Demonstration Fast Breeder Reactor Research Articles

The investigation and calculation of the transmutation paths for the production of 252cf in fast reactors

The generation and transmutation process of the transplutoniums in the conversion chains for 252cf production is calculated and analyzed. The production and conversion efficiency of the chains with different target nuclides and neutron spectrums including fast, thermal and moderated neutron spectrums is computed. The conversion paths for 252cf under different neutron spectrums and neutron flux levels are analyzed. The multi-group depletion program STEP1.0 based on linear chain method is used to calculate the nuclide densities. The model of the reactor is simulated by the transport-burnup coupling system based on the STEP1.0 and MCMG-II for the detailed description of the target designs in the fast reactor. The differences of the conversion efficiency caused by the neutron spectrums can be qualified more accurately with the multi-group cross section calculations. The higher conversion efficiency than the typical thermal neutron spectrum can be obtained by some moderated spectrum optimized for absorption reaction in fast reactor (CEFR China Experimental Fast Reactor) through the irradiation of some transplutoniums (242Pu, 241Am, 244Cm). The neutron flux of large scale demonstration fast reactor or prototype fast breeder reactor is usually higher than the experimental fast reactor (CEFR) or reactor device. China is also developing the demonstration fast reactor (CFR China Fast Reactor) which will have a maximum neutron flux higher than 5.0E15 n/(cm2·s). Therefore, the 252cf production will be more practical and efficient with CFR.

Numerical approach to study the thermal-hydraulic characteristics of Reactor Vessel Cooling system in sodium-cooled fast reactors

The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled fast reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

高速増殖実証炉概念の保守補修性向上策の検討:（２）原子炉構造の補修性改善方策の検討

高速増殖実証炉概念の保守補修性向上策の検討:（２）原子炉構造の補修性改善方策の検討

改良9Cr-1Mo鋼の高温高サイクル疲労特性評価

Mod.9Cr-1Mo steel has been widely used for elevated temperature components e.g. power boiler, chemical plant because of its superior strength properties at elevated temperatures. However some failure accident has recently happened in the weldments of power boiler steam pipes and it's due to the creep damage so-called Type IV damage. Many research have being carried out to prevent Type IV failures. On the other hand, Mod.9Cr-1Mo steel is also a candidate material for demonstration fast breeder reactor in Japan. A flow induced vibration is assumed in the piping due to the coolant liquid sodium flowing with high velocity inside the pipe. Therefore, the high cycle fatigue properties of this material at elevated temperatures have to be clarified to evaluate the integrity of the piping against a flow induced vibration. In this research, a series of high cycle fatigue tests were performed at room temperature, 400°C and 550°C. As a result, the fatigue limit appeared at room temperature and 400°C, while it didn't appear within 108 cycles at 550°C. This may be due to the suppression of crack initiation by the oxide film formed on the specimen surface at 550°C. Life prediction was also discussed beyond 108cycles.

A design study on Pb-208 cooled compact CANDLE burning reactor for future nuclear energy supply

A compact pool-type Pb-208 cooled CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) with a thermal power rating of 125 MWth is considered for the future nuclear energy supply. Natural Pb consists of Pb-204, Pb-206, Pb-207 and Pb-208. Pb-208 has a small capture and inelastic-scattering cross-section, which makes it possible to reduce neutron capture by coolant and to make neutron spectrum harder. In case of Pb-208 coolant instead of natural Pb, the core height and radius are reduced to 1.5 m and 1 m, respectively. The effective multiplication factor of the core, k eff, could be increased from k eff = 0.984 of natural Pb up to k eff = 1.006. For increasing natural circulation head, coolant velocities in each core zone are adjusted by orifice at the core inlet position. The reactor vessel height is equal to that of a typical loop-type demonstration FBR vessel to obtain natural circulation head.

Evaluation method for structural integrity assessment in core disruptive accident of fast reactor

A procedure to evaluate the structural integrity of a reactor vessel under CDA condition has been developed based on simulative experiments under CDA loading using 1/20- and 1/10-scale steel models of the demonstration FBR in Japan (DFBR). For the experiments, a constant pressure source has been developed using a gas accumulator. Simulation analyses were also conducted using the dynamic analysis program AUTODYN-2D, which can simulate fluid-structure interaction and non-linear material behaviour in two-dimensional geometry. The actual dynamic material properties, whose strain rates were from 0.1 to 10 s −1, were used in the analyses. The pressure history and the dynamic response of structures were compared between experimental results and the analysis, and the appropriateness of the numerical analytical method for the CDA problem was confirmed. In both experimental and numerical approaches, the effect of internal structures has been investigated, and a drastic strain reduction was observed owing to the energy absorption by internal structures. Furthermore, material tests and element tests including weld joint tests and burst tests were carried out to define a rationalised strain limit under multi-axial dynamic loading condition. Using the developed evaluation procedure, a preliminary evaluation for DFBR was carried out and the robustness of the reactor boundary in DFBR against CDA loading was also confirmed.

Monte Carlo Analysis of Streaming Effect through Annular Gaps around Piping Legs in Sodium Pool to Secondary Sodium Activation of the Demonstration Fast Breeder Reactor

A top-entry loop-type reactor is one of the favorable options for Demonstration Fast Breeder Reactor (DFBR) which is now under development in Japan as a part of conceptual design study. Annular gaps around top-entry piping raise neutron streaming in upward direction in the sodium pool of reactor vessel. It enhances neutron flux level around decay heat exchangers (DHX's) in the pool and at the penetrations in primary biological shield for piping connecting to the vessels of intermediate heat exchangers (IHX's), and consequently enhance secondary sodium activations in these heat exchangers, which is one of the main issues for shielding design. In this study, three-dimensional Monte Carlo analysis method was applied to make precise evaluation of the neutron streaming effect with combination of some techniques for reduction of statistical error within reasonable CPU time. It is established that the contribution of neutron streaming to the secondary sodium activation in DHX's hardly reaches a level of 30% and that design analysis with two-dimensional discrete ordinates method gives conservative evaluation of the secondary sodium activations in DHX's and IHX's.

Monte Carlo Analysis of Streaming Effect through Annular Gaps around Piping Legs in Sodium Pool to Secondary Sodium Activation of the Demonstration Fast Breeder Reactor.

A top-entry loop-type reactor is one of the favorable options for Demonstration Fast Breeder Reactor (DFBR) which is now under development in Japan as a part of conceptual design study. Annular gaps around top-entry piping raise neutron streaming in upward direction in the sodium pool of reactor vessel. It enhances neutron flux level around decay heat exchangers (DHX's) in the pool and at the penetrations in primary biological shield for piping connecting to the vessels of intermediate heat exchangers (IHX's), and consequently enhance secondary sodium activations in these heat exchangers, which is one of the main issues for shielding design. In this study, three-dimensional Monte Carlo analysis method was applied to make precise evaluation of the neutron streaming effect with combination of some techniques for reduction of statistical error within reasonable CPU time. It is established that the contribution of neutron streaming to the secondary sodium activation in DHX's hardly reaches a level of 30% and that design analysis with two-dimensional discrete ordinates method gives conservative evaluation of the secondary sodium activations in DHX's and IHX's.

Shaking table tests on failure characteristics of base isolation system for a DFBR plant

The building of a demonstration fast breeder reactor (DFBR) plant in Japan is planed to be base isolated in the horizontal direction. To verify the seismic safety of the isolation system, a series of shaking table tests was conducted using a reduced scale model with three types of base isolation system, natural rubber bearing with steel damper (NRB+SD), lead rubber bearing (LRB), and high damping rubber bearing (HRB). The results of these tests showed NRB+SD, LRB and HRB were within the stable domain (not hardening) at 1×S2 (maximum acceleration 3.80 m s −2) input, and were nearly hardening at 2×S2 input. None of the rubber bearings broke at 3×S2 input, which was the design limit. All these bearings broke at over 4×S2 input.

Evaluation Procedures for Irradiation Effects and Sodium Environmental Effects for the Structural Design of Japanese Fast Breeder Reactors

In the structural design of fast breeder reactors, irradiation effects and sodium environmental effects on structural materials have to be taken into account. In this paper, firstly, an evaluation procedure for irradiation effects on the mechanical properties of 316FR (FBR Grade 316 stainless steel), which is a newly developed stainless steel for the Japanese demonstration fast breeder reactor, is proposed. The procedure gives a limit of accumulated fast neutron fluence E>0.1 MeV as a function of temperature, so that the minimum tensile fracture elongation of 10 percent, which is the threshold for material to stay ductile, is maintained. Furthermore, the procedure determined a creep life reduction factor and a creep rate increase factor as a function of accumulated thermal neutron fluence E<0.4 eV, within the limitation of the accumulated fast neutron fluence, to account for the creep life reduction and the increase of creep rate due to irradiation. Secondly, an evaluation procedure for sodium environmental effects on the integrity of 316FR and modified 9Cr-1Mo steel was proposed. It gave a corrosion allowance as a function of temperature, oxygen content, and service time, based on corrosion tests. It determined that no correction factors that correspond to sodium environment on design allowable stresses, etc., are needed, because no adverse effects of sodium on the mechanical properties of 316FR and modified 9Cr-1Mo steel were to be expected in the service conditions of FBRs. Both the procedures have been incorporated into the Japanese Elevated Temperature Structural Design Guide for Demonstration Fast Breeder Reactor.

Elevated-Temperature Mechanical Properties of an Advanced-Type 316 Stainless Steel1

Type 316FR stainless steel is a candidate material for the Japanese demonstration fast breeder reactor plant to be built in Japan early in the next century. Like type 316L(N), it is a low-carbon grade of stainless steel with a more closely specified nitrogen content and chemistry optimized to enhance elevated-temperature performance. Early in 1994, under sponsorship of The Japan Atomic Power Company, work was initiated at Oak Ridge National Laboratory (ORNL) aimed at obtaining an elevated-temperature mechanical-properties database on a single heat of this material. The product form was 50-mm plate manufactured by the Nippon Steel Corporation. Data include results from long-term creep-rupture tests conducted at temperatures of 500 to 600°C with test times up to nearly 40.000 h, continuous-cycle strain-controlled fatigue test results over the same temperature range, limited creep-fatigue data at 550 and 600°C, and tensile test properties from room temperature to 650°C. The ORNL data were compared with data obtained from several different heats and product forms of this material obtained at Japanese laboratories. The data were also compared with results from predictive equations developed for this material and with data available for types 316 and 316L(N) stainless steel.

Shielding Design Study of the Demonstration Fast Breeder Reactor (2) -Shielding Design on the Basis of the JASPER Analysis-

Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR.

Radiation Shielding for Fission Reactors

Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium.

Shielding Design Study of the Demonstration Fast Breeder Reactor (1) -Evaluation of the JASPER Experiment and Its Analytical Results -

The JASPER Experiment had been conducted to obtain useful information on FBR shielding analysis accuracy. Results obtained from the post-experimental analyses are summarized in this paper. Both characteristics for bulk shielding attenuation and streaming of neutron through configurations consisting of several materials including B4C, stainless steel, sodium and etc. were clarified and the analysis accuracy confirmed. The shielding analysis system for fast reactors has been improved and verified. These experimental data and analytical results were reviewed from a viewpoint of the applicability to the shielding design analysis of the DFBR. Useful information to be utilized in the design study of the DFBR is accumulated through those activities.

Design study of the demonstration FBR in Japan

The objective of the plant design study Phase 2, conducted by the Japan Atomic Power Company since 1997 for 3 years, is to accomplished a plant overall concept of the Demonstrative FBR (DFBR) that has economical potential toward commercialization and offers high reliability to plant operators not to cause a long unexpected shutdown resulting from a trouble, i.e., sodium leakage or fires. This has been successfully achieved by establishment of a plant overall design of 672 MWe consisting of the reactor system with drastically simplified internals, the compact and double walled coolant boundaries, the well rationalized fuel handling system, the BOP systems introducing up-to-date LWR equipment, and the compact reactor building. The plant construction cost has been estimated based on the quantity of materials to be about 130 % on the bases of a 1000 MW LWR, which is well contented with the requirement. The DFBR plant concept, having economical potential toward commercialization, safety and reliability, has been established in the plant design study Phase 2.

Fast reactor decay heat removal: approach to the safety system design in Japan and Europe

In the framework of the cooperation on fast reactor between the European and Japanese electrical utilities, the design companies responsible for the demonstration fast breeder reactor (DFBR) in Japan and the European fast reactor (EFR) have performed a comparative evaluation of the safety qualified decay heat removal systems of the two reactor designs. At the level of overall safety and concept design there is an obvious similarity between the two DHR systems. In both cases heat is removed directly from the reactor vessel primary sodium by systems designed according to a similar deterministic methodology, with a probabilistic assessment performed to demonstrate achievement of the required reliability. Nevertheless, the evaluation revealed a number of differences resulting from different national practices. These include the application of diversity and redundancy philosophy, the extent of passivity taken into account, the consequences of postulated maintenance outage on the design and the decay heat curve.

Analytical Model for Self-Excited Vibration of an Overflow Flexible Plate Weir

An analytical model for the self-excited vibration of an overflow flexible weir as observed in the French demonstration fast breeder reactor, Super Phenix-1, is proposed. The instability condition was derived for the case in which the plate vibrates at the frequency of the downstream tank sloshing. In this analysis, the flexible plate weir is modeled as a simply supported-free-simply supported-clamped rectangular plate. Eigenfunction expansions were applied for analyzing both plate vibrations and the downstream tank sloshing. The effect of an overflow liquid is formulated based on the assumption that the momentum change due to the collision of an overflow liquid partly transmitted to the pressure rise on the free surface. As a result, the characteristic equation of the system yielding the theoretical stability boundary was obtained. The stability boundary thus derived agreed well with experimental results.

Further Evaluation of Creep-Fatigue Life Prediction Methods for Low-Carbon Nitrogen-Added 316 Stainless Steel

Low-carbon, medium-nitrogen 316 stainless steel is a principal candidate for a main structural material of a demonstration fast breeder reactor plant in Japan. A number of long-term creep tests and creep-fatigue tests have been conducted for four products of this steel. Two representative creep-fatigue life prediction methods, i.e., time fraction rule and ductility exhaustion method were applied. Total stress relaxation behavior was simulated well by an addition of a viscous strain term to the conventional (primary plus secondary) creep strain, but only the letter was assumed to contribute to creep damage in the ductility exhaustion method. The present ductility exhaustion approach was found to have very good accuracy in creep-fatigue life prediction for all materials tested, while the time fraction rule tended to overpredict failure life as large as a factor of 30. Discussion was made on the reason for this notable difference.

Evaluation of Creep-Fatigue Life Prediction Methods for Low-Carbon Nitrogen-Added 316 Stainless Steel

Low-carbon, medium-nitrogen 316 stainless steel is a principal candidate for a main structural material of a demonstration fast breeder reactor plant in Japan. A number of long-term creep tests and creep-fatigue tests have been conducting for two heats of the steel. Two representative creep-fatigue life prediction methods, i.e., time fraction rule and ductility exhaustion method were applied. An introduction of a simple viscous strain term improved the description of stress relaxation behavior and only the conventional (primary plus secondary) creep strain was assumed to contribute to creep damage in the ductility exhaustion method. The present ductility exhaustion approach was found to have very good accuracy in creep-fatigue life prediction, while the time fraction rule overpredicted failure life as large as a factor of 30.

The present status of DFBR development in Japan

The first demonstration fast breeder reactor (DFBR-1) is being developed in Japan, as the essential step toward the commercial deployment of FBR around 2030. The construction of the DFBR-1, which follows the experimental FBR (JOYO) and the prototype FBR (MONJU), is scheduled for the early 2000s. The Japan Atomic Power Company (JAPC) has been conducting design studies of the DFBR-1 and relevant R&D since 1986, under a contract awarded by the nine electric companies and the Electric Power Development Co., Ltd. The electric utilities have settled the major design specifications of the DFBR-1 plant as the top-entry loop type system with an output of 660MWe. Based on these major design specifications, JAPC has been conducting the optimization design study of the DFBR-1 since FY 1994 in order to establish the overall consistent plant design with construction cost reduction below 1.5 times that of LWRs on 1,000 MWe basis and comfirm the licensability of the DFBR-1 design. This study will complete by the end of FY 1996 This paper mainly outlines the design concepts of DFBR-1 resulting from the optimization design study of DFBR-1.