Đang tải... (xem toàn văn)
Comprehensive nuclear materials 5 21 graphite Comprehensive nuclear materials 5 21 graphite Comprehensive nuclear materials 5 21 graphite Comprehensive nuclear materials 5 21 graphite Comprehensive nuclear materials 5 21 graphite Comprehensive nuclear materials 5 21 graphite
5.21 Graphite J Fachinger Furnaces Nuclear Applications Grenoble ZU Hanau Research and Development, Hanau, Germany ß 2012 Elsevier Ltd All rights reserved 5.21.1 Introduction 540 5.21.2 5.21.2.1 5.21.2.2 5.21.2.3 5.21.2.4 5.21.2.5 5.21.3 5.21.4 5.21.4.1 5.21.4.1.1 5.21.4.2 5.21.4.2.1 5.21.4.2.2 5.21.5 5.21.5.1 5.21.5.2 5.21.5.3 5.21.5.4 5.21.6 5.21.6.1 5.21.6.2 5.21.6.3 5.21.6.4 5.21.7 5.21.8 References Appendix Amounts of i-Graphite and Its Origin Russia United Kingdom France United States Others Retrieval of i-Graphite Graphite Properties Physical Properties Wigner energy Chemical Properties Oxidation in gaseous phases Graphite reactions with liquids Graphite Radioactivity Formation of 3H Formation of 14C Formation of 36Cl Diffusion of Radionuclides in Graphite Graphite Treatment for Disposal or Recycling Waste Packages and Encapsulation Thermal Treatment The Russian ‘Self-Propagating High-Temperature Synthesis SHS’ Recycling of i-Graphite Final Disposal Summary 540 540 541 541 541 542 542 542 542 543 543 543 545 545 546 546 547 549 550 550 551 552 553 553 556 557 558 Amounts of Irradiated Graphite in Different Countries Abbreviations AGR AM-1 AMB AVR b BEPO CP1 EDF EL2 FRJ-1 Advanced gas-cooled reactor Prototype of RBMK Aтoм Mиpный Бoльшoй Allgemeiner Versuchsreaktor (a small HTR prototype reactor in Germany) Barn (10À24 cmÀ2) British experimental pile ‘0’ Chicago pile-1 E´lectricite´ de France SA Graphite moderated test pile in France Research Reactor Juălich GLEEP Graphite low energy experimental pile HTGR High-temperature gas-cooled reactor HTR High-temperature reactor IAEA International Atomic Energy Agency MAGNOX Magnesium alloy graphite moderated gas cooled uranium oxide reactor RBMK Reaktor Bolschoi Moschtschnosti Kanalny THTR Thorium high-temperature reactor UNGG Uranium naturel graphite gaz reactor WAGR Windscale’s advanced gas-cooled reactor 539 540 Graphite 5.21.1 Introduction Graphite has been used in nuclear technology since the birth of this technology The first artificial nuclear reactor, the Chicago Pile-1, consisted of a pile of uranium and graphite It was the fundament for future developments in the different graphite-moderated nuclear power reactors such as the Uranium Naturel Graphite Gaz reactors (UNGG) in France, Magnox and advanced gas-cooled reactors (AGR) in the United Kingdom, or RBMK in Russia The culmination of this development was the high-temperature reactor, for example, the Fort Saint Vrain reactor in the United States or the Thorium-Hochtemperaturreaktor (THTR) in Germany Worldwide, more than 230 000 tons of irradiated graphite (i-graphite) exist, which will eventually need to be managed as radioactive waste.1 The major part of i-graphite is still in operational or shutdown nuclear power reactors Actually, the reactor cores have been removed from Fort Saint Vrain, GLEEP, and BEPO The removal of the core from the Allgemeiner Versuchsreaktor (AVR) is proposed for the next few years Smaller quantities of i-graphite result additionally from operation, in the form of graphite sleeves, which have to be replaced during operation, and from different kinds of research reactors The total amount of i-graphite is assumed to be in the range of 220 000–250 000 tons A more detailed overview of the amount of i-graphite is given in Section 5.21.2 Graphite changes its properties during irradiation in a nuclear reactor Most important for the treatment and disposal of i-graphite is the possibility of storing energy in the form of structural defects, the so-called Wigner energy The entire Wigner energy could be released rapidly after an initial local release Furthermore, the graphite will be contaminated by radionuclides They result from the activation of 13C and impurities in the graphite matrix as well as from the depletion of fission products released from the fuel elements This is described in Section 5.21.5 Various treatment methods have been developed or are under development to decontaminate i-graphite or to optimize the disposal volume and behavior, respectively Section 5.21.6 gives a short overview of all these developments A major issue is the establishment of a close graphite cycle, which is essential for the future development of graphite-moderated nuclear reactors The last section is dedicated to the final disposal options for i-graphite and the behavior of i-graphite under different disposal conditions 5.21.2 Amounts of i-Graphite and Its Origin Generally, one can distinguish between four different types of nuclear reactors that utilize graphite as neutron moderator and reflector Aircooled graphite piles with a low power density as test facilities, prototypes, and first-generation plutonium production reactors Carbon-dioxide-cooled reactors (Magnox and UNGG) for electricity supply and/or plutonium production Helium-cooled high-temperature reactors for electricity generation and process heat generation Graphite-moderated water-cooled reactors for plutonium production and/or electricity generation The last feature, electricity generation, has been optimized especially in the Russian RBMK reactors 5.21.2.1 Russia The main sources of i-graphite are the RBMK nuclear power plants as well as high-capacity plutonium production reactors Five RBMK power plants with 11 reactors are still in operation in Russia.2,3 The original license foresaw a lifetime of 30 years However, lifetime extensions are already licensed or envisaged Therefore, the first shutdowns are expected in 2013, with a replacement program starting in 2015.3 The amount of graphite from these reactors is given in Table A.1 An important fact about this graphite-moderated water-cooled reactor type is that a helium–nitrogen mixture gives the graphite moderator a protective atmosphere, which will have an important impact on the generation of 14C.4 Four more graphite-moderated power reactors, Bilibino-1–4, are in operation These dual purpose reactors for electricity and heat contain 133 tons of graphite each.3 The AMB-1 and -2 in Beloyarsk, and the AM-1, prototypes of the RMBK reactors, were shutdown in 1981, 1989, and 2004 respectively The fuel has been removed from the reactor core and stored in cooling basins The reactor units with the graphite core are under safe storage conditions of IAEA Stage I, under surveillance; further dismantling is planned Graphite Besides these graphite-moderated power plants, 13 high-capacity plutonium power reactors had been in operation in Russia All of them were shutdown between 1987 and 2008 The reactor units with the graphite core are under safe storage conditions of IAEA Stage II The first dismantling concepts proposed the transformation of the reactor shafts into a final radwaste repository However, this disposal concept is not in accordance with the new Russian waste disposal regulations.5 Therefore, a further 21 000 tons of graphite internals of the reactor units, as well as 8000 tons of graphite rings at on-site storage facilities, have to be managed as radioactive waste in Russia An additional source of i-graphite are research reactors and other critical assemblies The determination of the exact number of these reactors and assemblies is complicated However, it can be assumed that more than 110 such facilities exist in Russia Furthermore, it was not possible to figure out which moderator had been used in these facilities and therefore the amount of i-graphite could not be evaluated 5.21.2.2 The first UNGG were built and operated by CEA in Marcoule as plutonium production reactors for the French nuclear deterrent forces The main characteristics of these reactors are given in Table A.3 While the G1 was still air cooled, all other UNGG used CO2 as cooling gas The graphite bricks, used as moderator as well as shielding for the internal walls of the reactor, were mounted to a horizontal reactor core The decommissioning has achieved IAEA level Six more UNGG reactors have been built and operated by EDF for electricity production The main characteristics of these reactors are also given in Table A.3 All of them are under decommissioning With the exception of Chinon A1 the design was changed from a horizontal to a vertical shaft for improved fuel handling However, the fuel cartridges had to be protected by graphite sleeves to withstand the mechanical forces of the upper fuel cartridges These graphite sleeves are temporarily stored in silos except those from Buggy, which have been disposed at the final disposal sites at Manche and Aube United Kingdom The United Kingdom has the largest amount of i-graphite that has to be managed as radioactive waste1,6 because most of the British nuclear power plants are gas-cooled graphite-moderated reactors as opposed to those in other countries, which utilize water-moderated reactors as an alternative The Magnox reactor type was utilized after test reactor and prototype development in the late 1950s The name was derived from the fuel cladding made from a magnesium–aluminum alloy The last Magnox reactor was commissioned in 1971 The next generation of gas-cooled reactors were the AGR, commissioned between 1976 and 1989 Both reactor types were graphite-moderated and cooled with CO2 Totally about 80 000 tons of graphite have to be handled as radioactive waste now or in the next two decades after the shutdown of the still operational AGR reactors An overview of the amount of graphite in the different UK reactors is given in Table A.2 in the Appendix 5.21.2.3 541 France The first graphite-moderated reactor in France was the pile EL2 built at Saclay in 1952.7,8 It was an experimental reactor like the EL3 built in 1957 The total mass of graphite in these two reactors was 109 tons The operation of these reactors ended in 1965 and 1979, respectively 5.21.2.4 United States The development of nuclear reactors started in the United States with the graphite-moderated pile (CP1) in Chicago The largest amounts of i-graphite in the United States are from the plutonium production reactors at the Hanford site The Hanford B-Reactor was the first large-scale plutonium production reactor in the world The reactor was graphite moderated and water cooled It consisted of an 8.5  11 m horizontal tube and contained 1100 tons of graphite.9 All reactors at the Hanford site were intended for plutonium production In all, nine plutonium production reactors were operated at the Hanford site The reactors were shutdown between 1964 and 1971 after an average life span, except the lastbuilt N Reactor (1963), a dual purpose facility for civil electricity generation (shutdown in 1987) Most of the reactors have been entombed after defueling and act as interim storage for the graphite moderator and structural materials to allow the decay of radioactive material until dismantling is possible with a low dose risk Reactors exclusively for civil application were the HTGR at Peach Bottom and Fort Saint Vrain with a graphite block as moderator Both were shutdown after a relatively short operational time of and 13 years, respectively An overview of the amount of graphite in the different reactors is given in Table A.4 in the Appendix 542 Graphite 5.21.2.5 Others Table A.5 gives an overview of graphite-moderated reactors in other countries Most of them are RBMK reactors in countries of the former Soviet Union or Magnox reactors High-temperature reactors have been constructed in China, Germany, and Japan Two graphite-moderated high-temperature reactors, the AVR and the THTR, were operated in Germany About 1000 metric tons of i-graphite and irradiated carbon bricks have to be managed as radioactive waste Furthermore, about million irradiated fuel pebbles were produced during the operational time of these reactors They consist mainly of a graphite matrix that contains about 10 000 so-called coated particles (see Chapter 3.06, TRISO Fuel Production and Chapter 3.07, TRISO-Coated Particle Fuel Performance) These particles safely enclose the nuclear fuel and the major part of the fission products The AVR is under dismantling to the green field Therefore the reactor vessel with the graphite core will be pulled out as one piece and transferred to an interim storage facility It will stay there for about 80 years before further treatment The development in Japan is based on a graphite block core similar to that in the United States The Chinese HTR-10 is a pebble HTR like the one in Germany whole reactor core was flooded before cutting the graphite internals This procedure provides two advantages First, the water acts as shielding, which minimizes the dose rate of the employees, and second, the water prevents the formation of graphite dust However, water purification requires additional effort A third approach is being used for dismantling the AVR in Juălich, Germany The whole reactor vessel, including the graphite internals, built as one part, has been lifted out of the reactor and transferred to an interim storage Before lifting, the reactor core was filled with light concrete to consolidate the internals and to reduce impacts, in case the vessel falls down during the lifting procedure Further dismantling will be performed after the decay of the main part of g-emitters, especially 60Co This method represents a way between complete dismantling and long-term safe storage and enables fast cleanup of a site The choice of the best applicable retrieval method depends on several site-specific facts and therefore there is no ‘best procedure.’ They include the mechanical and physical properties of the graphite, the dose rate of the graphite, and the surrounding structures, as well as the specific side constructions, which determine the space available for the installation of equipment A detailed overview of retrieval and available procedures and tools is under compilation by an expert team of the European Carbowaste project.10 5.21.3 Retrieval of i-Graphite 5.21.4 Graphite Properties The retrieval of i-graphite is based on two main principles, dry and wet retrieval.10 Dry retrieval has been chosen for the decommissioning of WAGR and GLEEP GLEEP was a low-energy and low-radiation test reactor The resulting total activity was so low that the graphite could be removed manually without shielding Only protective overalls and gloves were required to avoid incorporation The graphite should be treated in an industrial incinerator that is licensed for the discharge of small amounts of radioactivity It was noted that the graphite blocks showed only small effects after treatment at 1400 K for h in air Less then 2% of the graphite was lost during the process However, about 87% of 3H and 63% of 14C were released from the graphite The activity and dose rate were so high that a manual retrieval of the graphite was not acceptable for the WAGR A remote removal system was developed for the retrieval of the graphite stack Wet retrieval was utilized for dismantling the hightemperature reactor at Fort Saint Vrain Therefore, the 5.21.4.1 Physical Properties The properties of graphite are related to its polycrystalline structure (see Chapter 2.10, Graphite: Properties and Characteristics) Graphite crystallites are built by graphite planes, which are loosely bound by van der Waal’s forces The single planes consist of carbon six rings with a sp2 electron configuration of the carbon atoms and a dislocated p-system on both sides of the plane Therefore the properties of the crystallites are anisotropic with respect to the orientation parallel or vertical to the planes Irradiation induces damages in these graphite crystallites, which lead to anisotropic effects in the crystallites A good example is a radiation-induced expansion in one direction and shrinkage in the other direction Therefore the macroscopic changes can be anisotropic, if the crystallites have an overall preferred orientation direction, or isotropic, if the crystallites are randomly distributed This depends on the shape of the crystallites as well as on the production process For example, Graphite extruded graphite shrinks parallel to the extrusion direction and expands perpendicular to the extrusion direction at temperatures below 300 C and shrinks in both directions at higher temperatures More isotropic molded graphite initially exhibits shrinkage in all directions under all irradiation conditions The irradiation-induced shrinkage proceeds to a point of maximal density Further irradiation causes an expansion to the original density and beyond Besides being important for reactor operation, this effect is also a key issue for waste management It affects the mechanical stability, which has a large influence on the retrieval of graphite piles from the reactor for decommissioning Furthermore, the density and porosity may influence the radionuclide release in intermediate storage and especially in final disposal Another important parameter for disposal is the reduction of the thermal conductivity Small amounts of fast neutrons will reduce thermal conductivity and can be decreased further by orders of magnitude to W mÀ1 KÀ1, depending on neutron dose and irradiation temperature Another very important property for disposal is porosity, which allows the penetration of aqueous phases into the graphite matrix and therefore an undisturbed transport of radionuclides through the graphite matrix 5.21.4.1.1 Wigner energy The Wigner effect is named after its discoverer Eugene Paul Wigner This effect describes the displacement of atoms in a solid caused by neutron irradiation, which can occur in any solid However, it has a special importance for solid moderator materials such as graphite An atom can be moved from its position in a crystal lattice by collision with neutrons, if they have energies above 25 eV Therefore, high-energy neutron, for example, MeV, causes cascades of damages with about 900 displacements in a graphite moderator Not all of the displacements lead to lattice defects because the displaced atoms could also fill lattice vacancies Atoms that cannot be placed in lattice vacancies remain as interstitial atoms between the lattice planes and therefore they are associated with a higher energy.11 When such an atom has sufficient thermal energy, it is able to move to normal lattice position and release excess energy if the position energy is higher than the energy required for the return to a lattice place If such a process has been initiated, the hole-stored Wigner energy can be released immediately and heat up a graphite pile This was the cause of the Windscale fire accident.12 543 Wigner energy can cause the following problems related to the management of i-graphite: Initiation of an uncontrolled release of Wigner energy during retrieval Sawing or cutting of the graphite core can lead to a local heat increase, which may lead to an uncontrolled release of Wigner energy Therefore such operations should be performed with sufficient cooling Release of the Wigner energy during final disposal The temperature of the final disposal sites for lowand medium-level wastes is normally strictly limited These limits depend on corrosion processes or microbial degradation, and higher temperatures may disturb the integrity of the disposal or increase the reaction rates of release mechanisms Therefore, the Wigner energy should be dissipated before storage by annealing the graphite at temperatures above 250 C or it has to be demonstrated that the disposal site will not be affected by such an energy release This has been tried by NIREX but they concluded that Wigner energy is not adequately understood to guarantee that a release of Wigner energy cannot affect the safety of a disposal site.13 Despite this potential risk, only low amounts of i-graphite have considerable amounts of Wigner energy They are related to reactors operated at low temperatures, for example, reflectors of material test reactors High reactor operation temperatures, for example, achieved in an AGR would directly cause the annealing of the graphite.13 5.21.4.2 Chemical Properties ‘Burning of radioactive graphite’ has been in public discussion since the accident at Chernobyl But graphite has an extremely low chemical reactivity, which explains its geochemical stability, proved by the presence of natural graphite ores in the earth’s crust Graphite needs extremely powerful oxidation agents to convert it into the thermodynamic-favored carbon dioxide This also allows the utilization of graphite under extreme conditions in industrial processes, for example, as electrode in arc melting at temperatures up to 3000 C or its use as fire extinguisher 5.21.4.2.1 Oxidation in gaseous phases Graphite can react with gases such as O2, CO2, H2O, and H2 at elevated temperatures and the temperature depends on the perfection of the graphite’s crystal structure14 and therefore on the amount of impurity 544 Graphite Generally, heterogeneous reactions involving a porous solid and a gas can be controlled by one or more of three idealized steps: Mass transport of the reacting gas from gas stream to the exterior graphite surface Mass transport of the reacting gas from the exterior surface to an active site and mass transport of the products in the opposite direction Chemical reaction at the active sites The variation of the reaction rate with temperature for gas–carbon reactions can be divided into three main zones (Figure 1) In the low-temperature zone (zone I), the reaction is controlled by the chemical reactivity of the solid (step 3) There will be almost no concentration gradient of reacting gases throughout the whole volume of the sample because of low reaction rate, and this provides uniform access to the interior surface of porous materials For graphite – oxygen reaction, the upper limit for temperature will be 500 C, and for graphite–steam system, it will be 850 C.15 In the intermediate-temperature zone (zone II), step becomes important The diffusion of reactants in pores will influence the oxidation rate of material At higher temperatures, the concentration gradient of the reacting gas becomes steeper within graphite and the gas concentration becomes zero at a distance R nearer the surface The activation energy Ea in this zone amounts to half of the true activation energy Et For graphite–steam reaction, this temperature region is characterized by a temperature range of 850–1350 C and graphite oxygen reaction, by 500–900 C In the high-temperature zone (>900 C for graphite oxygen and >1250–1400 C for graphite–water steam) – zone III – the concentration of the reacting gas is low at the exterior of the solid and the rate is controlled by step As bulk gas-transfer processes have low activation energies, the apparent activation energy for gas–carbon reactions in zone III is low The reactions occurring in the gas–graphite system are Reaction with oxygen ðÁr H : standard enthalpy of reaction at 25 Cị ẵ1 ẵ2 Ln (reaction rate) II a r H ẳ 283:0 kJmol1 COgị ỵ 1=2O2 gị!CO2 gị ẵ3 Reaction with carbon dioxide Boudouard reaction : Csị ỵ CO2 gị ! 2CO r H ẳ ỵ 172:5 kJ molÀ1 ½4 The equilibrium can be shifted with increasing CO pressure16,17 or in the presence of a catalyst Reaction with water CsịỵH2Ogị!COgịỵH2gị r H ẳỵ131:3 kJ mol1 ẵ5 COgị ỵ H2 Ogị ! CO2 gị ỵ H2 gị r H ¼ À213:7 kJ molÀ1 ½6 The hydrogen and CO2 produced can then react with carbon C ỵ H2gị ! CH4 gị b r H ẳ 110:5 kJmol1 Csị ỵ 1=2O2 gị!COgị C ỵ CO2gị ! 2COgị III r H ẳ 394:5 kJmol1 Csị ỵ O2 gị ! CO2 gị r H ẳ ỵ172:5 kJ mol1 ẵ4 r H ẳ 71:81 kJ molÀ1 ½7 The presence of hydrogen can shift reaction [5] left [4] Reaction with hydrogen I 1/ T Figure Ideal reaction zones in graphite: I – reaction rate is controlled by chemical reactivity of the sample; II – reaction rate is controlled by diffusion in pores; III – reaction rate is controlled by gas transport to the exterior surface of the sample; a and b are transition zones C ỵ H2gị ! CH4 gị r H ẳ 71:81 kJ mol ½7 The mechanism and kinetics of these reactions are described by Walker15 The approximate relative rates of gas–carbon reactions at 800 C and 0.1 atm are given in Table Graphite Table Approximate relative rates of gas–carbon reactions at 800 C and 0.1 atm pressure Reaction Relative rate C–O2 C–H2O C–CO2 C–H2  105 3  10À3 In the literature, there are a number of investigations of nuclear graphite reactivity in different oxidation conditions Results of oxidation of HTR-10 nuclear graphite IG-1118 exhibited three regimes: 400–600 C with an activation energy of 158.56 kJ molÀ1, 600–800 C, at which the activation energy was 72.05 kJ molÀ1, and the ‘third-zone’, over 800 C regime with a very low oxidation energy The comparison of reactivity of the two types of graphite used in HTR in oxygen and air at 650–900 C (regime II) leads to the conclusion that there is no difference in the behavior of matrix graphite (A3-27) and standard graphite V483T during oxidation.19 At the same time, at a lower temperature (400 C, regime I) matrix graphite is more reactive with respect to air For the temperature range 350–520 C, the activation energy Ea for A3-3 graphite matrix amounts to 110 kJ molÀ1.20 The oxidation in air and moisture has to be considered for dismantling and interim storage, whereas the reaction with humidity and aquatic phases is important for final disposal Several investigations into virgin and irradiated graphite have been carried out, mostly in air at ambient pressure A comprehensive review was made by Stairmand,21 who concluded that significant graphite oxidation can be excluded at temperatures below 350 C However, graphite oxidation in air can occur in high irradiation fields Duwe et al.22 showed the consumption of oxygen and the production of carbon dioxide during the interim storage of HTR fuel pebbles in sealed cans But the dose of the irradiation field from freshly irradiated fuel pebbles is normally not relevant to the interim storage or final disposal of i-graphite 5.21.4.2.2 Graphite reactions with liquids Graphite does not react with alkaline and acidic solutions if no oxidizing agent is present Dissolved oxidants such as nitric acid, ozone, hypochlorides, and hydrogen peroxide attack graphite to different degrees.23–26 An important factor is the surface area, which depends mainly on pore volume and pore size 545 The reaction with oxidizing agents, for example, concentrated nitric acid, leads finally to the evolution of carbon dioxide: C ỵ 4HNO3 ! 2H2 O ỵ 4NO2 ỵ CO2 However, different stable intermediate reaction products can be formed: graphitic oxide (C7H2O4), mellitic acid (C6(CO2H)6), and hydrocyanic acid (HCN) The yield of these products and carbon dioxide depends on the reaction conditions and the nature of the graphite material Contact of i-graphite with aqueous phases during interim storage or final disposal cannot be excluded with an absolute certainty In such a case, the oxidation of i-graphite depends mainly on the irradiationinduced production of highly reactive species by radiolysis of the aqueous phase and the accessible graphite surface Corrosion experiments with A3-3 graphite show that the corrosion rate of graphite is increased in final repository relevant aqueous phases by external g-irradiation with a dose rate of $2 kGy hÀ1.27 The obtained corrosion rates are in the range from 10À5 down to 10À7 g mÀ2 day High chloride concentrations accelerate the graphite corrosion probably by the formation of hypochlorides This clearly indicates that irradiation-induced corrosion processes are relevant to the final disposal of graphite However, this is an extremely high dose rate not relevant to the disposal of i-graphite The first attempt to determine the relation between the dose rate and the corrosion rate was made in the framework of the European RAPHAEL project.28 However, the low number of performed measurements and the scattering of the obtained data did not allow the derivation of a validated data set for such a correlation 5.21.5 Graphite Radioactivity The utilization of graphite in a reactor leads to two different types of radioactive contamination in the graphite material, the contaminants being Activated impurities in the bulk graphite material Radioactive isotopes occurring in the reactor circuit The activation products are more or less homogenously distributed in the graphite, depending on the original location of the impurities, as well as on the possibility of their migrating in the graphite by thermal gradients induced by the reactor conditions and repulse effects during the activation process itself The radioactive isotopes from the reactor 546 Graphite circuit are located (adsorbed) primarily at the graphite surface and migration into the bulk material requires a transport force, which could be a thermal gradient The depleted isotopes have different origins: Activation products of the coolant Impurities in the coolant Corrosion products from steel constructions of the reactor distributed in the coolant and activated in the reactor core Release of fission products from fuel elements with a cladding failure These different sources of the radioactive contamination indicate that the activities of i-graphite depend on the reactor type, the type of the utilized virgin graphite material, and the operational conditions of the reactor Therefore, even i-graphite of similar reactor types can show different contamination levels and different isotope ratios and a detailed characterization of i-graphite is required before retrieval from a specific reactor in addition to calculated radionuclide inventories A good example of such an approach was the compilation of the so-called Aktivitaătsatlas des AVR’29 which was calculated and validated on the basis of radiochemical analysis of i-graphite from different locations in the reactor core However, a detailed consideration of the different i-graphite materials from different reactor types and different graphite types will not be helpful Furthermore, the amount of detailed information would definitely be out of the scope of this review and not indicate the significant general problems of the waste management of i-graphite The dose rate, one of the key parameters for the retrieval and interim storage of i-graphite, depends mainly on the 60Co activity 60Co has a half-life of 5.3 years The main source of 60Co in i-graphite is the abrasion of fine metal parts from the pebble loop system, which has been built up in the pipes by neutron activation Therefore, waiting for some decay periods can be helpful to reduce the dose per person for the workers at dismantling Another important parameter for retrieval is the release of radionuclides into air This could occur in the form of contaminated dust, which can be handled by adequate exhausting methods More problematic is the release of tritium as gaseous component However, it must be ensured that information specific to the reactor is retained during this period For final disposal, 14C and 36Cl have been identified as key nuclides with respect to the long-term safety, due to their long half-life, mobility, and biocompatibility 5.21.5.1 Formation of 3H The radionuclide 3H, tritium has a half-life of 12.3 years The contribution of radioactivity in nuclear graphite resulting from tritium is significant.29–31 It is produced by the following reactions: Fission reactions of uranium impurities in the graphite and fuel cladding failure, such as 235 U(n,f) 3H reactions Lithium impurities in the graphite via 6Li(n,a) H reactions 3He (n,p) 3H in HTR reactors, which utilize helium as coolant 10B (n,2a) 3H reactions in absorber rods (negligible for designs without core rods) The chemical properties of tritium are essentially the same as those of ordinary hydrogen Tritium generated from lithium impurities is produced mostly in graphite bulk The release of tritium is controlled by its diffusion out of the grain boundaries and into the pore system 5.21.5.2 Formation of 14C Three routes, shown in Table 2, have to be considered for the formation of 14C In the reactor core materials, nitrogen is present only as an impurity, whereas carbon and oxygen are in some cases major constituent elements of the coolant, moderator, or fuel In spite of this fact, the 14N activation reaction is usually more important for 14C production due to its large cross-section Therefore, the location and the chemical form of nitrogen are important for the location of the formed 14C Nitrogen levels vary widely from 10 to 100 ppm in different reactor graphite types30 and sometimes they are not known very precisely A comprehensive study of 14C has been carried out by Marsden et al.31 Calculations showed that about 70% of the 14C originates from nitrogen impurities with an assumed amount of 50 ppm by weight Table Activation reactions generation 14C Reaction Abundance of isotope in natural element (%) Capture crosssection (barn) 14 99.63 1.07 0.04 1.81 0.0009 0.235 N(n,p)14C C(n,g)14C 17 O(n,a)14C 13 Graphite Another source of 14C is the oxygen pathway from the coolant A birth ratio of 14C has been calculated for an AGR from 17O:14N:13C as 25:21:1, assuming 50 ppm nitrogen Besides the level, the location of nitrogen impurity in reactor core materials is an important parameter The nitrogen content of graphite is reduced during manufacture by several high-temperature treatment steps However, the different heating and cooling processes cause the formation of cracks and closed pores, which could be filled with air Therefore, the absorption of nitrogen on graphite surfaces as well as the nitrogen diffusion in the graphite matrix is one of the major parameters for the local distribution of 14C in i-graphite Takahashi et al reported that the nitrogen content in graphite depends on the surface area of the graphite and decreases from the surface to a depth of about 30 nm32 and that 14C produced from nitrogen will remain at its original position This is in agreement with the 14C distribution of HTR fuel pebbles from the German AVR33 (Figure 2) and with the observation of a preferential release of 14C by surface oxidation of i-graphite from the German high-temperature reactor AVR in a nitrogen/steam atmosphere.36 Takahashi et al.32 reported that the kinetic energy of the formed 14C atom is about 470 kJ molÀ1, which is in the range of covalent carbon bonding energies and therefore the 14C atom will be attached to nodes of the carbon lattice They suggested that 14C will not be released by radiolytic oxidation of the graphite µCi g–1(C) 547 However, this is also a surface-related reaction and a similar release should be assumed as observed for thermal surface oxidation This is in contradiction to Finn reporting a backscattering energy of 40 keV ($4  106 kJ molÀ1) for 14C formation.37 This would be significantly above any chemical bonding energy and would lead to movements in the lattice and the formation of new species This high backscattering energy as well as the large number of displacements of carbon atoms during irradiation should lead to a more homogenous distribution of 14C However, the displacements are in the range of 1–2 mm so that 500 displacements in one direction would be required for a transport of mm So generally, it can be assumed that 14C produced by activation of 13C is more or less homogenously distributed as opposed to 14C generated from 14N which is located in near-surface areas However, it cannot be concluded in general that the 14C in i-graphite at the end of the reactor life is generated mainly by nitrogen activation Surface oxidation of i-graphite irradiated in carbon dioxide during reactor operation could reduce surface-bound 14 C This reaction, as well as low amounts of nitrogen impurities, could result in the remaining 14C inventory being generated mainly by activation of 13C Activation calculation for the Bugey reactor shows that the 13C activation leads to 96% of the measured inventory 14C in i-graphite and only 4% of the inventory must be addressed to nitrogen activation.38 This would also be in agreement with the results obtained by Pichon,39 which show a fast release of about 0.1% of the 14C inventory followed by a negligible leaching phase (Figure 3) A possible explanation could be a covalent bonding of 14C resulting from 13C activation in the graphite matrix and leachable 14C fraction from nitrogen activation loosely absorbed at the surface 5.21.5.3 10 20 30 mm Figure Distribution of 14C in an high-temperature reactor (HTR) pebble fuel element Adapted from Schmidt, P Alternativen zur Verminderung der C-14-Emission bei der Wiederaufarbeitung von HTR-Brennelementen; Forschungszentrum Juălich: Juălich, 1979 Formation of 36Cl The dominant formation of 36Cl is by the neutron activation of 39K (2.2 barn), the main stable natural chlorine isotope with an occurrence of about 75% Chlorine itself is used for the removal of metal impurities in graphite by the formation of volatile chlorides However, residual amounts of chlorine remain in the graphite Therefore, new cleaning methods for nuclear graphite avoid the utilization of chlorine Furthermore 36Cl can be built by n,a-reaction of 39 K (4.3 mbarn) or from 34S via an n,g-reaction to 35 S (2.3 barn) followed by a b-decay to 35Cl But these reactions have no significant relevance 548 Graphite Cumulative released fraction of 14C (%) 0.08 N Њ8 – Lime water 0.07 N Њ10 – Lime water N Њ9 – Pure water 0.06 0.05 0.04 0.03 0.02 0.01 0 50 100 150 200 250 300 Time (days) 350 400 450 500 Figure Leaching behavior of 14C from French G2 Adapted from Pichon, C.; Guy, C.; Comte, J Cl-36 and C-14 behaviour in UNGG graphite during leaching experiments, 2008 Fraction of data less than concentration 0.9 0.8 0.7 0.6 0.5 0.4 0.3 Measured by NAA Inferred from 36Cl 0.2 0.1 0 0.2 0.4 0.6 0.8 1.2 Initial chlorine concentration (ppm) Figure Initial chlorine concentration in Oldbury moderator graphite as measured by NAA and as inferred from 36 Cl activation Adapted from Brown, F.; Palmer, J.; Wood, P Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination In IAEA Technical Committee Meeting on Nuclear Graphite Waste Management, Manchester, UK, 1999 An investigation of core graphite from the Oldbury reactor shows the correlation of the initial chlorine impurities and the 36Cl inventory (Figure 4).40 Furthermore, the investigation shows chlorine loss during irradiation This is explained by the release of chlorine from open pores and an activation of chlorine in the closed pores However, radiolytic oxidation during operation will open the closed pores by graphite oxidation, which results in an additional release path for 36Cl (see Figure 5) Leaching experiments with French i-graphite from G2 showed that a major amount of 80–85% will be leached from the graphite in month A further 5–10% will be leached in a period of about 1½ years (Figure 6).39 This is in agreement with the proposed chloride form of 36Cl located at water-accessible surfaces and its high solubility A small part of 5–10% 36Cl remained in the graphite This could be explained by 36Cl located in graphite areas that are not in contact with the leaching media, for example, closed pores, or by covalent bonds between the chloride and the carbon atoms of the graphite lattice Further investigations are required to clarify the nature of the nonleachable 36Cl Graphite Closed porosity 35Cl Open porosity Oxidation Activation 549 Primary circuit Release 35Cl 35Cl Activation Oxidation 36Cl Release 36Cl 36Cl Figure Schematic of activation and release of chlorine in graphite Cumulative released fraction of 36Cl (%) 100 90 80 70 N Њ5 – Lime water N Њ10 – Lime water N Њ8 – Lime water 60 50 N Њ6 – Pure water N Њ9 – Pure water N Њ2 – Pure water 40 100 200 300 400 500 Time (days) Figure Leaching behavior of 36Cl from French G2 Adapted from Pichon, C.; Guy, C.; Comte, J Cl-36 and C-14 behaviour in UNGG graphite during leaching experiments, 2008 5.21.5.4 Diffusion of Radionuclides in Graphite Diffusion in polycrystalline graphite is a complex topic strongly related to the structure of the graphite The three general diffusion types, listed in the order of increasing diffusion velocity, are: Volume diffusion by movements of atoms due to the presence of lattice defects or exchange of lattice positions.41 Diffusion along grain boundaries Pore diffusion All the three different diffusion types can occur in graphite: Volume diffusion in the graphite crystallites, grain boundary diffusion at the micropores between the crystallites, and pore diffusion in the pores between the graphite particles Self-diffusion of carbon in graphite occurs at temperatures about 2000 C42 and may be important for central zones of HTR fuel elements Diffusion of fission products in graphite has been studied intensively with respect to radionuclide release from HTR fuel elements All these processes become effective at higher temperatures and can be neglected at temperature ranges relevant to retrieval, interim storage, and final disposal However, they might be interesting for decontamination processes, especially for tritium Table shows some diffusion coefficients measured for A3-3 graphite from HTR fuel pebbles, pitch coke (AS1-500), and petrol coke (AL2-500) after irradiation.43,44 The diffusion and release processes of radionuclides in i-graphite depend strongly on the nature of the graphite and especially on the anisotropy of the graphite.45,46 Tritium can be released from graphite more or less completely by thermal treatment under inert atmosphere at temperatures in the order of 1300 C.36 550 Graphite Table Diffusion coefficients of tritium nuclear graphite Type of graphite Temperature ( C) Diffusion coefficient, D0 (sÀ1) A3a A3a A3a A3b AS1-500b AL2-500b 800 850 900 1000 1050 1025 1.72  10À9 9.09  10À9 6.89  10À8 8.18  10À9 9.83  10À11 1.83  10À10 a Irradiation at temperatures < 100 C Irradiation at temperatures $500 C b 5.21.6 Graphite Treatment for Disposal or Recycling 5.21.6.1 Waste Packages and Encapsulation Containers or drums are used as a packaging option for i-graphite, mainly for safe handling in the operational phase of waste management and not as a barrier for long-term safety aspects No special designs of containers or drums have been made for i-graphite and standards from common waste management are applied Therefore, this aspect is not covered in this chapter However, it must be mentioned that graphite can act as a noble metal and accelerate the galvanic corrosion of stainless steel containers and measures should be implemented to isolate the graphite from the container or container materials with a guaranteed lifetime until final disposal should be used Another problem may arise while filling the waste container, especially if more or less rectangular graphite bricks are filled into drums The disposal costs depend on the classification and volume of the radioactive waste and not on the weight Therefore different methods have been developed to achieve a high packing density but all methods will generate secondary wastes in the form of graphite dust Bach et al compared grinding, plasma cutting, jet cutting, wire sawing, and hydraulic breaking of graphite, especially with respect to the related release of graphite dust The encapsulation aims at a safe enclosure of the waste by retardation or immobilization of radionuclides to avoid a release into the environment or at least to reduce the release to an unobjectionable level Generally two options exist for encapsulation Embedding in a matrix material Impregnation of the graphite to fill the open pores The reference waste package concept for graphite waste envisages the embedding of i-graphite in cement pastes The cement will establish an alkaline environment in the pore water, which will reduce the solubility of many key nuclides Especially 14C will form insoluble carbonates if it is oxidized to CO2 by radiolysis processes Further, the different cement phases combined with a large pore surface area will be able to absorb radionuclides or build insoluble secondary phases On the other hand, the porous structure will not hinder the contact between aquatic phases and the waste and therefore a radionuclide release cannot be excluded, especially for 36Cl, which shows no significant retardation by cement Alternative embedding materials could be glass, polymers or resins, bitumen, low-melting metals, or ceramics The organic materials would all result in a dense waste package that protects the graphite from leaching However, the production process and the handling are related to a potential fire hazard Furthermore, the long-term stability could be less than the half-life of the key nuclides due to radiolysis or ageing processes and therefore water exclusion cannot be guaranteed for the final disposal time scale Low-melting metals may have sufficient corrosion stability, which has not been sufficiently determined for disposal conditions However, their own toxicity may create a problem For example, the license for the German low-level waste underground disposal site Konrad allows only the disposal of 72 Mg tin, 539 Mg zinc and 33 400 Mg lead due to the water protection law of Lower Saxony The vitrification of graphite will result in a wellknown product similar to vitrified high-level waste Besides the known problems of the final disposal of high-level waste such as fracturing, the graphite may react with the glass melt and form dispersed gas pebbles (bubbles?), which is known from the embedding of coated particles from HTR fuel The closing of the open pore system of graphite has been successfully tested by impregnation with bitumen, epoxy resins, and tar Therefore, the graphite has to be evacuated and then immersed in bitumen or resin under high pressure at elevated temperatures to obtain a sufficient fluidity Leaching tests with such an impregnated material have proved the high immobilization potential of this procedure However, this would lead to problems similar to those already described for materials used as embedding A new approach to fill the pore system of i-graphite is a process that can be classified between embedding and impregnation It foresees the granulation of the Graphite i-graphite and a high-temperature compaction after mixing with glass in an amount equal to the pore volume First attempts show a density of about 2.2 cm3 gÀ1, which is close to the theoretical density, and that the glass does not increase the total volume Furthermore, this method would lead to volume-optimized waste packages because the produced block dimensions can be adjusted to the waste container dimensions However, the proposed good leaching resistance and mechanical properties are yet to be demonstrated 5.21.6.2 551 Centre, Juălich (former KFA Juălich) This development was related to the reprocessing of HTR fuel pebbles Another process, based on inductive heating, has been developed by Westinghouse for graphite fuel compacts However, the incineration of graphite would result in a total release of 14C as CO2 together with the bulk 12CO2, which may causes local increases of the 14C activity in the surrounding area of the incineration plant Therefore no public acceptance could be achieved for such a graphite treatment option, even if the released 14C activity would be negligible in comparison with the natural 14C amounts The trapping of CO2 is no alternative Solidification of the CO2 as insoluble calcium carbonate from 1.2 tons of graphite ($0.7 m3) would produce 10 tons CaCO3 (3.7 m3) However, such a process has the advantage that the 14 C has been transferred into a defined species and will have a more or less homogenous distribution An advanced thermal treatment method has been developed first at the Research Centre, Juălich It was shown that the majority of the carbon 14C inventory could be removed from the AVR reflector and fuel graphite and graphite from the thermal column of the research reactor FRJ-1 by partial oxidation.34 The AVR Graphite was irradiated at a high temperature in an inert helium atmosphere and the other graphite at room temperature in an air atmosphere The thermal treatment process for 14C separation was performed in nitrogen or argon plus 2% oxygen or humidified nitrogen or argon First examinations by Podruzhina showed a 14C release of about 70% with a total graphite oxidation in the range of 20 to 30%.34 This results were Thermal Treatment The most effective volume reduction would be the complete oxidation of the i-graphite with a small ash residue which contains the nonvolatile radionuclides Volatile radionuclides like tritium, 36Cl, or 137Cs may cause some problems but could be trapped from the off-gas and solidified for final disposal (Tritium in the form of HTO could be used for the production of cement paste used as embedding material for radioactive waste.) Another problem is the incineration of nuclear graphite due to its chemical purity The high thermal conductivity will conduct the heat from the surface into the bulk material, which inhibits incineration The poor combustibility of graphite was shown by the first attempt of CEA, utilizing a coal stove Therefore, the material must be crushed before incineration Milling can be performed technically without dust release but requires great effort The burning itself could be performed in furnaces or in fluidized bed reactors The burning of crushed graphite has been demonstrated at the Research 100 Release rate (%) 80 60 14 C release rate; nitrogen + steam Total carbon release rate; nitrogen + steam C release rate; nitrogen + 2% oxygen Total carbon release rate; nitrogen + 2% oxygen 14 40 20 0 Figure 50 100 150 200 Time (min) 250 300 350 C release and total carbon oxidation by thermal treatment of Allgemeiner Versuchsreaktor graphite at 1230 C 14 552 Graphite confirmed by Jansen.35 Higher release rates were obtained by Florjan.36 Up to 60% of 14C will be released within the first 60 followed by a slower release of 20–30% in the next 2–7 h (Figure 7) The best 14C release rates have been obtained at temperatures of about 1200 C, whereas the separation of 12C and 14C is better at lower temperatures (Tables and 5) But the release rates of Florjan could not be repeated until now Furthermore, these results show the different 14C release behavior of the different graphite types under similar treatment conditions The best results have been obtained with AVR graphite This is explained by the inhomogenous distribution of 14C with higher 14C concentrations on the surface and the existence of more reactive 14C containing species This indicates that the irradiation conditions have an important influence on this process and that further investigations will show whether this process can be applied to CO2-cooled reactors or RMBK reactors However, this process could be an alternative waste treatment option only when 5% of the graphite materials have to be oxidized and captured as CaCO3, if sufficient decontamination factors can be achieved with these graphite types The removal of 14C from graphite has been considered the main problem in decontaminating graphite However, the separation of radionuclides other than 14C has to be managed, which can be performed by different methods Table 5.21.6.3 The Russian ‘Self-Propagating High-Temperature Synthesis SHS’ Graphite is homogenously mixed aluminum and titanium dioxide The amounts are related to the following reaction: 3C ỵ 4Al ỵ TiO2 ! Al2 O3 ỵ 3TiC The exothermic reaction is self-propagating and only an initial start is required The formed stable 14 C release by thermal treatment in N2 + 2% O2 Sample R7 Origin AVR Treatment time (h) Temperature ( C) Total carbon release (%) 14 C release (%) Separation factor Table An option would be the complete incineration of the residual graphite, which would result in two more waste streams The residual 14C including the CO2 stream could be released in the environment if sufficient decontamination rates can be achieved or fixated as CaCO3, which can be sent to a surface disposal site The second waste stream will be a very small quantity of high active ashes and filter dust which must be disposed as high-level waste after an appropriate fixation A typical volume reduction would be in the range of 160 for an incineration process for nuclear graphite.1 A second option would be direct disposal in a nearsurface disposal site However, this would require a sufficient reduction of the 36Cl inventory (see Chapter 1.06, The Effects of Helium in Irradiated Structural Alloys), which has not been investigated yet R8 K8 K9 M3 M4 FRJ2 Reflector graphite Fuel graphite 900 2.85 61.0 21 900 2.69 62.2 23 1230 2.94 78.8 27 900 1.87 43.2 23 1230 2.32 64.4 28 K5 K6 M2 MS2 1230 4.04 79.8 20 Thermal column 14 C release by thermal treatment in N2 + steam Sample R6 Origin of graphite AVR Treatment time (h) Temperature ( C) Total carbon release (%) 14 C release (%) Separation factor R10 FRJ2 Reflector graphite Fuel graphite 900 0.85 41.0 48 900 1.48 70.0 47 900 1.55 45.0 29 Thermal column 1280 5.40 92.6 17 900 4.12 69.8 17 900 0.02 49.0 2250 Graphite titanium carbide contains 14C and the other radionuclides incorporated into the corundum and titanium carbide lattice Additional confining additives can be added to the reaction mixture, for example, zirconium, which build even more stable crystalline phases with selected radionuclides such as uranium and plutonium Furthermore, additives are used to improve the final product quality and to minimize the volatilization of radionuclides Therefore, this process is also suitable for graphite contaminated with actinides from the Russian production reactors The process requires a carefully controlled regime to minimize the radionuclide release 5.21.6.4 Recycling of i-Graphite The reuse of i-graphite may open a waste management route that has the potential to reduce the amount of i-graphite for disposal The easiest attempt would be the direct use of i-graphite without any treatment for the production of new materials for the nuclear industry Generally, it is known that used graphite can be recycled as additive material for graphite production However, this cannot be done with i-graphite in the existing graphite production facilities Even lowcontaminated graphite must be handled in a closed manufacturing unit to avoid the release of contaminated graphite dust Furthermore, the amount of i-graphite suitable for direct reuse is small in comparison with the total amount of i-graphite Therefore reuse will be associated with decontamination of i-graphite The success of the decontamination depends on the achievable decontamination factors, especially of 14C In principle, two methods could be proposed for decontamination: The wet method of leaching the graphite with suitable decontamination agents such as mineral acids or alkaline solutions Decontamination by thermal treatment in steam or diluted oxygen atmosphere Both options are under investigation in the European Carbowaste project.47 At the moment, there are not enough results from the leaching process to evaluate the feasibility of this method, whereas the thermal treatment mentioned in Section 5.21.6.2 has already proved its potential to remove 14C from the i-graphite matrix Figure shows potential routes to obtain feedstock material for graphite recycling Two options can be considered after thermal 553 14 C decontamination The first option is the total oxidation of the residual graphite to find out if the remaining 14C amount in the graphite would allow a free release of the off-gas into the atmosphere The expected residues are in the range of a few kilograms per Mg graphite The alternative is treatment with graphite cleaning methods known from the graphite industry to remove the residual nonvolatile radionuclides to a level that can be handled in further production steps Potential products could be silicon carbide, waste additives, and feedstock material for new graphite materials in the nuclear industry The production of materials will definitely be cheaper if fresh feedstock materials are used, but the benefit will be the reduced waste volume (see next section) Another interesting aspect of the separation of the 14 C is the option to replace 14C production as tracer material for scientific purposes by irradiation of nitrates 5.21.7 Final Disposal Figure shows the general routes for radioactive waste classification in different countries Among the European Union states, the Belgian and French schemes are very similar and are closely related to the EU classification scheme, which is based on the general IAEA recommendations These schemes formally recognize the lifetimes of the predominant radionuclides within waste packages, and segregate low- and intermediate-level waste into short-lived and long-lived categories, on the basis of whether the half-lives of these nuclides are less than or greater than 30 years respectively Generally i-graphite can be assumed to be low- or medium-level radioactive waste by these regulations, whereby the classification for final disposal of i-graphite is determined mainly by the inventory of long-living radionuclides 14C and 36Cl The high bio-compatibility and the good solubility of 36Cl if it occurs as chloride and therefore its high mobility require larger efforts to provide a safe enclosure from the environment The French surface disposal site Centre de l’Aube has a total limit for the disposal of 0.4 TBq for 36Cl Figure 10 shows a calculation of the 36Cl inventory of the stack of Bugey plus some measured data The graphite core of the reactor Bugey would have a 36Cl inventory of about 0.1 TBq, assuming an average level of 50 Bq 36Cl gÀ1, which is probably too low as shown in Figure 10 Measured values for Bugey reveal an 554 Graphite i-graphite Disposal Solidification 14 C depleted i-graphite Partial oxidation in steam or diluted oxygen 14 C enriched off-gas in the form of CO or CO2 14 C separation 14 C products Option Total oxidation Option Free release of the off-gas (depending on the residual 14C inventory) Reconversion of CO and CO2 off-gas to carbon Solidification and disposal Further graphite cleaning by high temperature treatment may be in presence of halogens as decontamination agent Other carbon-based products (e.g., SiC, lamp black, waste additives) Additive for new graphite products Figure Process scheme of a potential process for thermal treatment of i-graphite average of about 200 Bq gÀ1 Furthermore 36Cl is easily leached, which has been discussed in Section 5.21.5.3 Therefore, a near-surface disposal of graphite is not an acceptable waste management option in France The situation in the United Kingdom is similar An estimate of the total 36Cl inventory is given by David Lever.48 It is in the range of 23 TBq for the British i-graphite This 36Cl inventory will not allow the near-surface disposal at the Drigg site if the release could not be excluded over geological time scales A further aspect of UK reactors is the release of C if it is in the form of methane Figure 11 shows a risk analysis of such a release if all the 14C is released as methane The assumption will not be true for i-graphite; however, no quantitative results that give a clear figure about the relation between 14CO2, 14CO, and 14CH4 in the long-term release of 14C under disposal conditions are available However, there is some evidence that organic 14C compounds cannot be neglected Leaching of HTR fuel spheres shows a 14 Belgium France VLLW < 100 Bq g-1 Cal A – low concentrations short half-lives (Criteria X and Y) LLW Short-lived halflives < 30 years; activity between 100 and 105 Bq g-1 ILW Short-lived halflives < 30 years; activity between 105 and 108 Bq g-1 European Union Transition waste IAEA United Kingdom EW – Exempt waste VLLW – less than 400 kBq of b/g activity per 0.1 m3 material LILW-SL LILW-SL Short-lived half-lives < 30 years Short-lived half-lives < 30 years LLW Long-lived halflives > 30 years; Cal B – medium or activity between 100 LILW-LL long half-lives in and 105 Bq g-1 relatively high Long-lived half-lives ILW concentrations > 30 years Long-lived halfpower < 20 W m-3 lives > 30 years; LILW-LL Long-lived half-lives > 30 years activity between 105 and 108 Bq g-1 Cal C – substantial HLW; amounts of b- and activity between 108 a-emitters and 1010 Bq g-1 power > 20 W m-3 HLW HLW Figure Comparison of radioactive waste classification schemes Spent nuclear fuel Waste with negligible heat generation LLW – < GBq t-1 of a and 20 years nuclear weapons ILW >4 GBq t-1 of a or >12 GBq t-1 of b/g activity, no heating consideration in storage HLW as ILW and with cooling in storage facilities United States Uranium mill tailings Naturally occurring radioactive material Heat-generating waste Low-level radioactive waste (LLW): by definition: everything else Graphite Generic disposal routes Germany 555 556 Graphite 1.00E + 03 Bq g de 36Cl 1.00E + 02 Measures C2J6 C1K7 B319 C4H2 B8J0 C2L4 = C2G6 ABlB C3F9 D9Jl = A619 Risque = 50.00% Risque = 2.50% Risque = 0.10% Initial centre Initial + sygma 1.00E + 01 1.00E + 00 36Cl measurements Activation calculation results for each channel where the samples come from 1.00E – 01 11 13 15 17 19 21 Height (m) in the graphite stack (cooling gas flow direction) 23 Figure 10 Activation calculation results on BU1 stack with data assimilation method: 36Cl 1E – 02 1E – 03 1E – 04 Total Magnox spheres 1E – 05 1E – 06 1E – 07 Uranium spheres Stainless steel spheres Carbon steel spheres 1E – 08 Organic degradation 1E – 09 1E – 10 Radiolysis organics Release from graphite Risk target 1E – 11 1E – 12 1E – 13 1E + 00 1E + 01 1E + 02 1E + 03 1E + 04 1E + 05 1E + 06 Years postclosure (postclosure starts calendar year 2150) Figure 11 Radiological risk versus time for 14CH4 by contributory sources Adapted from Lever, D Graphite Wastes: Disposal Issues; Manchester, UK, 2006 higher release of organic 14C than 14CO2, but as dissolved organic species and not as gaseous species.49 In Germany, radioactive waste is divided into two classes: waste with and without significant heat development Deep underground disposal is planned for both types The former Konrad iron mine has been designated as the disposal site for the nonheat developing waste and is proposed to be ready for waste disposal in 2013 The graphite from the reactor core of the AVR and THTR clearly belongs to the category of non heat developing waste and therefore could be disposed of at this site However, the 14C inventory of about 2.9  1014 Bq for the ceramic core interior from the AVR will claim a major share of the licensed C inventory (4  1014 Bq) of this disposal site and will limit the disposal of other radioactive waste Furthermore, the actual interim storage stage will extend beyond the proposed operational time of this disposal site and therefore alternatives are required 14 5.21.8 Summary A general solution for the management of i-graphite has not been established yet Only France has an ambitious final disposal plan for its i-graphite, with the proposal of the near-surface underground disposal site at a depth of about 200 m Other countries like Graphite the United Kingdom have not made a final decision for a reference waste management route until now Three main challenges have been identified for the waste management of i-graphite The first is the retrieval of the major amounts of i-graphite from the reactor cores Some experience is available from the decommissioning of the BEPO, GLEEP, and Fort Saint Vrain reactors However, no general methodology can be recommended because the retrieval depends on many site-specific factors A major concern is the need for more data on i-graphite This includes data on property changes of the structure and mechanical properties due to irradiation and the radionuclide inventory, as well as fundamental data concerning the behavior of i-graphite during treatment procedures, and disposal behavior Future research should focus also on the speciation of the chemical form of radionuclides because the chemical form determines the long-term release behavior under final disposal conditions Furthermore, the localization of the key nuclides 14C and 36Cl at a nano-scale is a major challenge because a near-surface distribution and a homogenous distribution in the bulk would lead to completely different release characteristics Another challenge is the development of a safe waste management route Generally, two principal methodologies could be utilized: the decontamination of i-graphite by chemical or thermal treatment to obtain a carbonaceous material for further use in nuclear technology or the final disposal of i-graphite The most advanced plan for a final disposal has been achieved by France, which is planning an underground disposal site for low-level radioactive waste containing longliving radionuclides, especially 36Cl, 14C, and radium Therefore, i-graphite will be grouted in drums or containers and disposed of afterward This could be assumed as the actual reference concept Other conditioning methods, which ensure the safe long-term enclosure of 36Cl and 14C as the Russian RSH method or a long-term stable sealing of the graphite pore system, for example, with glass,50 may be alternatives to enable a near-surface disposal of i-graphite from reactor cores 10 11 12 13 14 15 16 17 18 19 20 21 22 23 References Marsden, B.; Wickham, A Characterization, Treatment and Conditioning of Radioactive Graphite from Decommissioning of Nuclear Reactors; IAEA: Vieanna, 2006 Energoatom Concern OJSC > Nuclear Power Plants Labyntseva, M Russian Federation Country Report; Marseilles, France, 2008 24 25 26 27 557 RBMK – Wikipedia Russian Standards Catalog GOST/R/50996/96/// Collection/storage/treatment/and/burial/of/radioactive/ waste/Terms/and/definitions/GOST/R/50996/96///Sbor/ khranenie/pererabotka/i/zakhoronenie/radioaktivnykh/ otkhodov/Terminy/i/opredeleniia/ Holt, G The decommissioning of commercial magnox gas cooled reactor power stations in the United Kingdom (Magnox Electric plc Berkeley Centre, Berkeley (United Kingdom)); IAEA-TECDOC–1043, pp 71–83 Mary Byrd Davis: Natural uranium graphite gas reactors (UNGG), Nuclear France: Materials and Sites, WISE-Paris, 2002 Bastien, D (CEA Centre d’Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France) )HTGR Knowledge Base: Twenty-nine years of French experience in operating gas-cooled reactors; http://www.iaea.org/inisnkm/nkm/ aws/htgr/abstracts/abst_iwggcr19_16.html Hanford Cultural and Historic Resources Program Hanford Site Historic District: History of the Plutonium Production Facilities, 1943–1990; Battelle: Columbus, OH, 2003 EC Project, Carbowaste, Contract FP7-211333, Workpackage & Wigner effect – Wikipedia, the free encyclopedia; http:// en.wikipedia.org/wiki/Wigner_energy, accessed on Mar 1, 2010 Windscale fire – Wikipedia, the free encyclopedia; http:// en.wikipedia.org/wiki/Windscale_fire, accessed on Mar 1, 2010 Wise, M Management of UKAEA Graphite Liabilities In Proceedings of the IAEA Technical Committee Meeting on Nuclear Graphite Waste Management; http://www.iaea org/inisnkm/nkm/aws/htgr/abstracts/abst_manchester html, 1999 Laine, N.; Vastola, F.; Walker, P J Phys Chem 1963, 67, 2030 Delle, W.; Koizlik, K.; Nickel, H Graphitische Werkstoffe fuăr den Einsatz in Kernreaktoren Teil 2: Polykristaliner Graphit und Brennelementmatrix; Thiemig AG: Muănchen, 1983 Hedden, K.; Loăwe, A Carbon 1967, 5, 339 Hedden, K.; Wicke, E About some influences on the reactivity of carbon, In Third Biennial Carbon Conference 1957, S249–256 Xiaowei, L.; Jean-Charles, R.; Suyuan, Y Nucl Eng Des 2004, 227, S281 Moormann, R.; Hinssen, H K.; Kuăhn, K Nucl Eng Des 2004, 227, 281 Katscher, W.; Moormann, R.; Hinssen, H K.; Stauch, B ersuche zur Graphitkorrosion in Luft bei Temperaturen unterhalb 873 K, Vorschlag zur Modifizierung des Verbrennungs-Head-Ends fuăr HTR-Brennelemente, Juălich, 1984 Stairmand, J Graphite Oxidation – A Literature Survey; AEA Technology Report, AEA-FUS-83, 1990 Duwe, R.; Bruăcher, H.; Fachinger, J R&D on Intermediate Storage and Final Disposal of Spent HTR Fuel; IAEA: Vienna, 1997 Gomez-Serrano, V.; Acedo-Ramos, M A.; Lopez-Peinado, A J Thermochim Acta 1995, 254, 249 Donnet, J B.; Schutz, A Carbon 1967, 5(2), 113–125 Donnet, J B.; Ehrburger, P Carbon 1973, 11(4), 309–316 Gmelin Handbuch der Annorganischen Chemie, Verlag Chemie GmbH, 1966 Fachinger, J.; den Exter, M.; Grambow, B.; Holgerson, S.; Landesmann, C.; Titov, M.; Podruhzina, T Behaviour of 558 28 29 30 31 32 33 34 35 36 37 Graphite spent HTR fuel elements in aquatic phases of repository host rock formations In Proceedings of the 2nd International Topical Meeting on High Temperature Reactor Technology, Bejing, China, 2004 EC Projekt ‘‘RAPHAEL’’; EC Contract 516508 (FI6O); Internal report D-B3.5 Aktivitaătsatlas des AVR, Juălich: WTI Bush, R.; Smith, G.; White, I F Carbon-14 Waste Management; Office for Official Publications of the European Communities: Luxemburg, 1984 Marsden, B.; Wickham, A.; Hopkinson, K The chemical form of Carbon-14 within graphite, Serco Assurance, 2002 Takahashi, R.; Toyahara, M.; Maruki, S.; Ueda, H.; Yamamoto, T Investigation of morphology and impurity of nuclear-grade graphite, and leaching mechanism of carbon-14 In Proceedings of IAEA Technical Meeting, Manchester, UK, on Nuclear Graphite Waste Management, published on CD-ROM IAEA-NGWM/CD 01-00120, October, 1999 Schmidt, P Alternativen zur Verminderung der C-14-Emission bei der Wiederaufarbeitung von HTR-Brennelementen; Forschungszentrum Juălich: Juălich, 1979 Podruzhina, T Graphite as radioactive waste: Corrosion behaviour under final repository conditions and thermal treatment, Forschungszentrum Juălich (Diss RWTH Aachen 2004), Berichte des Forschungszentrum Juălich 4166, 2004 R Jansen, Thermische Dekontamination von aktiviertem Reaktorgraphit – Planung, Konstruktion und Test eines Ofensystems Forschungszentrum Juălich (Dipl RWTH Aachen 2005), 2005 Florjan, M Dekontamination von Nuekleargraphit durch thermische Behandlung (im Druck), RWTH Aachen 2009 Finn, R.; Ache, J.; Wolf, A J Phys Chem 1969, 73(11), S3928–S3933 38 39 40 41 42 43 44 45 46 47 48 49 50 Poncet, B Personal communication Pichon, C.; Guy, C.; Comte, J Personal communication Brown, F.; Palmer, J.; Wood, P Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination In IAEA Technical Committee Meeting on Nuclear Graphite Waste Management, Manchester, UK 1999 Mrowec, S Acta Cryst 1981, A37, 141–142 Sach, R.; Williams, W Carbon 1974, S425 Fischer, P G Verhalten von Tritium in Reaktorgraphiten; Forschungszentrum Juălich GmbH: Juălich, 1975 Malka, V J.; Raitz von Frentz, R Adsorption und Desorption von Tritium an Graphit; Forschungszentrum Juălich GmbH: Juălich, 1978 Ashida, K.; Watanabe, K J Nucl Mater 1991, 183, 89–95 Saeki, M J Nucl Mater 1985, 131, 32 EC Project, Carbowaste, Contract FP7–211333, Workpackage Lever, D Graphite Wastes: Disposal Issues, Presentation at the University Manchester, UK, 2006; http://www.hse gov.uk/aboutus/meetings/iacs/nusac/031006/ presentation1.pdf Fachinger, J.; Zhang, Z X.; Brodda, B G Graphite corrosion and hydrogen release from HTR fuel elements in Q-brine In Proceedings of the International Conference on Radioactive Waste Management and Environmental Remediation, 5th, Berlin, Sept 3–7, 1995; Vol 1, pp 637–640 Fachinger, J.; Hrovat, M.; Grosse, K.; Seemann, R Impermeable graphite: A new development for embedding radioactive waste and an alternative option of managing irradiated graphite In Proceedings of the Waste Management Symposium 2010, 10027, Phoenix, 2010 Appendix Amounts of Irradiated Graphite in Different Countries Table A.1 Graphite-moderated reactors in Russia Reactor Graphite amount (Mg) Scheduled shutdown Kursk Kursk Kursk Kursk Leningrad Leningrad Leningrad Leningrad Smolensk Smolensk Smolensk Total 2000 2000 2000 2000 2638 1798 1798 1798 2158 1798 1798 19 988 2021 2024 2013 2015 2018 2020 2009 + 20 years 2011 + 20 years 2013 2020 Graphite Table A.2 559 Graphite-moderated reactors in United Kingdom Location Reactor Type Graphite in reactor (tons) Graphite total (tons) Dungeness Dungeness Hartlepool Hartlepool Heysham Heysham Heysham Heysham Hunterston Hunterston Hinkley Point Hinkley Point Torness Torness Bradwell Bradwell Calder Hall Calder Hall Calder Hall Calder Hall Chapelcross Chapelcross Chapelcross Chapelcross Dungeness Dungeness Hinkley Point Hinkley Point Oldbury Oldbury Sizewell Sizewell Wylfa Wylfa Berkeley Berkeley Hunterston Hunterston Trawsfynydd Trawsfynydd Windscale Winfrith Windscale Windscale Harwell Harwell Total B-1 B-2 Unit I-1 Unit I-2 Unit II-1 Unit II-2 B1 B2 B1 B2 Unit Unit Unit Unit Unit Unit Unit Unit Unit Unit A1 A2 A1 A2 Unit Unit A1 A2 A1 A2 Unit Unit A1 A2 Unit Unit WAGR Dragon Pile Pile BEPO GLEEP AGR AGR AGR AGR AGR AGR AGR AGR AGR AGR AGR AGR AGR AGR Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox Magnox AGR HTR Air cooled Air cooled Air cooled Air cooled 850 850 1360 1360 1520 1520 1520 1520 970 970 970 970 1520 1520 1810 1810 1164 1164 1164 1164 1164 1164 1164 1164 2150 2150 2210 3310 2061 2061 2237 2237 3470 3470 1938 1938 1780 1780 1900 1900 285 40