Fast Reactor Cycle Research Articles

THE IMPACT OF FUEL CYCLE OPTIONS ON THE SPACE REQUIREMENTS OF A HLW REPOSITORY

Because of increasing concerns regarding global warming and the longevity of oil and gas reserves, the importance of nuclear energy as a major source of sustainable energy is gaining recognition worldwide. To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.

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 UO 2 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.

A NEXT GENERATION SODIUM-COOLED FAST REACTOR CONCEPT AND ITS R&D PROGRAM

Critical issues in the development targets for the future fast reactor(FR) cycle system, including sodium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercialization is described, to be followed up in a new framework of the Fast reactor Cycle Technology development(FaCT) Project launched in 2006.

Uranium Crystallization Test with Dissolver Solution of Irradiated Fuel

The crystallization process has been developed as a part of the advanced aqueous process, NEXT (New Extraction System for TRU recovery) for fast reactor (FR) cycle. In this process, a large part of U is separated from dissolver solution by crystallization as UO2(NO3)2.6H2O. The U crystallization test was carried out with real dissolver solution of irradiated FR fuel to investigate the influence of cooling rate on the crystal size and the behavior of fission product (FP) compared with that of Pu(IV). In regard to the influence of the cooling rate, it was confirmed that the crystal size was smaller as the cooling rate is faster. Although it was expectable that the decontamination performance was improved by diminishing the specific surface of the crystals, it was suggested that a large crystal produced by crystallization was not always high purity. Concerning the behavior of FPs, Eu behaved similarly to Pu(IV). Cs accompanied with U into the crystals under the condition in this test.

ＭＯＸ燃料を使用した高温ガス炉の核熱設計

Design studies of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C), which can produce both on electric power and hydrogen, have proceeded in Japan Atomic Energy Agency. In future, it will be obliged to operate using not enriched uranium but plutonium to coexist with fast reactors after the full deployment of a fast reactor cycle. Therefore, a nuclear and thermal design has been performed to confirm the feasibility of the reactor core using Mixed-Oxide (MOX) fuel. The reactor core with operation period of 450days and average burn-up of 123 GWd/ton for discharged fuel was designed. The reactor core met safety requirements of the maximum fuel temperature of less than 1,400°C during normal operation, the maximum power density of less than 13 W/cm3, shutdown margin of more than 1.0%Δk/kk′ and negative reactivity coefficient. The results proved that it is possible to operate the GTHTR300C using MOX fuels without consuming natural uranium resources.

Main R&D issues for fast reactor structural design standard

For realization of economical and reliable fast reactor (FR) plants, the Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC) are cooperating on the “Feasibility Study on Commercialized FR Cycle Systems”. To certify the design concepts through evaluation of the structural integrity of FR plants, the research and development of the “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)” is recognized as an essential theme. The FDS focuses on particular failure modes of FRs such as ratchet deformation and creep-fatigue damage due to cyclic thermal loads. For precise evaluation of these modes, the research and development for three main issues is in progress. First, the “Refinement of Failure Criteria” needs to be addressed for particular failure modes of FRs. Secondly, the development of “Guidelines for Inelastic Design Analysis” is conducted to predict elastic plastic and creep deformation under elevated temperature conditions. Lastly, efforts are being made toward preparing “Guidelines for Thermal Load Modeling” for the design of FR components where thermal loads are dominant.

06/02007 Reprocessing of AHWR spent-fuel: challenges and strategies: Kant, S. International Journal of Nuclear Energy Science and Technology, 2005, 1, (2–3), 204–214

A fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel is given. The present cermet fuel consists of an oxide solid solution of Th and minor actinides and Mo-inert matrix. It has been proposed as a high-performance device that can enhance minor actinide incineration in a fast reactor cycle. It is used in an independent small sub-cycle, whereby dedicated cycle technologies are adopted. Two-step reprocessing process was proposed for the present cermet fuel; it consists of a pre-removal of Mo-inert matrix and an actinide recovery. A preliminary test for the pre-removal of Mo-inert matrix was carried out using a surrogate cermet fuel. Burnup characteristics of a fast reactor core loaded with the cermet fuel were investigated by using neutronic calculation codes. It was revealed that a heterogeneous composition of Mo-inert matrix in inner and outer cores may lead to an effective transmutation of minor actinides and a flattened power density. It was concluded that the present cermet fuel was potentially promising as a high-performance incineration device of minor actinides for fast reactors.

06/01203 Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors: Kotake, S. et al. Progress in Nuclear Energy, 2005, 47, (1–4), 300–313

06/01203 Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors: Kotake, S. et al. Progress in Nuclear Energy, 2005, 47, (1–4), 300–313

Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor

An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors

The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under progress in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been also carried out. Crucial development issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. As a result, it has been confirmed that the sodium-cooled FR concept is highly suited to the development targets and R&D issues are related enhancing the economy with certain perspectives for realization. A flexible and robust development program for the FR cycle system will be proposed taking account of the characteristics for each FR concept until the end of the Phase II study.

431. 高速増殖炉燃料再処理における新脱被覆システムの開発

As a part of the feasibility study on commercialized fast reactor cycle systems, Japan Nuclear Cycle Development Institute (JNC) has been developing the fuel decladding technology for the dry reprocessing process (oxide electrowinning process) and aqueous reprocessing process. Particularly, in the oxide electrowinning process, the spent fuel should be reduced to powder for quick dissolution in the molten salt at electrolyzer. Therefore, JNC proposes new decladding system with innovative mechanical decradding devices. The decladding system consists of fuel crushing stage, hull separation stage and hull rinsing stage. In the fuel crushing stage, disassembled spent fuel pins are crushed and powdered by mechanical decladding device, then the following stage. the hull and the fuel powder are separated by magnetic separator. Only the fuel powder is fed to the electrolyzer. On the other side, the separated hull is melted by induction heating method, and the small amount of oxide included in the hull fragments is recovered at the hull rinsing stage. The recovered oxide fuel is fed back to the electrolyzer. In this paper, the basic performance of the element equipment that composes this new decladding system will be descried.

Advanced MOX Core Design Study of Sodium-Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan

Sodium-cooled mixed-oxide core design studies are performed with a target burnup of 150 GWd/t and possible measures against the recriticality issues in core disruptive accidents. Four types of core are compared from the viewpoints of core performance and reliability. Results show that all the types of core satisfy the target and that a homogeneous core with an axial blanket partial elimination subassembly is the superior concept, although experimental demonstration is required of molten fuel motion for mitigation of recriticality following fuel melting and loss of fuel pin integrity.

Conceptual Design Study on Advanced Aqueous Reprocessing System for Fast Reactor Fuel Cycle

The design study of an aqueous reprocessing system has been progressed for the feasibility study on commercialized fast reactor cycle systems in Japan. A simplified PUREX process, with the addition of a uranium crystallization step and a minor actinide (MA) recovery process was proposed as the NEXT process. For the simplified PUREX process and the SETFICS/TRUEX process for MA recovery, small-scale hot tests were conducted. The test results showed the decontamination factor (DF) of 105 for the U, Pu and Np product was a reasonable target for the simplified PUREX process. Concerning to the Am and Cm product, the DF of rare earth elements (REs) and the other fission products (FPs) were over 10 and over 102, respectively. For economical competitiveness, an aqueous reprocessing plant with the capacity of 200 tHM/yr cleared the target cost, while the plant with the capacity of 50tHM/yr did not clear it. For efficient utilization of resources, the calculated U and TRU recovery of both plants was over 99wt%. Using salt-free reagents and repeating concentration of waste solutions, the environmental impact was reduced. Pu was not isolated through the whole processes in the NEXT process in order to keep the proliferation resistance.

実用化戦略調査研究でのNa炉の1次主冷却系設計(液体金属冷却炉,革新型原子炉の開発および多目的利用技術,原子力要素技術開発)

In the "Feasibility Study on Commercialized Fast Reactor Cycle Systems", an innovative concept of a sodium-cooled loop-type reactor to pursue high economic competitiveness has been developed by adopting several innovative concepts. For one of the concepts, 12Cr-steel with high strength and low thermal expansion rate has been employed as a structural material of the primary piping. As a result, the diameter is enlarged and length is shortening and the cooling system has been remarkably simplified. For the design concept of the hot leg piping, it is designed as a simple L-shaped top-entry type piping selected on the view point of economical and technical issues. And the structural integrity against a thermal stress and the instability of the vibrations of primary piping are evaluated.

Conceptual Design Study on Advanced Aqueous Reprocessing System for Fast Reactor Fuel Cycle

The design study of an aqueous reprocessing system has been progressed for the feasibility study on commercialized fast reactor cycle systems in Japan. A simplified PUREX process, with the addition of a uranium crystallization step and a minor actinide (MA) recovery process was proposed as the NEXT process. For the simplified PUREX process and the SETFICS/TRUEX process for MA recovery, small-scale hot tests were conducted. The test results showed the decontamination factor (DF) of 10 5 for the U, Pu and Np product was a reasonable target for the simplified PUREX process. Concerning to the Am and Cm product, the DF of rare earth elements (REs) and the other fission products (FPs) were over 10 and over 10 2 , respectively. For economical competitiveness, an aqueous reprocessing plant with the capacity of 200tHM/yr cleared the target cost, while the plant with the capacity of 50 tHM/yr did not clear it. For efficient utilization of resources, the calculated U and TRU recovery of both plants was over 99 wt%. Using salt-free reagents and repeating concentration of waste solutions, the environmental impact was reduced. Pu was not isolated through the whole processes in the NEXT process in order to keep the proliferation resistance.

03/02262 R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC: Aizawa, K. Progress in Nuclear Energy, 2002, 40, (3–4), 349–356

03/02262 R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC: Aizawa, K. Progress in Nuclear Energy, 2002, 40, (3–4), 349–356

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.

C109 FBR サイクルに関する実用化戦略調査研究(フェーズ 1)について

Japan Nuclear Cycle Development Institute (JNC) have started the following feasibility studies (F/S) in order to develop a commercialized fast breeder reactor (FBR) cycle system satisfying the needs of the future society since JFY1999 with the electric companies. In the F/S, a number of candidate concepts will be selected from various options about the FBR system and the related fuel cycle system (reprocessing and fuel fabriction), featuring innovative technologies. The options are examined considering the attainable perspective for the followings : 1) safety, 2) economic competitiveness to future LWRs, 3) efficient utilization of uranium resources, 4) reduction of environmental burden and 5) enhancement of nuclear non-proliferation in order to determine the promising concepts making the most of the merits of the FBR cycle system and to define the necessary R&D tasks. In the end of JFY2000,we will determine the promising concepts from various options and summarize the results of the F/S (Phase I). Current status of these F/S activities was described in this paper.

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.

4598308 Easily repairable, low cost, high speed electromechanical assembly of integrated circuit die

Japan Atomic Energy Agency (JAEA) initiated “Fast Reactor Cycle Technology Development (FaCT)” project since 2005 to accomplish necessary R&Ds goals for Japan sodium-cooled fast reactor (JSFR) cycle (the loop type sodium-cooled fast reactor with MOX fuel and its related fuel cycle system). Since the concept of multi-transuranic (TRU) fuel recycling has been adopted to reduce environmental burden and to enhance proliferation resistance, decay heat and radiation of MA, etc. became major consideration in designing fresh fuel shipping cask.The use of dry cask and helium gas filling for its high thermal conductivity has been selected, since water as a coolant is not preferable in order to avoid corrosion of fuel steel by sodium–water reaction products. Hence the helium gas cask with five fuel subassemblies storage was designed. To assure the thermal integrity, the cladding temperature of the fuel assembly has to be kept below the creep temperature to prevent any creep damage before fuel loaded into the core. In this study, a whole cask analysis using three-dimensional FEM model has been carried out to confirm absence of creep damage on fuel cladding and thermal integrity complying with the IAEA safety standards. Furthermore, detailed fuel subassembly analysis using two-dimensional FEM model estimated the maximum cladding tube temperature in detail to confirm the heat transport performance. In addition, the cask filled with water is assessed by Monte Carlo method for its integrity at criticality.