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A comparative study was performed with different designs of the fast breeder reactor (FBR) blanket to evaluate the performance of plutonium breeding accompanied with denaturing, and trans-uranium (TRU) transmutation for enhancement of the proliferation resistance and actinide management in the future fuel cycle. Four designs of the FBR blanket fuel were intercompared, conventional depleted uranium (DU) oxide fuels, DU with TRU oxide fuels, DU with TRU oxide fuels with homogeneous/heterogeneous moderators. As results, with heterogeneous moderator in the axial and the radial blanket fuels, the greatest TRU transmutation and plutonium denaturing performance were attained without deteriorating fuel breeding. Design optimization was performed as for ZrH 1.65 moderator rod/pellet arrangement in the axial and the radial blanket fuel assemblies. The optimum number of moderator rods per assembly was 25 (11% of the number of total rods) with proper mutual interval (2–3 cm) and symmetrical arrangement in the radial blanket, and insertion of 1 cm thick ZrH 1.65 moderator pellets to 4 and 10 cm from the top and the bottom planes of the active core was the optimum design in the axial blanket, in order to achieve the highest TRU transmutation ratio and plutonium denaturing performance without deteriorating fuel breeding. From the non-proliferation aspect, material attractiveness of plutonium were evaluated comparing with the criterion based on decay heat, and required TRU doping ratio in DU with TRU oxide fuel were proved to be as 1.6–2.5 wt.% with heterogeneous moderator, 2.1–3.0 wt.% with homogeneous moderator, and 4.1–5.0 wt.% without moderator. The heterogeneous loading of moderation showed the great advantages in fuel treatment in clear the heat generation limit criterion, as well as mass requirement to the initial and multi-cycle phase.

For improvement of proliferation resistance, the protected plutonium production (PPP) concept has been proposed. The concept is to produce 238 Pu which has high decay heat in fuels with appropriate 238 Pu/ 239 Pu ratio for higher proliferation resistance. Estimation of Pu production with high accuracy under various irradiation conditions should be needed for evaluation of fuel specification and its proliferation resistivity. In previous studies, 238 Pu production from 237 Np and 241 Am has been experimentally investigated in our group ( Osaka et al., 2005 , Koyama and Mitsugashira, 2008 ; Koyama et al., 2010 ). However, experimental data of 239 Pu production from 238 U is not enough opened. In this study, 239 Pu production from 238 U is investigated on the basis of experimental data. Nine small 238 U samples were irradiated in the experimental fast reactor Joyo under various irradiation conditions. Isotopic composition of U, Pu and Nd was measured by chemical analysis with accuracy. Plutonium-239 production ratio (a ratio of 239 Pu atoms to initial 238 U atoms) and burnup were also determined. Behavior of 239 Pu production depended on sample loading position (e.g. neutron spectrum), and the tendency is different between fuel stack and reflector region. In order to evaluate this spectrum dependence quantitatively, the 239 Pu production rate is defined as the ratio of 2 39 Pu production per 10 26 n/m 2 of neutron fluence. The 239 Pu production rate was decreased linearly with increasing fast flux ratio (a ratio of fast flux with energy of more than 0.1 MeV to the total flux). Therefore, the 239 Pu production from 238 U by neutron capture reaction could be evaluated under wide irradiation area and/or position based on experimental results.

A feasibility study on natural uranium spallation target in accelerator-driven system (ADS) for minor actinide (MA) transmutation was performed. As a result of comparative study of uranium and lead-bismuth (PbBi) targets in the bare case without blanket surrounding, it was found that uranium target had better neutron generation performance, but limited by the geometrical size due to high neutron absorption in 238U. In ADS for MA transmutation, uranium used as target instead of PbBi also absorbs neutrons passing the target area. More realistic concept of pin type uranium spallation target cooled by liquid PbBi was considered aiming at enhancing spallation target performance in terms of neutron generation efficiency and operation temperature. The uranium pin target design had nothing better effects on neutron balance of such system than a conventional PbBi target in ADS and it was concluded that uranium target was not suitable for the full-scale ADS.

Inventories of secondary nuclear materials such as plutonium (Pu), minor actinides (MAs) and reprocessed uranium (RepU), generated in spent fuel in the wake of nuclear electricity production were predicted for the world and Japanese cases by IAEA-VISTA code, which can be applied to the evaluation of the nuclear fuel cycle with thermal power reactors. Three strategic input parameters, i.e.. nuclear capacity growth, reprocessing ratio and MOX fuel use ratio, were incorporated in the code as well as other technical parameters with respect to reactor-type share, burnup of individual reactor type, load factor, enrichment, etc. The secondary nuclear material inventories in spent fuel discharged from PWR and BWR were calculated for the nuclear fuel cycle in the world and in Japan. VISTA calculation clarified anticipated quantities of Pu in spent fuel and separated form, quantities of MAs (Np, Am, and Cm) and RepU until 2050. MA data resulted from the VISTA calculation were applied to evaluate the mass balance of Np and MAs in the nuclear fuel cycle for Protected Plutonium Production (PPP: P 3 ) with the 1 GWe class PWR, and possible numbers of PWR with available quantities of Np and MAs in the world and in Japan were appraised.

The paper deals with comparison of proliferation resistance of essential fissile/fissionable compositions emerged in potential fuel cycles oriented on production, use and storage of denatured plutonium. The main focus is made on elaboration of associated criteria to bring this comparison on a quantitative base. New evaluation function – attractiveness is introduced that largely relies on a ratio of function of α-Rossi (that reflects energy yield) to characteristics reflecting technological difficulties in bringing nuclear material to supercritical state.

The present study provides a quantitative basis for assessment of proliferation resistance properties of U and Pu isotopes. Within the large area of issues covered by term of proliferation resistance, the only technical characteristics were chosen as a subject for discussion. The present study concentrates on intrinsic proliferation resistance features focusing the material characteristics associated with bare critical mass (BCM), decay heat (DH) and spontaneous-fission neutrons (SFN), which are treated as the main constituents of isotopic barriers against proliferation. The present study aims at determining major proliferation resistant attributes for U and Pu. This determination is largely based on analysis of their critical bare spheres. The MCNP code coupled with JENDL 3.2 data library is the main tool applied for this study. The other characteristics were summarized from the data available in scientific literature.

This paper focuses on improving the proliferation resistance of plutonium resulting from uranium-based fuel irradiation. Intrinsic properties of plutonium isotopes with even mass numbers (238Pu, 240Pu and 242Pu) — in terms of their intense decay heat and high spontaneous fission neutron rates — were used as a measure to improve the proliferation resistance of plutonium itself. The present study explores MA addition effect into LEU (5%235U) and HEU (20%235U) with regard to plutonium proliferation resistance characteristics. Consideration goes beyond critical condition to examine the potential of subcritical system in enhancing the plutonium proliferation properties. Results show that even the doping level of 1% of Np, TrPu or all MA elements into low enriched uranium improves the proliferation-resistant properties of plutonium. A potential for further improvement is achieved by higher doping of minor actinides into high enriched uranium irradiated in a subcritical mode.

The long-term radiological burden associated with nuclear power production is usually attributed to long-lived fission products (LLFP). Their lifetime and large equilibrium mass and hence radioactivity accumulated in the course of fission energy generation make their storage a rather formidable task to solve. Therefore the idea of artificial incineration of LLFP through their transmutation has been quite naturally incorporated into the concept of self-consistent nuclear energy system (SCNES) based primarily on fast breeder reactor technologies. However it is now acknowledged that neutron environment of fission facilities including fast breeder reactors does not seem most appropriate for LLFP transmutation. The issue has been then extensively developed within the framework of multi-component self-consistent nuclear energy system (MC-SCNES). Neutrons of specific quality required for LLFP transmutation are proposed there to be of non-fission origin. Given neutron excess available and neutron quality, a fusion neutron source (FNS) is appearing as the candidate No. 1 to consider for LLFP transmutation. Research on LLFP transmutation by means of FNS has very long history and has received an additional boost during the decade passed. In the present study, potential of thermal flux blanket of FNS is exemplified by transmutation of 93 Zr and 126 Sn, the most difficult LLFP to transmute. It is shown that transmutation of 93 Zr is effective even with a rather modest neutron loading of 1 MWt·m −2 , typical for ITER project. Transmutation of 126 Sn, however, requires neutron loading of as high as 3 MWt·m −2 for DD fusion and is quite unattractive for DT fusion. In the latter case, transmutation through the threshold (n,2n) reaction may be preferable.

The paper deals with transmutation of elemental palladium separated from the spent fuel of fission reactors into the stable isotopes of silver and cadmium. The most efficient transmutation environment is achieved with resonance neutron spectrum in a blanket of fusion-driven transmuter. Three types of plasma conditioning are considered: DT-plasma with equal fractions of deuterium and tritium, DT-plasma with reduced tritium fraction and pure DD-plasma.

The concept of a nuclear fuel recycle system with a nitride fueled FBR core has been investigated as a part of related studies towards the Self-Consistent Nuclear Energy System (SCNES). Nitride fuel has been given attention because of its relatively high fuel density and high thermal conductivity. To materialize the SCNES concept, it is important to adequately use the excess neutrons produced in the chain reaction. The high fuel density of the nitride fuel brings out more of the excess neutrons and has a higher potential to transmute the long-lived fission products (LLFP's). The high thermal conductivity, in addition, provides margin of fuel melting, and gives negative feedback due to the Doppler reactivity in unprotected loss of flow accidents. In this paper, we discuss good use of nitride fuel in the SCNES.

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and removal of stable nuclides from radioactive nuclides with isotope separation using tunable laser. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with a metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.

The present paper focuses on analysis of radiological hazard of spallation products that appears to be an important factor that might affect the choice of beam/target performance in designing the accelerator-driven systems. The analysis is done in terms of toxicity expressed in units of Annual Limit on Intake (ALI). It reveals the significant contribution of alpha emitting rare earths ( 146 Sm, 148 Gd, 150 Gd, 154 Dy) into overall toxicity of spallation targets.

Environmental harmonization of nuclear energy technology is considered as an absolutely necessary condition in its future successful development for peaceful use. Establishment of Self-Consistent Nuclear Energy System, that simultaneously meets four requirements — energy production, fuel production, burning of radionuclides and safety, strongly relies on the neutron excess generation. Implementation of external non-fission based neutron sources into fission energy system would open the possibility of approaching the Multi-Component Self-Consistent Nuclear Energy System (MC-SCNES) with unlimited fuel resources and zero radioactivity release. This provides strong evidence that nuclear energy could be considered as a base for the future sustainable growth in long perspective.

Feasibility of transmutation of the major long-lived FPs (I, Pd, Tc, Sn, Se, Zr, Cs) while maintaining fuel breeding capability for the Self-Consistent Nuclear Energy System is evaluated based on the actinide recycle metal fuel core of a fast reactor. It is shown that I, Pd, Tc, Sn, Se, and Zr can be transmuted simultaneously by an aid of the isotope separation of Pd-107, Zr-93, Sn-126 and Se-79. Cs, which is difficult to transmute with the other FPs, is planned to be utilized as an in-reactor shielding material to confine in the system The overall assessment based on those results indicates that the developed system has the great potential toward the Self-Consistent Nuclear Energy System.

Among fission products (FP) discharged from a fission reactor, long-lived fission products are considered as of primary concern. Their transmutation has been of high priority to reduce the long-term consequences of nuclear energy generation. A self-consistent nuclear energy system (SCNES) in which we center fast breeder reactor may not have enough degree of excess neutron sources to transmute the fission products that potentially would poses environmental hazards in long-term period if they are buried in geologic disposal. Here we propose a so-called multi-component SCNES in which fission reactor systems can be combined with fusion reactor systems mainly for compensating the loss of enough capability for the transmutation. Amongst long-lived fission products, major concern has been paid for iodine and technetium and little attention was given to radioactive 93 Zr, although its hazard appears to be rather substantial. The importance of 93 Zr transmutation is emphasized and the transmutation capability was examined with fusion neutron sources by incorporating adequate moderation structures. As a result, we have demonstrated that the fusion neutron sources with high-flux blanket can be applied to transmute 93 Zr sufficiently and resolve the problem of its accumulation within the time period of several decades.

The potential of a large MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in the self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long -lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a radial blanket region and part of a lower axial blanket region without any significant impact on the reactor's nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 1000 years is as small as that of a typical uranium ore. To realize self-controllability (passive safety), the proposed FBR core concept employs gas expansion modules and a sodium plenum above the core. To realize self-terminability, even if MOX fuel melting should cause a core compaction, re-criticality of the core can be avoided by a fuel dilution and relocation module. The results show the MOX fueled FBR core has potential applicability to the SCNES. The fundamental applicability of various coolants and fuels is evaluated based on neutron balance toward the final goal of the ideal SCNES. The results show that gas coolant has a potential for increasing the transmutation efficiency of LLFPs. And an improved SCNES with several conventional FBRs and a FP transmutation reactor is also studied.

The new cross-section data library BISERM2 elaborated for radiation damage analysis is described. Data base contains neutron induced displacement, hydrogen and helium production cross-sections for 259 stable isotopes of structural materials from aluminum to bismuth for energy range up to 1 GeV, and displacement cross-sections for most important alloys, namely Zircaloy, Inconel and various types of stainless steel.

The potential of a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in the so-called self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99 Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a radial blanket region and a part of a lower axial blanket region without any significant impact on its nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 1000 years is as small as that of a typical uranium ore. To realize self-controllability (passive safety), the proposed FBR core concept employs gas expansion modules and sodium plenum above the core. To realize self-terminability, even if MOX fuel melting should cause a core compaction, recriticality of the core can be avoided by a fuel dilution and relocation module. The results show the MOX fueled FBR core has potential applicability to the SCNES. With the final goal of the ideal SCNES, fundamental applicability of various coolants and fuels is evaluated based on neutron balance. It is shown that the harder the core spectra is, the larger the potential for transmuting LLFPs would be.

Transmutation of radiowaste is often referred to as a crucial step toward environmentally harmonized nuclear energy system. This area concerns with elimination of two classes of radionuclides: minor actinides and fission products. In a great majority of transmutation studies the stress has been given to some particular transmuters, mainly fast and accelerator driven reactors, and potential advantage of their use in waste management. The logic of the present paper goes in an opposite direction. Having finished general discussion on the final goal of transmutation, it turns to identify the radionuclides being the most severe obstacle on this way and, only after that, advocates the appropriate transmutation environments. Among the problematic nuclides cesium isotopes and 244 Cm are accentuated with advocated correspondingly fusion and accelerator-driven transmutation.

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and LLFPs burning capability. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.

On the commercial stage of FBR, the improvement of the fuel handling system which directly removes spent fuels from the reactor without in-vessel storage, will be useful to enhance an effective utilization of nuclear resources. Minor actinides (MA) loading measures such as the heterogeneous loading (two kinds of MA contents-fuel assemblies are loaded to the same reactor) and the homogeneous loading were extracted, and their suitability for the commercial FBR was estimated from standpoints of the fuel handling system cost and the contribution to the smooth fuel recycle flow. And it was clarified that the heterogeneous MA loading was useful for the transition period from LWRs to FBRs due to the promotion for the Pu utilization and the MA transmutation, and the homogeneous MA loading for the period of only FBRs due to the reduction for the TRU (Pu and MA) inventory outside the reactor.

The key element of a Self-Consistent Nuclear Energy System is excess neutron generation. External neutron source is supposed to be included in the nuclear energy structure to improve its neutron balance. The paper presents accelerator based 14 MeV neutron source that might contribute to neutron excess generation required for transmuting of those fission products which are difficult to transmute in ordinary fission reactors.

Self-Consistent Nuclear Energy System simultaneously satisfying four objectives: energy generation, fuel breeding, transmutation of radionuclides and system safety has been proposed as a new phylosophical basis for further nuclear development. Transmutation of radionuclides is closely connected with the key element of Self-Consistent option - excess neutron generation which is assumed to be provided by external neutron source. The paper presented makes a special emphasis on excess neutron generation in fusion neutron source based on DD plasma.

The potential for a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in a self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a two layer radial blanket region without a significant impact on core nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 10 2 years is as small as that of a typical uranium ore. Regarding self-controllability in the system's safety, the proposed FBR core concept has an inherent negative reactivity feedback with a gas expansion module, sodium plenum above the core and burnup reactivity compensation module. So sodium boiling and fuel melting will be avoided in anticipated transient without scram events. Regarding self-terminability, even if the MOX fuel melting should cause a core compaction process, re-criticality of the core can be avoided by a fuel dilution and relocation module.

Feasibility of transmutation of long-lived Fps in BWR is studied from the view point of neuron balance. Pointing out the equilibrium mass amount of LLFP such as 129 I, 99 Tc, and 126 Sn in BWR core, the neutron balance analysis shows that the available neutrons produced by the current BWR core modification is satisfactory to the required amount of neutrons for LLFP transmutation. Core performance of the LLFP-recycled BWR is then analyzed and the result shows satisfactory results for the present core design criteria. The results confirm the potential of the transmutation of LLFP in BWR and that the eventual radioactive hazard of the radwaste coming form the LLFP recycled BWR can be reduced well below the hazard level of the burned uranium.

Self-Consistent Nuclear Energy System simultaneously meets four requirements: energy generation, fuel breeding, burning of radionuclides and system safety. Multi-component nuclear energy system composed of fast and thermal reactors has a potential to approach to self-consistency provided external neutron sources are included in its structure. This paper deals with general analysis of such a system with focus on neutron production, minor actinide fissioning and plutonium vector at equilibrium stage.

Feasibility of burning of the major long-lived FPs (I, Pd, Tc, Sn, Se, Zr, Cs) while maintaining fuel breeding capability for the Self-Consistent Nuclear Energy System is evaluated based on the actinide recycle metal fuel core of a fast reactor. It is shown that I, Pd, Tc, Sn, Se, and Zr can be burnt simultaneously by an aid of the isotope separation of Pd-107, Zr-93 and Se-79. Cs, which is difficult to burn with the other FPs, should be utilized as an in-reactor shielding material to confine in the system. The selection of the target FPs to be burnt are also validated by using the Burden Index. The overall assessment based on those results indicates that the developed system has the great potential to achieve the goal of the Self-Consistent Nuclear Energy System.

Self-Consistent Nuclear Energy System simultaneously meets four requirements: energy generation, fuel breeding, burning of radionuclides and system safety. The key element of the system is excess neutron generation. Analysis of various neutron sources is done with respect to transmutation requirements. Impact of neutron source on energy system performance is analyzed in terms of excess neutron cost. Special emphasis is made on Fusion Neutron Source.

The basic condition and concept to eliminate recriticality-related problems has been investigated from the viewpoint of the safety features of the SCNES. The passive shutdown capability of the intact core was achieved by self-controllability due to the flat core geometry of high thermal conductivity fuel. Recriticality during severe transients was found to be eliminated by the self-terminability due to the controlled material relocation of the leading channel, where the incoherent and localized fuel motion was generated in the degraded core.

Fast Breeder Reactor (FBR) is expected to be commercialized in Japan to overcome foreseeable problems such as reactor safety, increasing energy demand, final disposal of high level radioactive waste and fuel resource shortage. We have been studying three FBR core concepts enhancing its potential abilities; ultra-large type, simplified type and friendly to fuel cycle type core. This study is sponsored by Ministry of International Trade and Industry.

Transmutation of radioactive nuclides into stable or short-lived ones has been under discussion as a measure to deal with the wastes of nuclear technology since its very beginning. There has been a lot of research activities world-wide with various approaches to transmuter design, matrix for waste incorporation and end waste form. This diversity comes from differently formulated transmutation priorities which, in their turn, originated in differently formulated strategy for development of nuclear fuel cycles. The present paper aims at surveying the goals of waste transmutation in view of trends identified in the past decade and address some new characteristics for estimating its radiological cost.