Feasibility study of large MOX fueled FBR core aimed at the Self-Consistent Nuclear Energy System

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.     

Applicability evaluation to a MOX fueled fast breeder reactor for a self-consistent nuclear energy system

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 10² 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.     

An application of metal fuel cycle technology toward Self-Consistent Nuclear Energy System (SCNES) concept

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.     

Elimination of recriticality potential for the Self-Consistent Nuclear Energy System

The ultimate safety goal of the Self-consistent Nuclear Energy System (SCNES) is to eliminate the recriticality-problem based on a simple safety logic. The principle of the elimination of the recriticality-problem is the Controlled Material Relocation (CMR) to establish the neutronic shutdown by removing the molten fuel to the out of core before a large scale pool formation which has potential of energetics driven by a super prompt criticality.

The CMR concept should be reliable without significant impact on the core neutronic performance. As the typical core concepts to enhance this CMR characteristic, several design options are under consideration. They are fuel assemblies with inner duct structure (FAIDUS), fuel assemblies with hollow fuel pins in the axial blanket region (ABLE) for MOX fueled cores, and fuel assemblies without fuel pin bundle structure in the lower axial blanket region (ELAB) for the metallic fueled core. Based on the core design study and accident analyses, these CMR-oriented concepts have been found feasible without significant degradation of the neutronic performance

In order to experimentally confirm the effectiveness of the CMR concept for the MOX fueled core, the EAGLE project has been started in 1998 by Japan Nuclear Cycle Development Institute (JNC) and The Japan Atomic Power Company (JAPC). The EAGLE project is the experimental program utilizing the out-of pile test facility and in-pile facility IGR of the National Nuclear Center of the Republic of Kazakhstan (NNC/RK).     

A metallic fuel cycle for Self-Consistent Nuclear Energy System (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.     

Feasibility study on nitride fuel core and recycling system toward Self-Consistent Nuclear Energy System

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.     

Transmutation of long-lived radioactive waste based on double-strata concept

A considerable attention is directed toward the reduction in the long-term potential hazard by partitioning and transmutation (P-T): separating long-lived nuclides from the waste stream and converting them into either shorter-lived or non-radioactive ones. The effects of higher Pu and minor actinide (MA) compositions on the transmutation rates have been studied for a typical mixed oxide (MOX)-fuel fast breeder reactor (FBR) core with 2600 MWt. The calculations showed that the transmutation rate for (Pu, MA) compositions from MOX -LWR becomes one half than that from UO₂-light water reactor (LWR). Furthermore, MA accumulation and transmutation based on Double-Strata Scenario have been investigated for introducing the accelerator driven transmutation system (ADS) with 800 MWt. It was shown that in the scenario of nuclear plant capacities for maximum 140 GWe, which consists of LWRs and FBRs, the introduction of ADS can play a significant role as “Transmuter” in the back-end of fuel cycle.     

Study on MA and FP transmutation in fast reactors

Feasibility studies have been performed to develop an optimized fast reactor core for reducing long-term radiotoxicity of nuclear waste by minor actinide(MA) and long-lived fission product(FP) transmutation, taking into consideration fuel cycle technology. Systematic parameter survey calculations were implemented to investigate the basic characteristics of MA and FP transmutation in a fast reactor core. The hybrid MA-loading method, where Np nuclide is dispersed uniformly in the core and target subassemblies containing Am, Cm and rare earth nuclides are loaded into the blanket region, has the potential to achieve the maximum transmutation of MA with no special fuel design considerations. The introduction of target subassemblies using duplex pellets - a moderator annulus surrounding a ⁹⁹Tc core - has a great potential to transmute long-lived fission products in the radial blanket region of the fast reactor core.     

A feasibility study on FP incineration for self-consistent nuclear energy system (SCNES)

A fast reactor core and fuel cycle concept is discussed for the future “Self-Consistent Nuclear Energy System (SCNES)” concept. The present study mainly discussed long-lived fission products (LLFPs) burning capability and recycle scheme in the framework of metal fuel fast reactor cycle, aiming at the goals for fuel breeding capability and confinement for TRU and radio-active FPs within the system. Combining neutron spectrum-shift for target sub-assemblies and isotope separation using tunable laser, LLFP burning capability is enhanced. This result indicates that major LLFPs can be treated in the additional recycle schemes to avoid LLFP accumulation along with energy production. In total, the proposed fuel cycle is a candidate for realizing SCNES concept.     

Effect of TRU fuel loading on core performance and plutonium production of FBR

The effect of trans-uranium (TRU) fuel loading on the reactor core performances as well as the actinide and isotopic plutonium compositions in the core and blanket regions has been analyzed based on the large FBR type. Isotopic plutonium composition of TRU fuel is less than that of MOX fuel except for Pu-238 composition which obtains relatively higher composition. A significant increase of plutonium vector composition is shown by Pu-238 for TRU fuel in the core region as well as its increasing value in the blanket region for doping MA case. Excess reactivity can be reduced significantly (5% at beginning of cycle) and an additional breeding gain can be obtained by TRU fuel in comparison with MOX fuel. Doping MA in the blanket regions reduces the criticality for a small reduction value (0.1%) and it gives a reduction value of breeding ratio. Loading MA in the core regions as TRU fuel composition gives relatively bigger effect to increase the void reactivity coefficient mean while it gives less effect for loading MA in the blanket regions. Similar to the void reactivity coefficient profile, loading MA is more effective to the change of Doppler coefficient in the core regions in comparison with loading MA in the blanket regions which gives slightly less negative Doppler coefficient. Obtained Pu-240 vector compositions in the core region are categorized as practically unusable composition for nuclear device based on the Pellaud's criterion. Less than 7% Pu-240 vector compositions in the blanket region are categorized as weapon grade composition for no doping MA case. Obtaining 9% of Pu-238 composition by doping MA 2% in the blanket regions is enough to increase the level of proliferation resistance for denaturing plutonium based on the Kessler's criterion.     

功率展平的压水堆乏燃料发电包层中子学初步研究

聚变裂变混合堆在增殖核燃料、嬗变长寿命核废料及固有安全性等方面具有较大优势,同时,它比纯聚变堆在工程及技术方面要求低,因此较聚变堆更易实现。本工作基于目前国际聚变实验堆（ITER）所能达到的技术水平,提出一种直接利用乏燃料进行发电的聚变裂变混合堆包层概念,利用在不同位置放置不同乏燃料体积分数的方法对燃料增殖区实现了功率展平。计算结果表明：功率展平后的包层功率不均匀系数更小,且包层中燃料区的能量输出要比不展平情况下的能量输出高约21.7%。燃料富集度到运行末期最大可达5.23%。从中子学角度初步论证了该包层的可行性。     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Interpretation of actinide transmutation in thermal and fast reactors

The commonly used transmutation rate of minor actinides in nuclear reactors is decomposed into four components, overall fission rate, Pu production rate, MA production rate, and element production rate. The physical meanings of these factors are described. The transmutation rates of minor actinides in two types of highly-moderated PWRs, a MOX fueled Na cooled fast reactor, and a metal fueled Pb cooled fast reactor are interpreted using the four components. The metal fueled Pb cooled fast reactor can incinerate minor actinides most (79kg/GWth/year), and this amount is about 4 times larger than the thermal reactors. The thermal reactors have large relative overall fission rates for ²⁴¹Am and have a potential for the incineration of ²⁴¹Am.     

Time-dependent Neutronic Analysis for High Level Waste Transmutation in a Fusion-driven Transmuter

This study presents time-dependent transmutations of high-level waste (HLW) including minor actinides (MAs) and long-lived fission products (LLFPs) in the fusion-driven transmuter (FDT) that is optimized in terms of the neutronic performance per fusion neutron in our previous study. Its blanket has two different transmutation zones (MA transmutation zone, TZ_MA, and LLFP transmutation zone, TZ_FP), located separately from each other. High burn-up pressured water reactor (PWR)-mixed oxide (MOX) spent fuel is used as HLW. The time-dependent transmutation analyses have been performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m². The effective half-lives of the MA and LLFP nuclides can be shortened significantly in the considered FDT while substantial electricity is produced in situ along the OP.     

The implication of high-flux fusion neutron source for the self-consistency in multi-component 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 ⁹³Zr, although its hazard appears to be rather substantial. The importance of ⁹³Zr 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 ⁹³Zr sufficiently and resolve the problem of its accumulation within the time period of several decades.     

聚变裂变混合堆钍基增殖锕系元素嬗变包层中子学初步研究

聚变裂变混合堆比纯聚变堆在工程及技术方面要求低,且在产生核燃料、嬗变长寿命核废料以及固有安全性方面具有一定优势,因此,越来越受到人们的重视。增殖包层是混合堆系统的关键部件,已有的包层研究基本上是基于较成熟的铀-钚燃料循环技术。针对我国铀资源相对较少而钍资源较丰富的现状,本文就一种新型的钍基燃料增殖锕系元素嬗变包层进行了初步的中子学研究,利用一维离散纵标法燃耗程序BISONC以及Monte-Carlo粒子输运程序MCNP,对包层的关键核参数,诸如氚增殖比、少量锕系元素的嬗变质量、233U产量以及热功率等,进行了较详细的计算分析。计算结果表明,生成的核燃料233U的富集度可达到3.65%,从而满足压水堆燃料富集度要求。分析结果为下一步的包层优化设计提供了依据。     

Development of Long Life FBR Fuels with Particular Emphasis on Cladding Material Improvement and Fuel Fabrication

The development of FBR fuel systems with high reliability and long in-core residence capability is required to make the fast reactor economically competitive with other electrical energy sources. PNC program of fuels and materials development has been primarily focused on mixed uranium/plutonium oxide (MOX) fuel with cold-worked 316 stainless steel for the past 20 years. Modified 316 stainless steel with excellent swelling resistance and high creep rupture strength was obtained for cladding and duct of the fast prototype reactor MONJU. Advanced austenitic alloys and high strength ferritic alloys are also being investigated for high burnup fuel assemblies of a long life core in large scale FBRs.

In MOX fuel fabrication technology, extensive progress has been achieved during driver fuel fabrication for the experimental reactor JOYO. A new MOX production facility PFPF has been completed with fully automatic and remote handling systems. This facility serves for MONJU core fuel production. The improvement of fuel fabrication technologies promotes cost reduction, safety operation and security from a physical protection standpoint.     

Optimum utilization of nuclear fuel with gas and vapor core reactors

Gas and Vapor Core Reactors (G/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Furthermore, G/VCR systems feature a low inventory and fully integrated fuel cycle with exceptional sustainability and safety characteristics. With respect to fuel utilization, there is practically no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from almost solely a burner to a breeder is achievable. The continuous recycling of fuel in G/VCR systems allows for continuous burning and transmutation of actinides without removing and reprocessing of the fuel. The only waste product at the backend of the gas core reactors' fuel cycle is fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing.

G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than Light Water Reactors (LWRs). In addition, the continuous transmutation and burning of actinides further reduces the quality of the fissile plutonium inventory. The low fuel inventory (about two orders of magnitude lower than LWRs for the same power generation level) combined with continuous burning of actinides, significantly reduces the need for emergency planning and the vulnerability to external threats. Low fuel inventory, low fuel heat content, and online separation of fission fragments are among the key constituent safety features of G/VCR systems.     

压水堆高放长寿命裂变产物堆内嬗变平衡循环设计

基于压水堆多燃料循环管理计算,进行长寿命裂变产物（LLFP）核素堆内嬗变分析。基于长寿命裂变产物核素在乏燃料中的比重及核素的放射毒性,¹²⁹I和⁹⁹Tc作为当前嬗变研究的主要裂变产物。为避免碘同位素分离,参照乏燃料中¹²⁷I和¹²⁹I的组分比例,设计当前的碘化物嬗变靶件。将嬗变核素均匀弥散在惰性慢化材料ZrH₂中,放置在控制棒导向管内进行嬗变分析计算。基于该嬗变组件设计方案,对不同的换料方案进行评价和比较,进而搜索嬗变平衡循环。计算显示,当前带有靶件组件的布料方案可达到平衡循环,并能实现LLFP的嬗变。进一步嬗变优化方案设计受限于当前嬗变组件设计。     

Plutonium recycling in hydride fueled PWR cores

The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. The mono-recycling of plutonium in PWR transmutes less than 30% of the loaded plutonium, providing only a limited reduction in the long-term radiotoxicity and in the inventory of TRU to be stored in the repository. The primary objective of this study is to assess the feasibility of plutonium recycling in PWR in the form of plutonium hydride, PuH₂, mixed with uranium and zirconium hydride, ZrH_1.6, referred to as PUZH, that is loaded uniformly in each fuel rod. The assessment is performed by comparing the performance of the PUZH fueled core to that of the MOX fueled core. Performance characteristics examined are transmutation effectiveness, proliferation resistance of the discharged fuel and fuel cycle economics. The PUZH loaded core is found superior to the MOX fueled core in terms of the transmutation effectiveness and proliferation resistance. For the reference cycle duration and reference fuel rod diameter and pitch, the percentage of the plutonium loaded that is transmuted in one recycle is 53% for PUZH versus 29% for MOX fuel. That is, the net amount of plutonium transmuted in the first recycle is 55% higher in cores using PUZH than in cores using MOX fuel. Relative to the discharged MOX, the discharged PUZH fuel has smaller fissile plutonium fraction - 45% versus 60%, 15% smaller minor actinides (MA) inventory and more than double spontaneous fission neutron source intensity and decay heat per gram of discharged TRU. Relative to the MOX fuel assembly, the radioactivity of the PUZH fuel assembly is 26% smaller and the decay heat and the neutron yield are only 3% larger. The net effect is that the handling of the discharged PUZH fuel assembly will be comparable in difficulty to that of the discharged MOX assembly while the proliferation resistance of the TRU of the discharged PUZH fuel is enhanced.