org/wiki/Generation_IV_reactor \"Generation IV reactor\") high-temperature gas-cooled [pebble-bed](https://en.wikipedia.org/wiki/Pebble_bed_modular_reactor \"Pebble bed modular reactor\")[nuclear reactor](https://en.wikipedia.org/wiki/Nuclear_reactor \"Nuclear reactor\") design. It has received funding from private sources and various government grants and contracts, notably through the [Department of Energy](https://en.wikipedia.org/wiki/United_States_Department_of_Energy \"United States Department of Energy\")'s (DOE) Advanced Reactor Concept Cooperative Agreement in 2016 and its Advanced Reactor Demonstration Program (ARDP) in 2020.\n\nSee also\n--------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=19 \"Edit section: See also\")]\n\n*   ![Image 13](https://upload.wikimedia.org/wikipedia/commons/thumb/b/b5/Radioactive.svg/40px-Radioactive.svg.png)[Nuclear technology portal](https://en.wikipedia.org/wiki/Portal:Nuclear_technology \"Portal:Nuclear technology\")\n\n*   [Gas turbine modular helium reactor](https://en.wikipedia.org/wiki/Gas_turbine_modular_helium_reactor \"Gas turbine modular helium reactor\")\u2013 US/Russian design concept ~1997 - never built\n*   [Generation IV reactor](https://en.wikipedia.org/wiki/Generation_IV_reactor \"Generation IV reactor\")\u2013 New nuclear reactor technologies under development\n*   [Next Generation Nuclear Plant](https://en.wikipedia.org/wiki/Next_Generation_Nuclear_Plant \"Next Generation Nuclear Plant\")\u2013 Cancelled American reactor project\n*   [Very high temperature reactor](https://en.wikipedia.org/wiki/Very_high_temperature_reactor \"Very high temperature reactor\")\u2013 Type of nuclear reactor that operates at high temperatures as part of normal operation Pages displaying short descriptions of redirect targets\n*   [Nuclear fuel](https://en.wikipedia.org/wiki/Nuclear_fuel \"Nuclear fuel\")\n*   [Nuclear safety](https://en.wikipedia.org/wiki/Nuclear_safety \"Nuclear safety\")\n*   [Rainer Moormann](https://en.wikipedia.org/wiki/Rainer_Moormann \"Rainer Moormann\")\n\nReferences\n----------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=20 \"Edit section: References\")]\n\n1.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-1 \"Jump up\")**Williams, D.F. (March 24, 2006). [Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)](https://www.osti.gov/servlets/purl/885975-9IC4H7/) (Report). [doi](https://en.wikipedia.org/wiki/Doi_(identifier) \"Doi (identifier)\"):[10.2172/885975](https://doi.org/10.2172%2F885975).\n2.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-MIT_Kadak_2-0 \"Jump up\")**Kadak, A.C. (2005). [\"A future for nuclear energy: pebble bed reactors, Int. J. Critical Infrastructures, Vol. 1, No. 4, pp.330\u2013345\"](http://web.mit.edu/pebble-bed/papers1_files/Future%20for%20Nuclear%20Energy.pdf)(PDF).\n3.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-3 \"Jump up\")**Association of German Engineers (VDI), the Society for Energy Technologies (publ.) (1990). [_AVR - Experimental High-Temperature Reactor, 21 Years of Successful Operation for A Future Energy Technology_](http://www.nea.fr/abs/html/nea-1739.html). Association of German Engineers (VDI), The Society for Energy Technologies. pp.9\u201323. [ISBN](https://en.wikipedia.org/wiki/ISBN_(identifier) \"ISBN (identifier)\")[3-18-401015-5](https://en.wikipedia.org/wiki/Special:BookSources/3-18-401015-5 \"Special:BookSources/3-18-401015-5\").\n4.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-4 \"Jump up\")**[NGNP Point Design \u2013 Results of the Initial Neutronics and Thermal-Hydraulic Assessments During FY-03](https://www.inl.gov/technicalpublications/Search/Results.asp?ID=INEEL/EXT-03-00870)[Archived](https://web.archive.org/web/20060614093344/http://www.inl.gov/technicalpublications/Search/Results.asp?ID=INEEL%2FEXT-03-00870) 2006-06-14 at the [Wayback Machine](https://en.wikipedia.org/wiki/Wayback_Machine \"Wayback Machine\") pg 20\n5.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-5 \"Jump up\")**[\"Pebble Bed Modular Reactor - What is PBMR?\"](https://web.archive.org/web/20150503014959/https://www.hashdoc.com/documents/42279/the-frame-characteristics-of-jacketed-reactor). Archived from the original on May 3, 2015.\n6.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-6 \"Jump up\")**[\"How the PBMR Fueling System Works\"](https://web.archive.org/web/20080309163451/http://www.pbmr.com/download/FuelSystem.pdf)(PDF). Archived from [the original](http://www.pbmr.com/download/FuelSystem.pdf)(PDF) on March 9, 2008.\n7.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-ISR_7-0 \"Jump up\")**[[1]](http://www.fz-juelich.de/isr/2/tint-a_e.html)[Archived](https://web.archive.org/web/20060613225502/http://www.fz-juelich.de/isr/2/tint-a_e.html) June 13, 2006, at the [Wayback Machine](https://en.wikipedia.org/wiki/Wayback_Machine \"Wayback Machine\")\n8.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-8 \"Jump up\")**[Richard A. Muller](https://en.wikipedia.org/wiki/Richard_A._Muller \"Richard A. Muller\") (2008). [_Physics for Future Presidents_](http://www.google.com/products/catalog?q=physics+for+future+presidents&safe=on&um=1&ie=UTF-8&tbm=shop&cid=15719616496879467605&sa=X&ei=RkhOT7zhCtOUtwe1j5GmCA&ved=0CDQQ8wIwAQ). Norton Press. p.170. [ISBN](https://en.wikipedia.org/wiki/ISBN_(identifier) \"ISBN (identifier)\")[978-0-393-33711-2](https://en.wikipedia.org/wiki/Special:BookSources/978-0-393-33711-2 \"Special:BookSources/978-0-393-33711-2\").\n9.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-9 \"Jump up\")**[\"Fabrication of pyrolytic graphite rocket nozzle components\"](http://issuu.com/glass4/docs/the_cooling_jacketed_reactor/=html&identifier=AD0270153). _issuu.com_. Retrieved October 6, 2009.\n10.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-Key-diff-2002_10-0 \"Jump up\")**[\"_Key Differences in the Fabrication of US and German TRISO-COATED Particle Fuel, and their Implications on Fuel Performance_ Free, accessed 4/10/2008\"](https://web.archive.org/web/20040921163426/http://www.iaea.org/inis/aws/htgr/fulltext/htr2002_201.pdf)(PDF). Archived from [the original](http://www.iaea.org/inis/aws/htgr/fulltext/htr2002_201.pdf)(PDF) on September 21, 2004. Retrieved February 25, 2004.\n11.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-Key-diff-2003_11-0 \"Jump up\")**D. A. Petti; J. Buongiorno; J. T. Maki; R. R. Hobbins; G. K. Miller (2003). [\"Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance\"](https://digital.library.unt.edu/ark:/67531/metadc882322/). _Nuclear Engineering and Design_. **222** (2\u20133): 281\u2013297. [doi](https://en.wikipedia.org/wiki/Doi_(identifier) \"Doi (identifier)\"):[10.1016/S0029-5493(03)00033-5](https://doi.org/10.1016%2FS0029-5493%2803%2900033-5).\n12.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-12 \"Jump up\")**[\"NRC: Speech - 027 - \"Regulatory Perspectives on the Deployment of High Temperature Gas-Cooled Reactors in Electric and Non-Electric Energy Sectors\"\"](https://web.archive.org/web/20150503014959/https://www.hashdoc.com/documents/42279/the-frame-characteristics-of-jacketed-reactor). Archived from the original on May 3, 2015.\n13.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-13 \"Jump up\")**Rainer Moormann (2008). \"A safety re-evaluation of the AVR pebble bed reactor operation and its consequences for future HTR concepts\". Berichte des Forschungszentrums J\u00fclich. Forschungszentrum J\u00fclich, Zentralbibliothek, Verlag. [hdl](https://en.wikipedia.org/wiki/Hdl_(identifier) \"Hdl (identifier)\"):[2128/3136](https://hdl.handle.net/2128%2F3136). Berichte des Forschungszentrums J\u00fclich JUEL-4275.`{{cite journal}}`: Cite journal requires `|journal=` ([help](https://en.wikipedia.org/wiki/Help:CS1_errors#missing_periodical \"Help:CS1 errors\"))\n14.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-14 \"Jump up\")**Rainer Moormann (April 1, 2009). [\"PBR safety revisited\"](https://archive.today/20120530042951/http://www.neimagazine.com/story.asp?storyCode=2052589). Nuclear Engineering International. Archived from [the original](http://www.neimagazine.com/story.asp?storyCode=2052589) on May 30, 2012. Retrieved April 2, 2009.\n15.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-15 \"Jump up\")**Albert Koster (May 29, 2009). [\"Pebble Bed Reactor - Safety in perspective\"](https://web.archive.org/web/20100626054414/http://www.neimagazine.com/story.asp?sectioncode=76&storyCode=2053102). Nuclear Engineering International. Archived from [the original](http://www.neimagazine.com/story.asp?sectioncode=76&storyCode=2053102) on June 26, 2010.\n16.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-16 \"Jump up\")**[\"ORNL Review Vol. 36, No. 1, 2003 - Nuclear Power and Research Reactors\"](https://web.archive.org/web/20130701145044/http://www.ornl.gov/info/ornlreview/v36_1_03/article_01.shtml). Ornl.gov. Archived from [the original](http://www.ornl.gov/info/ornlreview/v36_1_03/article_01.shtml) on July 1, 2013. Retrieved September 5, 2013.\n17.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-17 \"Jump up\")**[\"E. Wahlen, J. Wahl, P. Pohl (AVR GmbH): Status of the AVR decommissioning project with special regard to the inspection of the core cavity for residual fuel. WM'00 Conference, February 27 - March 2, 2000, Tucson, AZ\"](http://www.wmsym.org/archives/2000/pdf/36/36-5.pdf)(PDF).\n18.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-18 \"Jump up\")**Der Spiegel (German news magazine), no. 24 (1986) p. 28\u201330\n19.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-19 \"Jump up\")**R. Baeumer, THTR-300 Erfahrungen mit einer fortschrittlichen Technologie, Atomwirtschaft, May 1989, p. 226.\n20.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-20 \"Jump up\")**[\"China leading world in next generation of nuclear plants\"](https://web.archive.org/web/20120211094120/http://daga.dhs.org/daga/readingroom/newsclips/2004/wto/41005scmp03.htm). _[South China Morning Post](https://en.wikipedia.org/wiki/South\\_China\\_Morning\\_Post \"South China Morning Post\")_. October 5, 2004. Archived from [the original](http://daga.dhs.org/daga/readingroom/newsclips/2004/wto/41005scmp03.htm) on February 11, 2012. Retrieved October 18, 2006.\n21.   ^ [Jump up to: _**a**_](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-HTR-PM-2021_21-0)[_**b**_](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-HTR-PM-2021_21-1)[\"China's HTR-PM reactor achieves first criticality: New Nuclear - World Nuclear News\"](https://www.world-nuclear-news.org/Articles/Chinas-HTR-PM-reactor-achieves-first-criticality). _www.world-nuclear-news.org_. Retrieved September 28, 2021.\n22.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-22 \"Jump up\")**Wang, Brian (December 13, 2023). [\"China's Pebble Bed Reactor Finally Starts Commercial Operation | NextBigFuture.com\"](https://www.nextbigfuture.com/2023/12/chinas-pebble-bed-reactor-finally-starts-commercial-operation.html). Retrieved December 15, 2023.\n23.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-23 \"Jump up\")**[\"South Africa: Energy and Environmental Issues\"](http://www.eia.doe.gov/cabs/safrenv.html). _EIA Country Analysis Briefs_. [Energy Information Administration](https://en.wikipedia.org/wiki/Energy_Information_Administration \"Energy Information Administration\"). [Archived](https://web.archive.org/web/20070204042807/http://www.eia.doe.gov/cabs/safrenv.html) from the original on February 4, 2007. Retrieved December 15, 2015.\n24.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-24 \"Jump up\")**[\"Earthlife Africa Sues for Public Power Giant's Nuclear Plans\"](http://www.ens-newswire.com/ens/jul2005/2005-07-04-03.asp). _[Environment News Service](https://en.wikipedia.org/wiki/Environment\\_News\\_Service \"Environment News Service\")_. July 4, 2005. Retrieved October 18, 2006.\n25.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-25 \"Jump up\")**Linda Ensor (September 17, 2010). [\"Hogan ends pebble bed reactor project\"](http://www.businessday.co.za/articles/Content.aspx?id=121307). Businessday.co.za. Retrieved September 5, 2013.\n26.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-26 \"Jump up\")**[\"HTMR-100 team aim for pebble bed SMR in South Africa: New Nuclear - World Nuclear News\"](https://www.world-nuclear-news.org/Articles/HTMR-100-team-aim-for-PBMR-SMR-in-South-Africa). _www.world-nuclear-news.org_. Retrieved June 24, 2023.\n27.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-Adams1994_27-0 \"Jump up\")**[US 5309492](https://worldwide.espacenet.com/textdoc?DB=EPODOC&IDX=US5309492), [Adams, Rodney M.](https://en.wikipedia.org/wiki/Adams_Atomic_Engines \"Adams Atomic Engines\"), \"Control for a closed cycle gas turbine system\", published May 3, 1994, issued May 13, 1993. Patent expired on 2006-05-03 due to failure to pay maintenance fees.[[2]](http://www.uspto.gov/web/offices/com/sol/og/2006/week26/patexpi.htm)\n28.   **[^](https://en.wikipedia.org/wiki/Pebble-bed_reactor#cite_ref-28 \"Jump up\")**[\"Company formerly known as Adams Atomic Engines\"](http://www.atomicengines.com/). Atomicengines.com. June 29, 2011. Retrieved September 5, 2013.\n\nExternal links\n--------------\n\n[[edit](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&action=edit&section=21 \"Edit section: External links\")]\n\n*   [IAEA HTGR Knowledge Base](http://www.iaea.org/inisnkm/nkm/aws/htgr/)\n*   AVR, experimental high-temperature reactor: 21 years of successful operation for a future energy technology [ISBN](https://en.wikipedia.org/wiki/ISBN_(identifier) \"ISBN (identifier)\")[3-18-401015-5](https://en.wikipedia.org/wiki/Special:BookSources/3-18-401015-5 \"Special:BookSources/3-18-401015-5\")\n*   [High Temperature Reactor 2006 Conference, Sandton, South Africa](https://web.archive.org/web/20060930044923/http://www.htr2006.co.za/)\n*   [MIT page on Modular Pebble Bed Reactor](http://web.mit.edu/pebble-bed/)\n*   [Research on innovative reactors in J\u00fclich](https://web.archive.org/web/20110702202227/http://www2.fz-juelich.de/ief/ief-6/research/nuclear-systems/innovative-reactors)\n*   [Differences in American and German TRISO-coated fuels](http://www.iaea.org/inis/aws/htgr/fulltext/htr2002_201.pdf)[Archived](https://web.archive.org/web/20040921163426/http://www.iaea.org/inis/aws/htgr/fulltext/htr2002_201.pdf) September 21, 2004, at the [Wayback Machine](https://en.wikipedia.org/wiki/Wayback_Machine \"Wayback Machine\")\n\nIdaho National Laboratory - United States\n*   [Conceptual Design of a Very High Temperature Pebble-Bed Reactor 2003](https://www.inl.gov/technicalpublications/Documents/2761726.pdf)\n*   [NGNP Point Design - Results of the Initial Neutronics and Thermal-Hydraulic Assessments During FY-03, Rev. 1](https://web.archive.org/web/20141025165808/http://www.inl.gov/technicalpublications/Documents/2699872.pdf), September 2003\n*   [Next Generation Nuclear Plant (NGNP) Project \u2013 Preliminary Assessment Of Two Possible Designs](https://www.inl.gov/technicalpublications/documents/2761728.pdf), March 21 \u2013 25, 2004\n*   [The Next Generation Nuclear Plant \u2013 Insights Gained from the INEEL Point Design Studies](https://www.inl.gov/technicalpublications/Documents/2761753.pdf), August 25 \u2013 September 3, 2004\n*   [Computation of Dancoff Factors for Fuel Elements Incorporating Randomly Packed TRISO Particles](https://www.inl.gov/technicalpublications/Documents/3028302.pdf), January 2005\n\nSouth Africa\n*   [Coalition Against Nuclear Energy South Africa](http://www.cane.org.za/category/pebble-bed-modular-nuclear-reactor/)\n*   [Eskom](https://en.wikipedia.org/wiki/Eskom \"Eskom\")\n*   [PBMR (Pty.) Ltd.](https://web.archive.org/web/20051030043434/http://www.pbmr.co.za/)\n*   [Pebble Bed Modular Reactor - PBMR - Home](http://www.pbmr.com/)\n*   [Atomic Energy in South Africa](https://web.archive.org/web/20040218025843/http://www.anti-atom.de/akwsudaf.htm)\n*   [Earthlife Africa: Nuclear Energy Costs the Earth campaign](https://web.archive.org/web/20050828122741/http://www.earthlife-ct.org.za/ct/article.php?story=20030630154051118)\n*   Steve Thomas (2005), [\"The Economic Impact of the Proposed Demonstration Plant for the Pebble Bed Modular Reactor Design\"](https://web.archive.org/web/20051018075200/http://www.psiru.org/reports/2005-09-E-PBMR.pdf), PSIRU, [University of Greenwich](https://en.wikipedia.org/wiki/University_of_Greenwich \"University of Greenwich\"), UK\n*   [NPR](https://en.wikipedia.org/wiki/NPR \"NPR\") (April 17, 2006) [NPR: South Africa Invests in Nuclear Power](https://www.npr.org/templates/story/story.php?storyId=5345501)\n\n| show * [v](https://en.wikipedia.org/wiki/Template:Nuclear_fission_reactors \"Template:Nuclear fission reactors\") * [t](https://en.wikipedia.org/wiki/Template_talk:Nuclear_fission_reactors \"Template talk:Nuclear fission reactors\") * [e](https://en.wikipedia.org/wiki/Special:EditPage/Template:Nuclear_fission_reactors \"Special:EditPage/Template:Nuclear fission reactors\") Types of [nuclear fission reactor](https://en.wikipedia.org/wiki/Nuclear_reactor#Fission \"Nuclear reactor\") |\n| --- |\n| **[Moderator](https://en.wikipedia.org/wiki/Neutron_moderator \"Neutron moderator\")** |\n| [Light water](https://en.wikipedia.org/wiki/Light-water_reactor \"Light-water reactor\") | * [Aqueous homogeneous](https://en.wikipedia.org/wiki/Aqueous_homogeneous_reactor \"Aqueous homogeneous reactor\") * [Boiling](https://en.wikipedia.org/wiki/Boiling_water_reactor \"Boiling water reactor\") * [BWR](https://en.wikipedia.org/wiki/GE_BWR \"GE BWR\") * [ABWR](https://en.wikipedia.org/wiki/Advanced_boiling_water_reactor \"Advanced boiling water reactor\") * [ESBWR](https://en.wikipedia.org/wiki/Economic_Simplified_Boiling_Water_Reactor \"Economic Simplified Boiling Water Reactor\") * [Kerena](https://en.wikipedia.org/wiki/Kerena_boiling_water_reactor \"Kerena boiling water reactor\") * [Natural fission](https://en.wikipedia.org/wiki/Natural_nuclear_fission_reactor \"Natural nuclear fission reactor\") * [Pressurized](https://en.wikipedia.org/wiki/Pressurized_water_reactor \"Pressurized water reactor\") * [AP1000](https://en.wikipedia.org/wiki/AP1000 \"AP1000\") * [APR-1400](https://en.wikipedia.org/wiki/APR-1400 \"APR-1400\") * [APR+](https://en.wikipedia.org/wiki/APR%2B \"APR+\") * [APWR](https://en.wikipedia.org/wiki/APWR \"APWR\") * [ATMEA1](https://en.wikipedia.org/wiki/ATMEA1 \"ATMEA1\") * [CAP1400](https://en.wikipedia.org/wiki/CAP1400 \"CAP1400\") * [CPR-1000](https://en.wikipedia.org/wiki/CPR-1000 \"CPR-1000\") * [EPR](https://en.wikipedia.org/wiki/EPR_(nuclear_reactor) \"EPR (nuclear reactor)\") * [HPR-1000](https://en.wikipedia.org/wiki/Hualong_One \"Hualong One\") * [ACPR1000](https://en.wikipedia.org/wiki/ACPR1000 \"ACPR1000\") * [ACP1000](https://en.wikipedia.org/wiki/ACP1000 \"ACP1000\") * [VVER](https://en.wikipedia.org/wiki/VVER \"VVER\") * [RITM-200](https://en.wikipedia.org/wiki/RITM-200 \"RITM-200\") * [KLT-40](https://en.wikipedia.org/wiki/KLT-40_reactor \"KLT-40 reactor\") * [OK-150/OK-900](https://en.wikipedia.org/wiki/OK-150_reactor \"OK-150 reactor\") * [OK-650](https://en.wikipedia.org/wiki/OK-650_reactor \"OK-650 reactor\") * [KN-3](https://en.wikipedia.org/wiki/KN-3_reactor \"KN-3 reactor\") * [VM](https://en.wikipedia.org/wiki/VM_reactor \"VM reactor\") * [IPWR-900](https://en.wikipedia.org/wiki/IPWR-900 \"IPWR-900\") * many others * [Supercritical (SCWR)](https://en.wikipedia.org/wiki/Supercritical_water_reactor \"Supercritical water reactor\") |\n| [Heavy water](https://en.wikipedia.org/wiki/Heavy-water_reactor \"Heavy-water reactor\") by [coolant](https://en.wikipedia.org/wiki/Nuclear_reactor_coolant \"Nuclear reactor coolant\") | | [D 2 O](https://en.wikipedia.org/wiki/Deuterium_oxide \"Deuterium oxide\") | * [Pressurized](https://en.wikipedia.org/wiki/Pressurized_heavy-water_reactor \"Pressurized heavy-water reactor\") * [CANDU](https://en.wikipedia.org/wiki/CANDU_reactor \"CANDU reactor\") * CANDU 6 * CANDU 9 * EC6 * AFCR * [ACR-1000](https://en.wikipedia.org/wiki/ACR-1000 \"ACR-1000\") * [CVTR](https://en.wikipedia.org/wiki/Carolinas%E2%80%93Virginia_Tube_Reactor \"Carolinas\u2013Virginia Tube Reactor\") * [IPHWR](https://en.wikipedia.org/wiki/IPHWR \"IPHWR\") * [IPHWR-220](https://en.wikipedia.org/wiki/IPHWR-220 \"IPHWR-220\") * [IPHWR-540](https://en.wikipedia.org/wiki/IPHWR#IPHWR-540 \"IPHWR\") * [IPHWR-700](https://en.wikipedia.org/wiki/IPHWR-700 \"IPHWR-700\") * [PHWR KWU](https://en.wikipedia.org/wiki/Nuclear_energy_in_Argentina \"Nuclear energy in Argentina\") * [MZFR](https://en.wikipedia.org/w/index.php?title=MZFR&action=edit&redlink=1 \"MZFR (page does not exist)\") * [R3](https://en.wikipedia.org/wiki/%C3%85gestaverket \"\u00c5gestaverket\") * [R4 Marviken](https://en.wikipedia.org/wiki/R4_nuclear_reactor \"R4 nuclear reactor\") | | --- | | [H 2 O](https://en.wikipedia.org/wiki/H2O \"H2O\") | * [HWLWR](https://en.wikipedia.org/w/index.php?title=HWLWR&action=edit&redlink=1 \"HWLWR (page does not exist)\") * [ATR](https://en.wikipedia.org/wiki/Fugen_Nuclear_Power_Plant \"Fugen Nuclear Power Plant\") * [HW BLWR 250](https://en.wikipedia.org/wiki/Gentilly_Nuclear_Generating_Station#Gentilly-1 \"Gentilly Nuclear Generating Station\") * [Steam-generating (SGHWR)](https://en.wikipedia.org/wiki/Steam-generating_heavy_water_reactor \"Steam-generating heavy water reactor\") * [AHWR](https://en.wikipedia.org/wiki/Advanced_heavy-water_reactor \"Advanced heavy-water reactor\") | | [Organic](https://en.wikipedia.org/wiki/Organic_matter \"Organic matter\") | * [WR-1](https://en.wikipedia.org/wiki/WR-1 \"WR-1\") | | [CO 2](https://en.wikipedia.org/wiki/Carbon_dioxide \"Carbon dioxide\") | * [HWGCR](https://en.wikipedia.org/w/index.php?title=HWGCR&action=edit&redlink=1 \"HWGCR (page does not exist)\") * [EL-4](https://en.wikipedia.org/wiki/Brennilis_Nuclear_Power_Plant \"Brennilis Nuclear Power Plant\") * [KKN](https://en.wikipedia.org/w/index.php?title=Kernkraftwerk_Niederaichbach&action=edit&redlink=1 \"Kernkraftwerk Niederaichbach (page does not exist)\") * [KS 150](https://en.wikipedia.org/wiki/KS_150 \"KS 150\") * [Lucens](https://en.wikipedia.org/wiki/Lucens_reactor \"Lucens reactor\") | |\n| [Graphite](https://en.wikipedia.org/wiki/Graphite-moderated_reactor \"Graphite-moderated reactor\") by [coolant](https://en.wikipedia.org/wiki/Nuclear_reactor_coolant \"Nuclear reactor coolant\") | | Water | | [H 2 O](https://en.wikipedia.org/wiki/H2O \"H2O\") | * [AM-1](https://en.wikipedia.org/wiki/Obninsk_Nuclear_Power_Plant \"Obninsk Nuclear Power Plant\") * [AMB-X](https://en.wikipedia.org/wiki/Beloyarsk_Nuclear_Power_Station#Early_reactors \"Beloyarsk Nuclear Power Station\") * [EGP-6](https://en.wikipedia.org/wiki/EGP-6 \"EGP-6\") * [RBMK](https://en.wikipedia.org/wiki/RBMK \"RBMK\") * [MKER](https://en.wikipedia.org/wiki/MKER \"MKER\") | | --- | | | --- | | [Gas](https://en.wikipedia.org/wiki/Gas-cooled_reactor \"Gas-cooled reactor\") | | [CO 2](https://en.wikipedia.org/wiki/Carbon_dioxide \"Carbon dioxide\") | * [_Uranium Naturel Graphite Gaz_ (UNGG)](https://en.wikipedia.org/wiki/UNGG_reactor \"UNGG reactor\") * [Magnox](https://en.wikipedia.org/wiki/Magnox \"Magnox\") * [Advanced gas-cooled (AGR)](https://en.wikipedia.org/wiki/Advanced_Gas-cooled_Reactor \"Advanced Gas-cooled Reactor\") | | --- | | [He](https://en.wikipedia.org/wiki/Helium \"Helium\") | * [GTMHR](https://en.wikipedia.org/wiki/Gas_turbine_modular_helium_reactor \"Gas turbine modular helium reactor\") * [MHR-T](https://en.wikipedia.org/w/index.php?title=MHR-T&action=edit&redlink=1 \"MHR-T (page does not exist)\") * [UHTREX](https://en.wikipedia.org/wiki/UHTREX \"UHTREX\") * [VHTR (HTGR)](https://en.wikipedia.org/wiki/Very-high-temperature_reactor \"Very-high-temperature reactor\") * PBR (PBMR) * [AVR](https://en.wikipedia.org/wiki/AVR_reactor \"AVR reactor\") * [HTR-10](https://en.wikipedia.org/wiki/HTR-10 \"HTR-10\") * [HTR-PM](https://en.wikipedia.org/wiki/HTR-PM \"HTR-PM\") * [THTR-300](https://en.wikipedia.org/wiki/THTR-300 \"THTR-300\") * [PMR](https://en.wikipedia.org/w/index.php?title=Prismatic_block_reactor&action=edit&redlink=1 \"Prismatic block reactor (page does not exist)\") | | | [Molten-salt](https://en.wikipedia.org/wiki/Molten_salt_reactor \"Molten salt reactor\") | | [Fluorides](https://en.wikipedia.org/wiki/FLiBe \"FLiBe\") | * [Fuji MSR](https://en.wikipedia.org/wiki/Fuji_Molten_Salt_Reactor \"Fuji Molten Salt Reactor\") * [Liquid-fluoride thorium reactor (LFTR)](https://en.wikipedia.org/wiki/Liquid_fluoride_thorium_reactor \"Liquid fluoride thorium reactor\") * [Molten-Salt Reactor Experiment (MSRE)](https://en.wikipedia.org/wiki/Molten-Salt_Reactor_Experiment \"Molten-Salt Reactor Experiment\") * [Integral Molten Salt Reactor (IMSR)](https://en.wikipedia.org/wiki/Integral_Molten_Salt_Reactor \"Integral Molten Salt Reactor\") * [TMSR-500](https://en.wikipedia.org/wiki/TMSR-500 \"TMSR-500\") * [TMSR-LF1](https://en.wikipedia.org/wiki/TMSR-LF1 \"TMSR-LF1\") | | --- | | |\n| None ([fast-neutron](https://en.wikipedia.org/wiki/Fast-neutron_reactor \"Fast-neutron reactor\")) | * [Breeder (FBR)](https://en.wikipedia.org/wiki/Fast_breeder_reactor \"Fast breeder reactor\") * [Integral (IFR)](https://en.wikipedia.org/wiki/Integral_fast_reactor \"Integral fast reactor\") * [Liquid-metal-cooled (LMFR)](https://en.wikipedia.org/wiki/Liquid_metal_cooled_reactor \"Liquid metal cooled reactor\") * [OK-550](https://en.wikipedia.org/wiki/OK-550_reactor \"OK-550 reactor\") * [BM-40A](https://en.wikipedia.org/wiki/BM-40A_reactor \"BM-40A reactor\") * [VT-1](https://en.wikipedia.org/wiki/VT-1_reactor \"VT-1 reactor\") * [Small sealed transportable autonomous (SSTAR)](https://en.wikipedia.org/wiki/Small,_sealed,_transportable,_autonomous_reactor \"Small, sealed, transportable, autonomous reactor\") * [Traveling-wave (TWR)](https://en.wikipedia.org/wiki/Traveling_wave_reactor \"Traveling wave reactor\") * [Energy Multiplier Module (EM2)](https://en.wikipedia.org/wiki/Energy_Multiplier_Module \"Energy Multiplier Module\") * [Reduced-moderation (RMWR)](https://en.wikipedia.org/wiki/Reduced_moderation_water_reactor \"Reduced moderation water reactor\") * [Fast Breeder Test Reactor (FBTR)](https://en.wikipedia.org/wiki/Fast_Breeder_Test_Reactor \"Fast Breeder Test Reactor\") * [Dual fluid reactor (DFR)](https://en.wikipedia.org/wiki/Dual_fluid_reactor \"Dual fluid reactor\") [Generation IV](https://en.wikipedia.org/wiki/Generation_IV_reactor \"Generation IV reactor\")* [Sodium (SFR)](https://en.wikipedia.org/wiki/Sodium-cooled_fast_reactor \"Sodium-cooled fast reactor\") * [BN-350](https://en.wikipedia.org/wiki/BN-350_reactor \"BN-350 reactor\") * [BN-600](https://en.wikipedia.org/wiki/BN-600_reactor \"BN-600 reactor\") * [BN-800](https://en.wikipedia.org/wiki/BN-800_reactor \"BN-800 reactor\") * [BN-1200](https://en.wikipedia.org/wiki/BN-1200_reactor \"BN-1200 reactor\") * [CFR-600](https://en.wikipedia.org/wiki/CFR-600 \"CFR-600\") * [Ph\u00e9nix](https://en.wikipedia.org/wiki/Ph%C3%A9nix \"Ph\u00e9nix\") * [Superph\u00e9nix](https://en.wikipedia.org/wiki/Superph%C3%A9nix \"Superph\u00e9nix\") * [PFBR](https://en.wikipedia.org/wiki/Prototype_Fast_Breeder_Reactor \"Prototype Fast Breeder Reactor\") * [FBR-600](https://en.wikipedia.org/wiki/FBR-600 \"FBR-600\") * [CEFR](https://en.wikipedia.org/wiki/China_Experimental_Fast_Reactor \"China Experimental Fast Reactor\") * [PFR](https://en.wikipedia.org/wiki/Dounreay#Prototype_Fast_Reactor_(PFR) \"Dounreay\") * [PRISM](https://en.wikipedia.org/wiki/PRISM_(reactor) \"PRISM (reactor)\") * [Lead](https://en.wikipedia.org/wiki/Lead-cooled_fast_reactor \"Lead-cooled fast reactor\") * [BREST-300](https://en.wikipedia.org/wiki/BREST_(reactor) \"BREST (reactor)\") * [Helium gas (GFR)](https://en.wikipedia.org/wiki/Gas-cooled_fast_reactor \"Gas-cooled fast reactor\") * [Stable Salt Reactor (SSR)](https://en.wikipedia.org/wiki/Stable_salt_reactor \"Stable salt reactor\") |\n| Others | * [Organic nuclear reactor](https://en.wikipedia.org/wiki/Organic_nuclear_reactor \"Organic nuclear reactor\") * [OMRE](https://en.wikipedia.org/wiki/Organic_Moderated_Reactor_Experiment \"Organic Moderated Reactor Experiment\") * [Arbus](https://en.wikipedia.org/w/index.php?title=Arbus-reactor&action=edit&redlink=1 \"Arbus-reactor (page does not exist)\") * [Piqua](https://en.wikipedia.org/wiki/Piqua_Nuclear_Generating_Station \"Piqua Nuclear Generating Station\") * [Aircraft Reactor Experiment](https://en.wikipedia.org/wiki/Aircraft_Nuclear_Propulsion \"Aircraft Nuclear Propulsion\") |\n| * ![Image 14](https://upload.wikimedia.org/wikipedia/commons/thumb/8/83/Symbol_template_class_pink.svg/20px-Symbol_template_class_pink.svg.png)[Nuclear fusion reactors](https://en.wikipedia.org/wiki/Template:Nuclear_fusion_reactors \"Template:Nuclear fusion reactors\") * ![Image 15](https://upload.wikimedia.org/wikipedia/en/thumb/d/db/Symbol_list_class.svg/20px-Symbol_list_class.svg.png)[List of nuclear reactors](https://en.wikipedia.org/wiki/List_of_nuclear_reactors \"List of nuclear reactors\") * ![Image 16](https://upload.wikimedia.org/wikipedia/commons/thumb/8/83/Symbol_template_class_pink.svg/20px-Symbol_template_class_pink.svg.png)[Nuclear technology](https://en.wikipedia.org/wiki/Template:Nuclear_technology \"Template:Nuclear technology\") * ![Image 17](https://upload.wikimedia.org/wikipedia/commons/thumb/8/83/Symbol_template_class_pink.svg/20px-Symbol_template_class_pink.svg.png)[Nuclear accidents](https://en.wikipedia.org/wiki/Template:Nuclear_and_radiation_accidents_and_incidents \"Template:Nuclear and radiation accidents and incidents\") |\n\n| [Authority control databases](https://en.wikipedia.org/wiki/Help:Authority_control \"Help:Authority control\"): National [![Image 18: Edit this at Wikidata](https://upload.wikimedia.org/wikipedia/en/thumb/8/8a/OOjs_UI_icon_edit-ltr-progressive.svg/20px-OOjs_UI_icon_edit-ltr-progressive.svg.png)](https://www.wikidata.org/wiki/Q1622414#identifiers \"Edit this at Wikidata\") | * [United States](https://id.loc.gov/authorities/sh85099093) * [Israel](https://www.nli.org.il/en/authorities/987007531423505171) |\n| --- |\n\n![Image 19](https://en.wikipedia.org/wiki/Special:CentralAutoLogin/start?type=1x1&usesul3=1)\n\nRetrieved from \"[https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&oldid=1279453464](https://en.wikipedia.org/w/index.php?title=Pebble-bed_reactor&oldid=1279453464)\"\n\n[Categories](https://en.wikipedia.org/wiki/Help:Category \"Help:Category\"): \n*   [Pebble bed reactors](https://en.wikipedia.org/wiki/Category:Pebble_bed_reactors \"Category:Pebble bed reactors\")\n*   [Nuclear power reactor types](https://en.wikipedia.org/wiki/Category:Nuclear_power_reactor_types \"Category:Nuclear power reactor types\")\n\nHidden categories: \n*   [Webarchive template wayback links](https://en.wikipedia.org/wiki/Category:Webarchive_template_wayback_links \"Category:Webarchive template wayback links\")\n*   [CS1: unfit URL](https://en.wikipedia.org/wiki/Category:CS1:_unfit_URL \"Category:CS1: unfit URL\")\n*   [CS1 errors: missing 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+        },
+        {
+            "title": "Is the graphite used in a nuclear reactor and the one used in ... - Quora",
+            "url": "https://www.quora.com/Is-the-graphite-used-in-a-nuclear-reactor-and-the-one-used-in-a-pencil-the-same",
+            "content": "No. The differences are the purity of the graphite used in reactors versus the adulterants added to pencils to adjust the softness,",
+            "raw_content": null
+        },
+        {
+            "title": "[PDF] A Safety Assessment of the Use of Graphite in Nuclear Reactors ...",
+            "url": "https://www.osti.gov/servlets/purl/6102304",
+            "content": "Concomitantly, coals, cokes and charcoals develop a porous white ash on the burning surfaces which greatly reduces radi- ation heat losses while simultaneously",
+            "raw_content": null
+        },
+        {
+            "title": "Why aren't graphite moderated reactors more popular? - Reddit",
+            "url": "https://www.reddit.com/r/nuclear/comments/w5ucmp/why_arent_graphite_moderated_reactors_more_popular/",
+            "content": "Graphite limits the lifetime a reactor can operate, it will eventually form cracks under neutron flux. Replacing it tends not to be worth the",
+            "raw_content": null
+        },
+        {
+            "title": "Review On the thermal oxidation of nuclear graphite relevant to high ...",
+            "url": "https://www.sciencedirect.com/science/article/abs/pii/S0022311522005785",
+            "content": "Thermal oxidation of nuclear graphite components is highly undesirable because it can cause structural and property degradation that negatively affect a",
+            "raw_content": null
+        }
+    ]
+}
\ No newline at end of file