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.     

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 of large MOX fueled FBR core and applicability of various coolants and fuels aimed at the Self-Consistent Nuclear Energy System

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.     

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.     

R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC

Research and development(R&D) activities on partitioning and transmutation of trans-uranium nuclides (TRU) and LLFP and future R&D program in JNC were summarized. Feasibility design studies have been conducting to investigate the characteristics of a fast reactor core with TRU and LLFP transmutation. It was reconfirmed that the fast reactor has a strong potential for transmuting TRU and LLFP, effectively. R&D for establishing partitioning process of TRU apart from the high-level radioactive wastes have been carried out. By several counter-current runs of the TRUEX process using highly active raffinates, a process flow sheet capable of selective partitioning of actinides and fission products was newly developed. JNC has settled a new R&D program concerning partitioning and transmutation of long-lived radioactive waste based on recommendation of check & review for OMEGA program performed by the Ad Hoc Committee under the Atomic Energy Commission of Japan (AEC). The R&D program is composed of the design studies and development of element technologies (fabrication, irradiation) in the “Feasibility Studies” on commercialized fast reactor system and the basic studies with experiments (nuclear data, reactor physics, fuel property, etc.) to establish database and analytical tools for the TRU- and LLFP- containing fuel and core design.     

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.     

Core concept of a passive-safety fast reactor “METAL-KAMADO” and reactivity coefficients

New concept of a passive-safety simple fast reactor “METAL-KAMADO” with metallic fuels is presented, which has same concept as a passive-safety thermal reactor “KAMADO”. A fuel element of the “METAL-KAMADO” consists of metallic fuel (U–10%Zr) and cooling holes of He gas flow. These fuel elements are located in a reactor water pool of atmospheric pressure (0.1 MPa) and low temperature (<60 °C). In case of LOF, decay heats of fuel elements are removed by natural heat transfer from surfaces of the fuel elements to the reactor water pool.

Preliminary neutronic calculations of the “METAL-KAMADO” show possibility of high burn-up of more than 120 GWd/t with 10% enriched U–Zr fuel. Reactivity coefficients of the core are also discussed.     

Transmutation characteristics of MA and LLFP in a fast reactor

Systematic studies were implemented to investigate the flexibility and attractive core concepts of MA and LLFP transmutation in fast reactors. The MA transmutation in the fast reactor core has no serious drawbacks in terms of core performance, provided that the homogeneous loading method can be employed with a small fraction of MA fuel (2˜5wt%). The recycling of MA in the fast reactor is feasible from neutronic and thermal-hydraulic points of view. For FP transmutation, the introduction of target subassemblies using duplex pellets — a moderator annulus surrounding a Tc-99 core — gives the maximum transmutation rate of Tc-99 in the radial shield region of the fast reactor. The fast reactor has an excellent potential for transmuting MA and LLFP effectively. The fast reactor will be able to play an important role for reduction of environmental burden in future energy system.     

Fusion-driven transmutation of selected long-lived fission products

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 ⁹³Zr and ¹²⁶Sn, the most difficult LLFP to transmute. It is shown that transmutation of ⁹³Zr is effective even with a rather modest neutron loading of 1 MWt·m⁻², typical for ITER project. Transmutation of ¹²⁶Sn, 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 closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

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.     

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO₂ spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.     

压水堆中嬗变99Tc和129I的计算研究

选取大亚湾压水堆作为嬗变参考堆,研究在压水堆中嬗变长寿命裂变产物⁹⁹Tc和¹²⁹I的可行性。计算结果表明：在1个换料周期（18个月）内,⁹⁹Tc的最大嬗变率为15.69%,¹²⁹I的最大嬗变率为9.18%。通过对不同堆芯方案进行安全性分析发现：添加⁹⁹Tc和¹²⁹I后,堆芯有效增殖因数k_eff降低且随燃耗变化的幅度变小;堆芯径向中子通量密度分布无明显变化但径向功率峰因子降低;考虑燃料温度系数、慢化剂温度系数、硼微分价值以及控制棒价值等,得出在反应性温度系数及反应性控制方面不会导致安全问题,相反有优化作用。因此,从安全角度分析,在压水堆中嬗变⁹⁹Tc和¹²⁹I是可行的。     

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.     

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.     

径向步进倒料行波堆的数值研究

行波堆是一种可实现自持增殖-燃耗的新概念快堆,它可直接使用天然铀、贫铀、钍等可转换核材料,实现非常高的燃料利用率。基于行波堆的原理,提出了具有现实应用价值的径向步进倒料行波堆的概念,并将其与典型钠冷快堆的设计相结合,采用数值方法对由外而内的径向步进行波堆二维渐近稳态特性进行了研究。计算结果表明:渐近keff随倒料循环周期近似抛物线分布,而渐近燃耗随倒料循环周期线性增长,满足临界条件的倒料循环周期中最大燃耗可达38%;堆芯功率峰随着倒料循环周期的增长,从燃料卸出区(堆芯中心)向燃料导入区(堆芯外围)移动,功率峰值逐渐降低,在高燃耗情况下,靠近堆芯中心的轴向功率分布呈M形。     

Innovative fast breeder reactor concept ‘RAPID’ for improvement of reactor performance and proliferation resistance

The 60 MWe metal fueled fast breeder reactor concept ‘RAPID’ to improve reactor performance and proliferation resistance has been demonstrated. The reactor can be operated without refueling for up to 5 years. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. RAPID concept enables quick and simplified refueling by replacing an IFA in which all the core and blanket fuel elements are comprised. An on-site storage cask achieves on-site decay heat removal of an IFA. After 3 years of on-site storage, an IFA together with an on-site storage cask can be transported directly to the reprocessing plant without any intermediate steps. Significant improvement of inherent safety features and plant availability has been discussed. Decay heat removal capability, safety consideration on criticality of the IFA and shielding design of the on-site storage cask has been confirmed. The RAPID refueling concept possesses high resistance to state-supported removal of plutonium for nuclear weapons production.     

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO₂ fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.     

Concept of optimization for geologic disposal combined with P&B/T (partitioning, and burning and/or transmutation) treatment in self-completed fuel cycle

Self-Completed Fuel Cycle combined with multiple-recycle P&B/T (partitioning, and burning and/or transmutation) treatment has the potential merits for improving the present geologic disposal scenario, and is discussed herein based on three plus one criteria for partitioning. The number reduction of GSC (glass solidified canister, i.e. waste package) is important concept to decrease the cost of geologic disposal. The cost needed for P&B/T treatment can be saved by the number reduction of GSC than the current disposal scenario.

The total cost for P&B/T treatment with thermal and fast B/T reactors can be minimized by the variation of recycle period, and by the adjustment of out-core optimization and in-core optimization. The partitioning can be done by three plus one criteria, that is (1) the selection of long-lived radio-nuclide with high Hazard Index at 1,000 years after the fuel discharge, (2) the selection of radio-nuclide with high mobility through deep geologic media, and (3) the grouping of MA (minor actinide) and LLFP (long-lived fission product) selected by the items (1) and (2), based on B/T characteristics and decay acceleration ratio of nuclear reactions.     

Measurement of Acoustic Velocity and Attenuation in an Air-Water Two-Phase Medium

Actinides, mainly responsible for the long term risk of spent fuel, are the principal candidates to transmutation due to their large absorption cross sections.

Systems driven by particle accelerators have been investigated in the past to produce fissile material. Recently these systems have been reconsidered to destroy minor actinides (MA) and long-lived fission products (LLFP), reducing the need for the traditional final confinement of radioactive waste.

Two Monte Carlo calculation models have been developped to determine the criticality safety conditions and the burning capability of MAS and of Pu.

A Pu burner, whose core is poisoned with Th to compensate by producing ²³³U the burnup reactivity due to the even Pu isotopes, can operate at a low proton current using perhaps a cyclotron, incinerating 70% of the charged Pu; its burning capability would be the production of about 1.5 PWRs.

Liquid fuel accelerator driven systems can be used in the future (due to the accelerator dimensions) for MA burning using D₂0 as carrier in a homogeneous core; such a system can burn the production of more than 15PWRs.

In the future, also the problem of LLFP burning could be solved definitively using a system with D₂0 as carrier.