Fast Reactor Cycle Technology Development Research Articles

Development of adjusted nuclear data library for fast reactor application

In Japan, development of adjusted nuclear data library for fast rector application based on the cross-section adjustment method has been conducted since the early 1990s. The adjusted library is called the unified cross-section set, which is an ABBN-type group constant set with 70-group energy structure. The first version was developed in 1991 and is called ADJ91. After that, ADJ98, ADJ2000, ADJ2000R, and ADJ2010 were constantly developed. For instance, ADJ2010 was developed based on JENDL-4.0, which provides covariance data needed to apply the cross-section adjustment method, by using 488 integral experimental data acquired in typical fast reactor systems. ADJ2010 has been used as the standard cross-section set for nuclear design in the fast reactor cycle technology development project (FaCT) and the succeeding fast reactor projects. In parallel, the integral experimental data were further expanded to improve the design prediction accuracy of the core loaded with MA and/or degraded Pu. Using the additional integral experimental data, development of the next version of ADJ2017 was started in 2017. In 2022, the latest unified cross-section set AJD2017R was developed based on JENDL-4.0 by using 619 integral experimental data. An overview of the latest version with a review of previous ones will be shown. On the other hand, the latest Japanese evaluated nuclear data library JENDL-5 was released at the end of 2021. In the development of JENDL-5, some of the integral experimental data used in ADJ2017R were explicitly utilized in the nuclear data evaluation. However, this is not reflected in the covariance data. This situation needs to be considered when developing a next version of the unified cross-section set based on JENDL-5. Preliminary adjustment calculation based on JENDL-5 is performed using C/E (calculation/experiment) values simply evaluated by a sensitivity analysis. The preliminary result of the JENDL-5-based adjustment will be also discussed.

Benchmarking of economic evaluation models for an advanced loop-type sodium cooled fast reactor

An economic calculation model based on detailed mass-flow (hereinafter referred to as the “JAEA model”) was developed for the comprehensive evaluation of an advanced loop-type sodium-cooled fast reactor (SFR) cycle system designed in the Fast Reactor Cycle Technology Development (FaCT) project launched in 2007 in Japan. The JAEA model enables calculation of the inventory or processing amount of nuclear fuel in a nuclear fuel cycle and its composition in each facility by simulating mass-flow among nuclear supply chains, and has a function to evaluate the economics and radiotoxicity of nuclear waste based on the processing amount. To identify the difference in economic evaluation methods between the JAEA model and an internationally-authorized code, and verify its calculation functions, the generation cost of an advanced loop-type SFR system was evaluated using the JAEA model and the Generation IV Excel Calculations of Nuclear Systems (G4-ECONS) developed by the Generation IV International Forum (GIF). It was clarified that the JAEA model is influenced to higher degree by the discount rate. When the present value was not taken into account, the results of both methods were quite similar, but it was found that the sensitivity of the load factor is relatively larger the G4-ECONS than in the JAEA model.

Design Studies of a Low Sodium Void Reactivity Core Able to Accommodate Degraded TRU Fuel

As part of the Fast Reactor Cycle Technology Development (FaCT) Project, JSFR (Japan Sodium-Cooled Fast Reactor) core design efforts have been made to cope with the transuranic (TRU) fuel compositions expected during the light water reactor (LWR)–to–fast breeder reactor transition period, during which various kinds of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and their recycling method. The sodium void reactivity, which is one of the major core safety parameters, is considerably influenced by TRU fuel compositions. The criteria assigned to the JSFR core include a void reactivity effect limited to ∼6 $; therefore, designing a core with reduced sodium void reactivity will offer a greater margin for the core to host TRU fuel.To this end, a new core concept called BUMPY is proposed. This homogeneous core exhibits a low sodium void reactivity, due to partial-length fuels with an upper sodium plenum interspersed within the core, among other standard fuel assemblies. This core configuration enhances the upward and lateral neutron leakage from the core fuel region toward the sodium plenum when voiding to reduce void reactivity. The BUMPY core is applied to the 750-MW(electric) JSFR core design. The core can meet the design target by adjusting the loading fraction of the partial-length fuels and the height of the step in fuel lengths. The calculated void reactivity of the selected BUMPY core is 2.5 $ (25% loading fraction, 30-cm step height), which is considerably reduced from the 5.3 $ value of the reference core. This allows the BUMPY core to accommodate 5% to 9% more minor actinides in the core compared to the reference core.

JSFR key technology evaluation on fuel handling system

A simplified fuel handling system design for the demonstration Japan sodium-cooled fast reactor (JSFR) has been proposed. Fast Reactor Cycle Technology Development project phase I results of key technology evaluations on a pantograph fuel handling machine (FHM), a fuel transfer pot with two core component positions, dry spent fuel cleaning and minor actinide-bearing fresh fuel shipping cask are summarized. A full-scale FHM mockup has been fabricated and tested in the air accumulating performance and seismic tolerance data. A mockup fuel transfer pot with fins and chromium carbide coating has been fabricated and tested with sodium accumulating heat transfer performance data. Several sodium cleaning tests using a dummy subassembly has been conducted accumulating cleaning performance data. For fresh fuel shipping cask, a design tool for evaluation of heat transfer capability has been developed and a helium gas cask shows cooling capability of minor actinide-bearing fresh fuel. Those experimental and analytical efforts have shown that key technologies to develop simplified fuel handling system are matured enough to proceed large-scale sodium experiments and conceptual design study for the demonstration JSFR.

Development of constitutive models for fast reactor design

R&D to enable a practical fast breeder reactor plant is proceeding in Japan, which is called “FaCT” (Fast reactor Cycle Technology development). One of the key issues of R&D is to realize a reasonably compact reactor vessel by eliminating the wall protection equipment which is installed inside the vessel in order to reduce thermal loading in the conventional design. Most important concern is the amount of the inelastic strain of the vessel accumulated around the liquid sodium surface which moves upward and downward cyclically with start-up and shut-down. The aim of this study is to develop rational constitutive models that enable prediction of this kind of complex inelastic behaviors precisely and to prepare the design guide based on inelastic analysis. We developed a high accuracy plasticity model and a simplified plasticity model and valuated them by organized experiments.

Under-Sodium Endurance Experiment of Selector Valve in Failed-Fuel Detection and Location System of JSFR

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress in the Fast Reactor Cycle Technology Development (FaCT) project in Japan. In the design of JSFR, a selector valve mechanism is adopted for its failed-fuel detection and location (FFDL) system. Since JSFR has only two FFDL units for [approximately]600 fuel subassemblies due to its compact design of the reactor vessel, one FFDL unit must handle a much larger number of subassemblies than in previous designs. In addition, during the long plant life of 60 years, the wear length of the selector valve will become longer than those of past reactors. Therefore, the endurance of the selector valve becomes important. To demonstrate the manufacturability and endurance of the selector valve, a full-size mockup valve including coating to protect the sliding mechanism was manufactured, and an endurance experiment of the mockup model under high-temperature sodium was conducted. Dimensional inspections and seal performance showed manufacturability of the selector valve. The cross-section observation, hardness measurement, and chemical assay results after the endurance experiment showed that the coating layer on the sliding surface still remains. Thus, the endurance of the JSFR selector valve was demonstrated.

Feasibility study on shipping cask for JSFR fresh fuel contained minor actinide

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.

Comparison of Sodium-Cooled Reactor Fuel-Handling Systems with and Without an Ex-Vessel Storage Tank

The JSFR is a commercial sodium-cooled fast reactor that has been studied in the Fast Reactor Cycle Technology Development (FaCT) project since 2006. For JSFR fuel handling, various fuel-handling systems (FHSs) were investigated, and an advanced FHS with an ex-vessel storage tank (EVST) was selected. This paper summarizes the various FHS concepts and comparisons among those concepts. In the reference system, spent-fuel subassemblies are cooled in the EVST before transfer to the spent-fuel storage pool. The other FHS concepts investigated are evolutional FHSs without an EVST. The result has indicated that the construction cost of the evolutional systems does not reduce the construction cost dramatically, which is mainly due to additional safety measures that required higher decay heat handling in a gas atmosphere and separated fresh and failed fuel storage. From an economics point of view, a longer plant outage of the evolutional systems offsets its advantage of the lower construction cost. Based on the results of this comparative study, JSFR selected the FHS with an EVST.

Water experiment and numerical simulation on failed fuel detection and location system of Japan sodium-cooled fast reactor (JSFR)

A conceptual design study of Japan sodium-cooled fast reactor (JSFR) is in progress as the “Fast Reactor Cycle Technology Development (FaCT)” project in Japan. A selector-valve type failed fuel detection and location (FFDL) system is applied to the JSFR design that has an upper internal structure (UIS) with a slit above the core and several sampling nozzles for the FFDL are set in the UIS around the slit to detect the fission product (FP) from the subassemblies below the slit. Therefore, mixing process in the UIS of complicated geometry should be known and appropriate arrangement of the sampling nozzles in the UIS is needed. A water experiment using a 1/5-scale model was carried out to investigate the mixing process in the UIS and concentration distribution of FP simulant. Experimental results showed that the sampling nozzles set in the UIS detected the FP simulant concentration within the criteria of FFDL signal detection, even in case of failed subassemblies under the UIS slit. In addition, identification of the failed fuel subassembly under the UIS slit was achieved by means of comparing concentration profiles in the UIS. Therefore, the suitable sampling nozzle arrangement was obtained for JSFR. A numerical simulation using a CFD code was carried out and the simulation method was validated based on the experimental data of the FP simulant concentrations. The simulation results showed that the simulation predicts the FP concentration distributions.

Status of Integrated IHX/Pump Development for JSFR

This paper illuminates the status of research and development on the integrated IHX/Pump concept. The integrated IHX/Pump is the incorporated component of the intermediate heat exchanger (IHX) and the primary pump. Among the innovative technologies of the Japan Sodium-Cooled Fast Reactor (JSFR) in the Fast Reactor Cycle Technology Development (FaCT) project, the integrated IHX/Pump concept is one of the major innovative ideas for plant economy by reducing the amount of material in the primary cooling system and the building volume. This report summarizes the view of the integrated IHX/Pump, a development plan, evaluation methods, and the present test results with the 1/4-scale IHX/Pump test device.

A Preliminary Assessment of the Adoption ofInnovative Technologiesinthe Fast Reactor Cycle Technology Development (FaCT)Project in Japan

JAEA has been implementingthe FaCTproject in cooperation with electric utilities toward the commercialization of fast reactorcycle system before2050.In this FaCTproject,many innovative technologies with technical challenges are actively used in order to provide significant improvements in economic competitiveness, enhancement of safety & reliability, sustainability, and nonproliferation.The workofdecidingon the adoption of innovative technologiesby the end of JFY2010 is in progress.This paper describes current preliminaryassessment results.

ICONE19-43819 WATER EXPERIMENTS ON THERMAL STRIPING IN REACTOR VESSEL OF JAPAN SODIUM-COOLED FAST REACTOR : COUNTERMEASURES FOR CONTROL RODS AND RADIAL BLANKET ASSEMBLIES

Design study of an advanced loop-type sodium-cooled fast reactor, JSFR, has been carried out in a frame work of Fast Reactor Cycle Technology Development Project (FaCT) in Japan. As the temperature differences among the control rod channels, blanket assemblies and the core fuel assemblies are around 100℃, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). Then, a water experiment was conducted using a 1/3 scale 60 degree sector model of the core and reactor upper plenum. Characteristics of temperature fluctuations near the cold fluid outlets were obtained and it was confirmed that several countermeasures can reduce temperature fluctuations at the bottom of UIS.

Status of Integrated IHX/Pump Development for JSFR

This paper illuminates the status of research and development on the integrated IHX/Pump concept. The integrated IHX/Pump is the incorporated component of the intermediate heat exchanger (IHX) and the primary pump. Among the innovative technologies of the Japan Sodium-Cooled Fast Reactor (JSFR) in the Fast Reactor Cycle Technology Development (FaCT) project, the integrated IHX/Pump concept is one of the major innovative ideas for plant economy by reducing the amount of material in the primary cooling system and the building volume. This report summarizes the view of the integrated IHX/Pump, a development plan, evaluation methods, and the present test results with the 1/4-scale IHX/Pump test device.

ICONE19-43720 ENGINEERING SCALE TESTS OF MECHANICAL DISASSEMBLY AND SHORT STROKE SHEARING SYSTEMS FOR FBR FUEL ASSEMBLY

Japan Atomic Energy Agency (JAEA) and The Japan Atomic Power Company (JAPC) have been developing an advanced head-end process based on mechanical disassembly and short stroke shearing systems as a part of Fast Reactor Cycle Technology Development (FaCT). Fuel pins for a fast reactor are installed within a hexagonal shaped wrapper tube made of stainless steel. In order to reprocess the fast reactor fuel pins, they must be removed from the wrapper tube and transported to the shearing system without failure. In addition, the advanced aqueous reprocessing process, called "NEXT" (New Extraction System for TRU Recovery) process requires a solution of the spent fuel with relatively high concentration (500g/L). JAEA and JAPC have developed the mechanical disassembly and the short stroke shearing technology which is expected to make fragmented fuel to satisfy these requirements. This paper reports the results of engineering scale tests on the mechanical disassembly and short stroke shearing systems. These tests were carried out with simulated FBR fuel assembly and removed pins. The mechanical cutting method has been developed to avoid fuel pin failure during disassembly operation. The cutting process is divided into two modes, so called "slit-cut" for cutting the wrapper tube and "crop-cut" for the end plug region of the fuel pin bundle. In the slit-cut mode, the depth of cutting was automatically controlled based on the calculated wastage of the cutting tool and deformation of the wrapper tube which had been measured before the cutting. This procedure was confirmed to minimize the fuel pin failure which was hard to prevent in the case of laser cutting. The cutting speed was also controlled automatically by the electric current of the cutting motor to lower the load of the cutting tool. The removed fuel pins were transported to the shearing machine, whose fuel shearing magazine width was set to be narrow to realize the suitable configuration for the short stroke shearing. In the short stroke shearing test, the simulated fuel pins of a commercial reactor and the prototype reactor "MONJU" were sheared at the length of 10±5mm with enough opening at both ends.

S054092 ナトリウム冷却高速炉1次系ホットレグ配管の流力振動評価(2) : ホットレグ配管1/3縮尺試験による流動状況と圧力変動特性([S05409]噴流,後流,およびはく離流れの基礎と応用:最前線(9))

A conceptual design study of the Japan Sodium-cooled Fast Reactor (JSFR) is in progress in "the Fast Reactor Cycle Technology Development (FaCT) project". In this project, a two-loop primary system is adopted in order to economize the plant construction cost. The reduction of the number of loops for the primary cooling system increases the primary coolant flow rate per loop. The mean sodium velocity in the pipes increases to over 9 m/s and Reynolds number in the hot-leg pipes is about 4x107. In the JSFR the pipe thickness is designed to be considerably thinner than that of a light water reactor, because the fast reactor is operated at low pressure and high temperature. One of the issues for the piping design is to understand the behavior of flow-induced vibration that is derived from the hydraulic characteristics under high Reynolds number conditions. We made a flow-induced vibration test facility that simulates the hot-leg piping with a 1/3-scale model .Flow-induced vibration experiments were performed using this facility to evaluate and confirm the integrity of piping against the flow-induced vibration. In these tests, we investigate flow patterns and fluctuating pressures on the pipe wall. This paper describes a summary of the 1/3-scale experimental results under the rectified-flow, swirl 庸low and deflected-flow conditions at the inlet of the hot-leg piping.

Development of Failed Fuel Detection and Location System in Sodium-Cooled Large Reactors (Sampling Method of Failed Fuels under the Slit)

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the “Fast Reactor Cycle Technology Development (FaCT)” project in Japan. JSFR adopts a selector-valve mechanism for a failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, appropriate sampling method for indentifying under-slit failed fuel subassemblies has been developed by water experiments.

Current Status and Future Plan of Nuclear Fuel Cycle in Japan, with Focus on Human Resource Development

Abstract Japan's basic nuclear policy isto reprocess spent fuel and to effectively use the recovered plutonium and uranium.MOX fuel utilization in LWRsispromotedin 16-18 reactors by FY2015. Commercial operation of Rokkasho Reprocessing Plant is planned to start in 2012. Prototype reactor “Monju” restarted operation in May 2010. From FY 2007, Fast Reactor Cycle Technology Development Project (FaCT project) started which focuses more toward the commercialization stage FBR cycle. Basic scenario of Japan's R&D aims for realization of demonstrationFBR by around 2025 and introducing commercial FBRs before 2050. Smooth transition fromLWR fuel cycle to FBR one is an important point.Fornuclear fuel cycle which requires long termR&D, human resources development and keeping isvitally important.

Development of System Based Code (1) Reliability Target Derivation of Structures and Components

The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code (SBC) by considering the safety point of view. This new method utilizes analysis models of a probabilistic safety assessment (PSA), and the reliability target is derived from (1) the proposal to a quantitative safety goal that was published by the nuclear safety commission (NSC) of Japan; (2) the quantitative safety design requirements on the core damage frequency (CDF) and the containment failure frequency (CFF) that were determined in the Fast Reactor Cycle Technology Development (FaCT) project by Japan Atomic Energy Agency (JAEA). This method was applied to determine the reliability target of the structures and components which constitute the reactor cooling system in the Japan sodium-cooled fast reactor (JSFR). The risk from the reactor is shown by the sum of combination of various elements in the PSA analysis model. Those elements include dynamic failures and static failures of the structures and components, and human errors. However, the present study focuses on the sequences including the static failure, and the probability of dynamic failures and human errors in those sequences is conservatively assumed as a unity. It was confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determine the reliability target associated with a random failure of the structures and components, and also confirmed that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.

Comparison of advanced fast reactor pool and loop configurations from the viewpoint of construction cost

Comparative evaluation of fast reactor pool/loop configurations in the JAEA feasibility study (FS) from 1996 to 2006 has been re-evaluated from the new point of view. In the FS, both pool and loop configurations have been investigated. The FS loop concept (FS-loop) is an advanced two-loop design which is being developed in the Fast Reactor Cycle Technology Development (FaCT) project as Japan Sodium-cooled Fast Reactor (JSFR). In this study, a brief description of the FS pool concept (FSpool) has been provided. The FS-pool pursues a compact reactor vessel structure using a flask shape intermediate heat exchanger. The original FS pool/loop comparison in 2000 concluded that material amounts of those systems were about similar between pool and loop configurations. However, the present comparison based on the reactor vessel structure and primary cooling system pointed out relatively good economic potential of the FS-loop concept, because differences in the secondary cooling systems of the pool/loop configurations are not essential from the viewpoint of pool/loop comparison. In addition, a rough estimation method of a reactor vessel diameter against power has been proposed and the result has been compared with RV diameters of various pool/loop concepts. The discussion on the reactor vessel diameter shows that the FS-pool reactor vessel is the most compact of the recent commercial reactor designs such as SPX-2, SNR-2, EFR, CDFR and BN-1600. This study points out better economic potential of FS-loop (present JSFR) as a result of competition with the most compact pool concept.

Experimental Study on Thermal Stratification in a Reactor Vessel of Innovative Sodium-Cooled Fast Reactor—Mitigation Approach of Temperature Gradient across Stratification Interface—

An innovative sodium-cooled fast reactor has been investigated as part of the fast reactor cycle technology development project (FaCT). Thermal stratification after a scram is one of the main thermal loads of a reactor vessel (R/V). R/V has an upper inner structure (UIS), which consists of perforated horizontal plates and control rod guide tubes, and has a slit in the radial direction for fuel handling. The UIS slit causes an asymmetric flow pattern in R/V. A water experiment using a 1/10-scale model was carried out. A steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. This means that the jet through the UIS slit entrains the bottom of the stratification interface. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at the dipped plates where a fuel handling machine was inserted during a fuel exchange operation, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 21% smaller than that in the case of the higher plug position. This reduction of the temperature gradient was sufficient to maintain the structural integrity of the R/V wall against the thermal stratification.