Comprehensive nuclear materials 3 02 nitride fuel

Comprehensive nuclear materials 3 02 nitride fuel Comprehensive nuclear materials 3 02 nitride fuel Comprehensive nuclear materials 3 02 nitride fuel Comprehensive nuclear materials 3 02 nitride fuel Comprehensive nuclear materials 3 02 nitride fuel Comprehensive nuclear materials 3 02 nitride fuel Comprehensive nuclear materials 3 02 nitride fuel

3.02 Nitride Fuel Y Arai Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Ibaraki, Japan ß 2012 Elsevier Ltd All rights reserved 3.02.1 Introduction 41 3.02.2 3.02.2.1 3.02.2.2 3.02.2.3 3.02.2.4 3.02.2.5 3.02.2.6 3.02.3 3.02.3.1 3.02.3.2 3.02.3.3 3.02.3.4 3.02.3.5 3.02.3.6 3.02.3.7 3.02.4 3.02.5 References Fabrication of Nitride Fuel Actinide Nitride Compounds Preparation from Metal or Hydride Carbothermic Reduction Other Nitride Formation Processes Nitride Pellet Fabrication Nitride Particle Fabrication Irradiation Behavior of Nitride Fuel Irradiation Experience Fuel Design Chemical Forms of FP Restructuring FP Gas Release Swelling and FCMI Fuel–Clad Chemical Interaction Reprocessing of Nitride Fuel Outlook of Nitride Fuel 43 43 44 44 45 46 47 47 47 47 48 49 50 50 51 51 52 53 Abbreviations ADS Accelerator-driven system CAPRA Fast reactor operated to burn rather than breed plutonium DP Direct pressing EPMA Electron probe microanalyzer FCCI Fuel–clad chemical interaction FCMI Fuel–clad mechanical interaction FIMA Fission per initial metal atom FP Fission products HLLW High-level radioactive liquid waste HLW High-level radioactive waste ITU European Institute for Transuranium Elements JAEA Japan Atomic Energy Agency LINEX Direct synthesis of actinide nitrides in the salt by the reaction with Li3N LOF Loss of flow MA Minor actinides MOX Uranium and plutonium mixed oxide PIE Postirradiation examination PSI Paul-Scherrer Institute PUREX Plutonium uranium recovery by extraction SPS TD TOP XRD Spark-plasma sintering Theoretical density Transient overpower X-ray diffraction 3.02.1 Introduction Nitride fuel has been proposed as an advanced fuel for fast reactors and developed since the 1960s in almost the whole nuclear world In this case, nitride fuel stands for a solid solution of uranium mononitride (UN) and plutonium mononitride (PuN), namely (U,Pu)N, in which the Pu/(U ỵ Pu) molar ratio roughly ranges from 0.15 to 0.25 In addition, UN was developed as a potential fuel for space reactors in the United States Although the interest in nitride fuel subsided under a global circumstance of slowdown of fast reactor programs in the 1980s, the solid solution of UN, PuN, and minor actinide (MA; Np, Am, and Cm) mononitride, (U,Pu,MA)N, has been proposed as one of the candidate fuels for Gen IV-type fast reactors Furthermore, as a 41 42 Nitride Fuel dedicated fuel for MA transmutation systems such as an accelerator-driven system (ADS), U-free nitride fuel, such as (Pu,MA)N diluted by ZrN, has been studied mainly in Japan At the beginning of the nuclear era, the development of fast reactor fuel cycles was centered on breeding ratio and doubling time The reason was that metallic fuel, the binary or certain ternary alloy of U and Pu, was adopted in the first generation of fast reactors The metallic fuel, however, had disadvantages for commercial use, such as anisotropic crystal structure, low melting temperature with phase transformations, and high fission product (FP) gas-induced swelling So a solid solution of uranium dioxide (UO2) and plutonium dioxide (PuO2), namely (U,Pu)O2 (MOX), has been a reference fuel for fast reactors and used in many test and prototype reactors all over the world, although the breeding ratio is smaller and the doubling time is longer than those of metallic fuel On the other hand, nitride fuel, as well as carbide fuel, has the advantages of both metallic fuel and oxide fuel as shown in Table It has a high thermal conductivity and high metal atom density like metallic fuel, while it has a high melting temperature and isotropic crystal structure like oxide fuel These characteristics led to the motivation for developing nitride fuel for fast reactors because the high thermal conductivity and high melting temperature allow a high linear power operation; alternatively, the largediameter fuel pins can be used for a given linear power The high metal atom density allows a low fissile material inventory with flexible core design and good neutron economics, leading to an improved breeding ratio and doubling time However, the development of nitride fuel has fallen behind that of carbide fuel, which has similar physical and chemical properties The reason includes an unexploited fuel fabrication process and Table Comparison of typical properties between oxide, metallic, and nitride fuels for fast reactors Chemical composition Theoretical density (TD) (g cmÀ3) Metal atom density (g cmÀ3) Thermal conductivity (W mÀ1 KÀ1) at 773 K at 1273 K Melting temperature (K) a the high neutron capture cross section of 14N (99.6% abundance in natural nitrogen) deteriorating neutron economics However, the fuel fabrication process has improved since the late 1980s, and the breeding ratio and doubling time have not been the center of the development of fast reactor fuel cycles Furthermore, it was found that nitride fuel is less hygroscopic in nature than carbide fuel, which will be advantageous for technological development It was also found that nitride fuel dissolves well in nitric acid without any formation of Pu oxalate, which will be compatible with hydrochemical reprocessing technology such as the PUREX process So since the late 1980s, the global interest has moved from carbide fuel to nitride fuel We can find two distinguished monographs about nitride fuel: one is written by Matzke1 published in 1986 and the other by Blank2 in 1994 These monographs describe nitride fuel and carbide fuel as MXtype fuel (X ¼ N or C) for fast reactors in detail from scientific and technological viewpoints It should also be mentioned that (U,Pu)N fuel with high Pu content, in which the Pu/(U ỵ Pu) molar ratio is roughly 0.45–0.55, was studied in France as a fast reactor fuel for incineration of Pu in the 1990s The good dissolution in nitric acid and stable crystal structure even at high Pu content led to the potential CAPRA core with (U,Pu)N fuel for incineration of Pu.3 Although not being described in this chapter, another interesting aspect of nitride fuel was pointed by Lyon et al.,4 who indicated the superior safety margin in case of hypothetical loss of flow (LOF) and transient overpower (TOP) events In a space reactor program called SP-100 in the United States, UN with highly enriched 235U was chosen as a reference fuel because it has the most favorable properties and will show the best performance for space reactor fuels.5 An extensive work was carried out in SP-100 program and Hayes et al At 0.1 MPa N2 pressure Oxide fuel Metallic fuel Nitride fuel (U0.8Pu0.2)O2 11.1 9.75 U–19Pu–10Zr (wt.%) 15.9 14.3 (U0.8Pu0.2)N 14.3 13.5 4.1 2.9 3083 18 31 1330 15 18 3053a Nitride Fuel summarized the physical,6 mechanical,7 transport,8 and thermodynamic properties9 of UN, while FP gas release and swelling of UN were summarized by Storms10 and Ross et al.,11 respectively Ross et al.12 also compiled and analyzed the thermal conductivity data of UN On the other hand, the diffusional and mechanical properties were reviewed by Routbort et al.13 previously Since the late 1990s, the partitioning and transmutation of MA has attracted global interest It may contribute to the decrease of toxicity of high-level radioactive waste (HLW) and the mitigation of burden for its final disposal Several transmutation systems and MA-containing fuels have been proposed so far Among them, the Japan Atomic Energy Agency ( JAEA) proposed a subcritical ADS as a transmutation system and MA nitride fuel as a dedicated fuel for transmutation.14 Besides the thermal and neutronic properties, the mutual solubility of actinide mononitrides in a wide range of composition and combination becomes an advantage of the fuel with high MA content Fabrication of MA nitride fuel and its property measurements have been carried out in JAEA.15–18 In this chapter, fabrication of nitride fuel and its irradiation behavior are summarized in Sections 3.02.2 and 3.02.3, respectively A brief description about reprocessing of spent nitride fuel is given in Section 3.02.4, because the reprocessing technologies are closely related with the specific issues of nitride fuel as 14C formation from natural nitrogen and 15N enrichment On the other hand, properties of nitride fuel are described in Chapter 2.03, Thermodynamic and Thermophysical Properties of the Actinide Nitrides In addition, an outlook of nitride fuel is briefly given in Section 3.02.5 43 Table Crystal structures and lattice parameters of nitrides of Th, U, Np, Pu, Am, and Cm Compounds Structure Lattice parameter (nm) ThN Th3N4 NaCl-type fcc Th3P4-type hexagonal 0.5167 a ¼ 0.3871 UN a-U2N3 þ x b-U2N3 À x NaCl-type fcc Mn2O3-type bcc La2O3-type hexagonal UN2 À x NpN PuN AmN CmN CaF2-type fcc NaCl-type fcc NaCl-type fcc NaCl-type fcc NaCl-type fcc N c ¼ 2.7385 0.4889 1.0685 a ¼ 0.3696 c ¼ 0.5840 0.531 0.4899 0.4905 0.4995 0.5027 U – Pu – N 1000 ЊC α-U2N3 + (U,Pu)N + N2 β + (U,Pu)N atm α-U2N3 β-U2N3 UN Solid U α + β + (U,Pu)N PuN α-U2N3 + (U,Pu)N Liquid Pu Figure Ternary U–Pu–N phase diagram at 1273 K Reproduced from Matzke, Hj Science of Advanced LMFBR Fuels; North-Holland: Amsterdam, 1986 3.02.2 Fabrication of Nitride Fuel 3.02.2.1 Actinide Nitride Compounds Although nitride fuel usually stands for a mononitride or its solid solution, such as UN and (U,Pu)N, higher nitrides other than mononitrides exist in the Th–N and U–N binary systems Table summarizes the crystal structures and lattice parameters of actinide nitrides reported in the Th–N, U–N, Np–N, Pu–N, Am–N, and Cm–N binary systems The binary U–N and Pu–N, and ternary U–Pu–N systems were investigated and reviewed by Holleck,19,20 Tagawa,21 and Potter22 in detail The ternary U–Pu–N phase diagram at 1273 K in Matzke’s monograph,1 originally calculated by Holleck,19 is shown in Figure The system is characterized by a complete solubility of UN and PuN It is considered that (U,Pu)N phase has a narrow composition range of the N/(U ỵ Pu) molar ratio Although Pu2N3 does not exist in the Pu–N system, a sesquinitride phase was identified in the U–Pu–N system at a Pu/(U ỵ Pu) molar ratio of 0.15.23 As seen in Table 2, actinide mononitrides have the same crystal structure with similar lattice parameters except for ThN, which leads to the mutual solubility In a mononitride lattice with NaCl-type structure, small nitrogen atoms are incorporated into a dense face-centered cubic packing of metal atoms Nitride Fuel 3.02.2.3 Carbothermic Reduction Carbothermic reduction is the most widely used process for preparing nitride fuel The starting material is a dioxide and carbon, and the general reaction is expressed as MO2 ỵ 2C ỵ 0:5N2 ẳ MN ỵ 2CO ẵI where M represents an actinide element, such as U and Pu The mixture of dioxide and carbon is heated in N2 gas stream, usually at 1773–1973 K It is considered that the carbothermic reduction could be applied in a technological production line as well as in a laboratory scale experiment, in contrast to the metal or hydride route.24 Furthermore, homogeneous products can be obtained by carbothermic reduction However, high amounts of oxygen, up to several thousand parts per million, are likely to remain in the products as impurity in case the initial carbon to dioxide mixing molar ratio, C/MO2, is 2.0 Therefore, an excess amount of carbon is usually added to the mixture to reduce the oxygen content and the residual carbon is removed from the products by heating in N2–H2 stream as CH(425) or HCN26 after carbothermic reduction The initial C/MO2 mixing ratio was chosen at 2.2–2.5 for the preparation of UN and (U,Pu)N Besides the two-step reaction constituted by the carbothermic reduction in N2 stream and the following decarburization in N2–H2 stream, a one-step reaction in N2–H2 or 60 61 62 63 64 2q (deg) (2 2) 65 PuN NpN AmN CmN PuN NpN Nitride preparation methods from metal or hydride were investigated mainly in the 1960s They include the nitridation of U or Pu metal in N2 or NH3 at about 1073–1173 K, arc-melting of U or Pu metal under N2 pressure, nitridation of fine grained U or Pu powder formed by the decomposition of hydrides with N2 or NH3 and direct reaction of UH3 or PuH2.7 with N2 or NH3 In the case of uranium nitrides, the products were often U2N3, which was subsequently decomposed to UN and N2 These reactions are exothermic and should be carried out slowly by temperature cycling for better control of the products Furthermore, these methods necessitate a high-purity inert gas atmosphere, since the fine-grained powders of metal, hydride, and nitride are chemically active and likely to react with moisture and oxygen in air even at room temperature So it is difficult to apply the metal or hydride route to a technological fuel production line and these methods were restricted to a laboratory scale experiment (3 1) Preparation from Metal or Hydride AmN 3.02.2.2 CmN 44 66 67 Figure X-ray diffraction pattern of (Np,Pu,Am,Cm)N prepared by carbothermic reduction.81 Reprinted with permission from OECD/NEA (2007), Actinide and Fission Product Partitioning and Transmutation, Ninth Information Exchange Meeting, Nıˆmes, France, Sept 25–29, 2006, p 119, www.nea.fr NH3 stream can be applied although a higher initial C/MO2 mixing ratio is necessary than that for the twostep reaction For the preparation of UN and (U,Pu)N, the atmosphere is changed to Ar or He from N2 or N2–H2 at a temperature lower than about 1673 K to prevent the formation of higher nitrides In the case of preparation of solid solution such as (U,Pu)N, both the reduction of the mixture of respective dioxides and the solid solution formation of respective mononitrides can be applied Figure shows the X-ray diffraction (XRD) pattern of (Np,Pu, Am,Cm)N prepared by the carbothermic reduction of the mixture of respective dioxides, from which the formation of quaternary mononitride solid solution was confirmed Mechanism and kinetics of carbothermic reduction were investigated by several authors, such as Muromura et al.,27–30 Lindemer,31 Greenhalgh32 and Bardelle et al.,26 mainly by chemical and XRD analyses, and weight change measurement for UN, PuN, and (U,Pu)N Muromura et al investigated the mechanism of carbothermic reduction at 1693–2023 K for UN in detail According to their results, the reaction is divided into four stages: (1) formation of UN1 À xCx from UO2, (2) decarburization of UN1 À xCx, (3) formation of UN1 À xCx with equilibrium composition, and (4) pure UN formation They also claimed that the carbothermic reduction followed the first-order rate reaction expressed as À lnð1 À aÞ ¼ kt ½1Š where a represents the reaction ratio, k the rate constant, and t the time, with an activation energy Nitride Fuel of 347 kJ molÀ1 This value is consistent with that reported by Greenhalgh,32 360 kJ molÀ1 On the other hand, Muromura et al claimed that the decarburization in N2–H2 or NH3 stream after carbothermic reduction followed the phase boundary-type rate reaction expressed as À ð1 À aị1=3 ẳ kt ẵ2 with activation energies of 285 kJ molÀ1 in 25% N2–75% H2 stream and 175–185 kJ molÀ1 in NH3 stream, respectively Kinetics was also investigated by thermogravimetry for (U,Pu)N33 and (U,Np)N.34 The results almost agreed with that for UN by Muromura et al.; the carbothermic reduction in N2 stream followed the first-order rate reaction with activation energies of 307 kJ molÀ1 for (U,Pu)N and 344–385 kJ molÀ1 for (U,Np)N Furthermore, the decarburization for (U,Np)N in 92% N2–8% H2 stream followed the phase boundary-type rate equation with an apparent activation energy of 210 kJ molÀ1 However, it should be pointed out that the decarburization includes both the removal of free carbon resulting in a decrease in weight and the replacement of carbon by nitrogen in carbonitride resulting in an increase in weight Typical impurities in nitride fuel prepared by carbothermic reduction are oxygen and carbon It was found that the level of impurities could be kept lower than 1000–2000 ppm for both oxygen and carbon by adjusting the initial C/MO2 mixing ratio Carbonitrides such as UN1 À xCx and PuN1 À xCx are characterized by complete solubility of the UN–UC and PuN–PuC systems, while solubility limits of hypothetical UO in UN and PuO in PuN were reported at 7% and 14%, respectively.35 It was reported that the carbon impurity content in mononitride prepared by carbothermic reduction is related to the thermodynamic equilibrium composition of carbonitride with free carbon under nitrogen atmosphere.17 When the same condition of carbothermic reduction was applied for UN, NpN, and PuN, the carbon impurity content decreased with the increase of atomic number of actinides Indeed, a rather high initial C/MO2 mixing ratio was chosen for the preparation of AmN and (Pu,Cm)N,36,37 since the monocarbides of Am or Cm are thermodynamically unstable It is well known that Am-bearing species have high vapor pressures in comparison with the other actinides Vaporization of Am during fuel fabrication process should be kept as low as possible In the case of 45 preparation of Am-bearing nitrides by the two-step reaction, the carbothermic reduction in N2 stream was carried out at 1573 K, which was lower than the cases for UN and (U,Pu)N by about 200 K.36,38 Then the temperature was raised to 1773 K for the decarburization in N2–H2 stream It is considered that the intermediate product of AmCO is likely to vaporize congruently during the carbothermic reduction On the other hand, the vaporization of Pu during carbothermic reduction can be neglected, which is different from the preparation of Pu-bearing carbides by carbothermic reduction carried out in vacuum 3.02.2.4 Other Nitride Formation Processes Four processes were reported for the preparation of nitride with regard to pyrochemical reprocessing of spent fuel The first one is the direct dissolution of spent nitride fuel in liquid Sn, followed by the pressurization with N2 It was reported that UN powder with high density sank to the bottom and could be mechanically separated from the liquid phase.39 The second and third processes concern the nitridation of actinides recovered in liquid Cd cathode by molten salt electrolysis The second one is the nitridation by N2 gas bubbling, in which N2 gas is passed into liquid Cd phase at 773–823 K Kasai et al reported that they succeeded in preparing UN or U2N3 granules by the N2 gas bubbling method.40 It was found, however, that the method was not applicable to the nitridation of Pu in liquid Cd because of the thermodynamic stabilization of Pu in liquid Cd phase.41 On the other hand, the third one is the nitridation–distillation combined reaction, in which the liquid Cd cathode-containing actinides are heated in N2 stream at 973 K In this method, the nitridation of actinides and distillation of Cd proceed simultaneously Preparation of (U,Pu)N, PuN, and AmN has been reported so far by the nitridation– distillation combined method.41,42 The fourth one is called LINEX process, in which actinides dissolved in the chloride molten salt are converted to nitride by the direct reaction with Li3N.43 In addition, a new process was reported by Yeamans et al.44 They successfully synthesized UN from UO2 by making it react first with NH3(HF)2 at ambient temperature to form (NH4)4UF8, and then with NH3 at 1073 K to UN2, followed by the decomposition to UN at 1373 K in Ar This method has the advantage of a low-temperature operation in comparison with the carbothermic reduction of dioxides 46 Nitride Fuel 3.02.2.5 Nitride Pellet Fabrication Nitride fuel pellets are usually prepared by a classical powder metallurgical manner; the product of carbothermic reduction is ground to powder by use of a ball mill, pressed into green pellets and sintered in a furnace at 1923–2023 K An organic binder is sometimes added to the ground powder to facilitate the pressing Finally, the diameter of sintered pellets is adjusted by use of a centerless grinder As is mentioned later, one of the characteristics of nitride fuel is that both He- and Na-bonded pins can be applied In general, an He-bonded fuel pin is characterized by low-density pellets (i.e., 80–85% of theoretical density (TD)) and a small gap width between pellets and cladding tube, whereas a Na-bonded fuel pin is characterized by high-density pellets (i.e., >90% TD) and a large gap width Actinide nitride powder has a low sinter-ability in comparison with that of oxide or carbide powder, which is derived from a low diffusion rate of metal atoms in mononitrides So a rather high sintering temperature (i.e., T >1973 K) is necessary for preparing dense UN or (U,Pu)N pellets higher than 90% TD.45 Although a small amount of Ni powder is an effective sintering aid for carbide fuel, it is not applicable to nitride fuel On the other hand, Bernard et al., reported that oxygen impurities tend to promote the sintering of (U,Pu)N pellets.24 However, the increase of oxygen impurities in UN and (U,Pu)N up to wt% resulted in the decrease of density and grain size of sintered pellets.46 Microstructures of (U,Pu)N pellets with different oxygen impurity contents are shown in Figure Sintering atmosphere also affects the sintered density of nitride fuel pellets It was reported that sintering in high N2 partial pressure, such as in N2 or N2–H2 stream, resulted in lower density than sintering in low N2 partial pressure, such as in Ar or Ar–H2 stream.45,47 This is an opposite tendency of the self-diffusion coefficient of Pu in (U,Pu)N at different N2 partial pressures.48 The residual oxygen impurity contents might affect the density of pellets sintered in different atmospheres On the other hand, sintering in N2 or N2–H2 stream is indispensable for Am-bearing nitride pellets from the viewpoint of mitigating loss of Am by evaporation It was reported that the density higher than 85% TD was attained for (Np,Am)N and (Pu,Am)N pellets by sintering in N2–H2 stream at temperatures lower than 1953 K.49 In addition to the classical powder metallurgical manner, a direct pressing (DP) method was proposed by Richter et al.50 In this method, the nitride compacts after carbothermic reduction were not ground to powder but directly pressed into green pellets, followed by sintering under the conventional manner The DP method has the advantage of avoiding dust production and shortening preparation period The (U,Pu)N pellets prepared by the DP method had a density of about 83% TD with levels of oxygen and carbon impurities lower than 0.1 wt%.51 The open porosity predominated in the pellets prepared by the DP method An isostatic hot-pressing technique was applied to fabrication of dense UN specimens for thermal and mechanical property measurements Speidel et al prepared UN pellets higher than 95% TD by consolidating the powder sealed in a refractory metal container under a pressure of 6.9 MPa at 1753–1813 K.52 Furthermore, a spark-plasma sintering (SPS) method for nitride fuel has been applied to preparation in a laboratory scale experiment recently.53 The SPS Oxide 20 μm (U,Pu)N pellet containing 0.21 wt% oxygen 20 μm (U,Pu)N pellet containing 0.99 wt% oxygen Figure Microstructures of (U,Pu)N pellets with different oxygen impurity contents Reproduced from Arai, Y.; Morihira, M.; Ohmichi, T J Nucl Mater 1993, 202, 70–78 Nitride Fuel method is a kind of pressure-assisted sintering that utilizes an electric current The method has the advantage of obtaining dense pellets at a drastically lower sintering temperature and a shorter sintering time than those of the conventional methods 3.02.2.6 Nitride Particle Fabrication Nitride particle fabrication method was vigorously developed in the Paul-Scherrer Institute (PSI) of Switzerland,54,55 then followed by India56 and Japan.57 The starting material is usually a nitric solution of actinides and this method has the advantage of avoiding dust production and feasibility of remote operation in comparison with the conventional powder process The nitride particles prepared may be directly filled into fuel pin (sphere-pac fuel) or pressed and sintered to fuel pellets The production of microspheres is carried out by a so-called sol–gel process The feed solution is mixed with an aqueous solution of gelation agent, urea, dispersed carbon black, and surfactants Different size of microspheres can be obtained by changing the nozzle used for microspheres production Besides the external gelation process using gelation agent, the internal gelation process developed by PSI consists of falling the droplets of feed material into hot silicon oil for microspheres production After washing, drying and calcining to MO2 ỵ C microspheres, they are subjected to carbothermic reduction In the case of preparing sphere-pac fuels, the carbothermic reduction is carried out at higher temperature than the conventional powder process to obtain dense nitride particles by reaction sintering The sol–gel process is proposed for the preparation of nitride fuel for the transmutation of MA under the double-strata fuel cycle concept.14 In this concept, MA partitioned from high-level liquid waste (HLLW) in a reprocessing plant is converted to nitride microspheres by the sol–gel process and carbothermic reduction, followed by mixing with diluent materials and sintering for pellet preparation Table 47 3.02.3 Irradiation Behavior of Nitride Fuel 3.02.3.1 Irradiation Experience The irradiation experience of nitride fuel is rather limited in comparison with the other fuels for fast reactors, such as oxide, metallic, and carbide fuels Especially, the number of (U,Pu)N fuel pins irradiated in fast reactors so far is smaller than 200 all over the world, which is summarized in Table The highest burnup was attained in the irradiation test in the EBR-II fast reactor, but still lower than 10% of fission per initial metal atom (FIMA).58 On the other hand, high burnups, that is, >15% FIMA, were attained in thermal reactors, such as ETR in the United States62 and HFR in the Netherlands.63 Most of them were irradiated in instrumented capsules In the United States, following the capsule irradiation in ETR and EBR-II, subassemblies constituted by 57 (U,Pu)N fuel pins were irradiated in EBR-II,64 whereas in Europe, more than 10 (U,Pu)N fuel pins were irradiated in fast test reactors, such as DFR, RAPSODIE, and PHENIX.59,60 Besides, in Japan, two (U,Pu)N fuel pins were irradiated in fast test reactor JOYO.61 With regard to nitride fuel other than (U,Pu)N, five subassemblies of 235U-enriched UN fuel were irradiated to about 9% FIMA in BR-10 in the 1980s.65 In addition, nitride fuels for the transmutation of MA have been subjected to the irradiation tests recently Besides (U,Pu,Np,Am)N and (Pu,Am, Zr)N fuels irradiated in PHENIX,66 (Pu,Zr)N fuels were irradiated in Russia67 and Japan.68 3.02.3.2 Fuel Design There are two typical bonding concepts of (U,Pu)N fuel pins for fast reactors: one is Na bonding and the other is He bonding Since (U,Pu)N fuel is compatible with liquid Na at operating temperatures, the gap between fuel pellets and cladding tube can be filled with liquid Na as well as gaseous He In a sense of Irradiation tests of (U,Pu)N fuel carried out in fast reactors Reactor Bonding Max linear power (kW mÀ1) Max burnup (% FIMA) References EBR-II DFR RAPSODIE PHENIX JOYO He and Na He Na He He 110 130 130 73 75 9.3 7.6 3.4 6.9 4.3 Bauer et al.58 Blank59 Blank59 Fromont et al.60 Inoue et al.61 Nitride Fuel liquid metal, liquid Li bonding was also suggested for UN-fueled space reactors In a He-bonding concept, the gap is filled with He of atmospheric pressure Besides the pellet-type fuel, vibropac (U,Pu)N fuel pins were irradiated in DFR by use of He for bonding gas.59 A Na-bonding concept is characterized by a large gap width (i.e., >0.5 mm) between fuel pellets and cladding tube and a high density of fuel pellets (i.e., >90% TD) This concept has the advantage of keeping the fuel temperature relatively low due to good thermal conductivity of liquid Na Furthermore, the temperature of fuel pellets is considered as quasiconstant A shroud tube was sometimes used in order to maintain the fuel fragments in their original geometry On the other hand, the disadvantage of a Na-bonding concept includes the difficulty in fuel pin fabrication and spent fuel reprocessing Furthermore, with regard to safety consideration, the possibility of loss of Na in a breached pin has to be evaluated At present, a He-bonding concept is considered as the reference for (U,Pu)N fuel A He-bonding concept is characterized by a small gap width (i.e.,