A conceptual design study on innovative FBR

An Actinide Recycle Reactor (ARR) with ductless fuel assemblies and mixed nitride fuel is studied in accordance with an Advanced Fuel Recycle System. The core is designed so that yield more economical efficiencies (high breeding ratio and high burnup), safety aspects (high Doppler reactivity coefficient, low void reactivity coefficient and reactor dynamic characteristics) in comparison with mixed oxide or metal fuel on a suitable condition. Preliminary calculations about key parameters of the core design performances had been done to compare with mixed oxide or metal fuel. Results that the mixed nitride fuel with a sodium bond and ZrH has promising capacity.     

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.     

Actinide recycling by pyro-process with metal fuel FBR for future nuclear fuel cycle system

Pyro-metallurgical technology is one of potential devices for future nuclear fuel cycle. Not only economic advantage but also environmental safety and strong resistance for proliferation are required for the fuel cycle. In order to satisfy the requirement, actinides recycling applicable to LWR and FBR cycles by pyro-process has been developed since more than ten years in CRIEPI. The main technology is electrorefining for U and Pu separation and reductive-extraction for TRU separation, which can be applied on oxide fuels through reduction process as well as metal fuels. The application of this technology on separation of TRU in HLLW through chlorination could contribute to the improvement of public acceptance on the geologic disposal.

The main achievements are summarized as follows:

• -|The elemental technologies, such as electrorefining, reductive extraction, injection casting and salt waste treatment and solidification, have been developed successfully with lots of experiments

• -|The fuel dissolution into molten salt and uranium recovery on solid cathode for electrorefining have been demonstrated by engineering scale facility in Argonne National Laboratory by using spent fuels and in CRIEPI by uranium tests.

• -|Single element tests, using actinides, showed the Li reduction to be technically feasible, remaining the subjects of technical feasibility on multi-elements system and on effective recycle of Li by electrolysis of Li₂O.

• -|Concerning on the treatment of HLLW for actinide separation, the conversion to chlorides through oxides has been also established through uranium tests.

• -|It is confirmed that more than 99% of TRU nuclides can be recovered from the high level liquid waste by TRU tests

• -|Through these studies, the process flow sheets for reprocessing of metal and oxide fuels and for partitioning of TRU separation have been established.

The subjects to be emphasized for further development are classified into three categories, that is, process development (demonstration), technology for engineering development, and supplemental technology.

The metal fuel FBR has a high potential for recycling actinides by integration with pyro-reprocessing. Alloys of U-Pu-Zr with minor actinides are investigated from points of fuel properties. The miscibility and other characteristics suggest that the maximum content up to ca. 5 wt% of minor actinides is allowable in the matrix. Nine pins of metal fuel including minor actinides are ready for irradiation at Phenix fast reactor.     

Compatibility of FBR materials with sodium

In order to incorporate a procedure for the evaluation of the sodium environmental effects on core and structural materials into the elevated temperature structural design guide lines for fast breeder reactors, R&D on the sodium compatibility of the materials has been in progress in Japan Atomic Energy Agency. This paper reviews corrosion behavior in the sodium of conventional austenitic and ferritic steel. Simultaneously, the corrosion and mechanical properties of the materials for advanced FBRs, 12Cr steel and ODS steels are summarized, including the results of recent research.     

Operating temperatures for an FBR

Structural mechanics aspects related to operating temperatures of a typical pool type 500 MWe fast breeder reactor are discussed. The critical high temperature components are analysed in detail based on elastic, inelastic and viscoplastic deformation theories, and life is predicted in accordance with the rules of design code RCC-MR 87. Analysis indicates that the control plug is the most critical component in the reactor which limits the reactor outlet temperature to 820 K with a temperature rise of 160 K across the core.     

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.     

Scenario with inert matrix fuel: a specific study

In this paper, we describe a strategy study concerning the future of the French nuclear energy infrastructure, with a scenario involving reactors loaded with inert matrix fuel. We select the problem of the inventory control of minor actinides by target introduction into fast reactors. Added to pressurized water reactors in the French nuclear infrastructure, this scenario permits one to balance plutonium and minor actinide production and consumption and to obtain a substantial reduction of the radiological impact compared to a non-reprocessing fuel scenario on a one million year scale.     

Applications of probabilistic fracture mechanics to FBR components

A probabilistic fracture mechanics code which evaluates fracture probability of a plate model with an elliptical surface crack caused by creep-fatigue crack growth has been developed. The code named PCCF (Probabilistic Fracture Mechanics Code for Creep-Fatigue Crack Growth) uses simplified methods of C^* and J-integral for evaluation of creep-fatigue crack growth and a stratified sampling method for two input variables to improve the solution convergency. According to the test analyses focused on an applied stress level using PCCF code, leak probability is sensitive to a stress level and increases rapidly when an applied stress is close to a yield stress level.     

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.     

Development of FBR piping bellows joints in Japan

For the practical use of fast breeder reactors (FBRs) reduction of construction costs is one of the most important factors. If the long and winding route of piping systems (needed to absorb the thermal expansion) can be shortened and simplified, sharp reductions in related apparatus, equipment and reactor building etc. can be expected (especially in the case of loop type FBRs). The use of bellows joints, which possess good ability to absorb thermal expansion, is one of the best means of shortening the piping system. From 1983 to 1988, the Power Reactor and Nuclear Fuel Development Corporation promoted extensive research and development on FBR piping bellows joints, which covered areas such as strength evaluation methods, manufacturing and inspection techniques, maintenance and repair techniques, investigation of safety logic etc. The purpose of that work was to ensure that the application of bellows joints to FBR main piping systems was a technical and practical possibility. The conclusion was that the use of FBR piping bellows joints was feasible. Consequently, both draft structural design rules and draft manufacturing and maintenance rules were formulated based on the results. This paper presents a summary of the program and the results of the research and development.     

快堆材料试验回路的静力分析

该回路有多个环形管系、布置紧凑、运行温度高。原设计未考虑支撑,因此“安全问题”比较突出。根据SDGJ6-78规定。利用PSDP程序,对此回路进行了一次、二次或一次加二次应力验算,改进了原设计,并对支撑作了合理布置。计算表明在给定条件下运行是安全的。     

Effect of minor actinides doping on plutonium produced in large-scale FBR blanket

The present study focuses on the effect of minor actinides (MAs) addition into the FBR blanket as ways of increasing fraction of even-mass-number plutonium isotopes, especially ²³⁸Pu, aiming at enhancing the proliferation resistance of plutonium produced in the blanket. The MA loading potential to enhance the proliferation resistance of plutonium is investigated, with considering actual design constraints on the fuel decay heat from the fuel handling and fabrication points of view, as MAs considerably generate decay heat. It reveals that depending on doping quantity of MAs, it is possible to denature produced plutonium by MA transmutation. MA addition in the blanket gives a significant increment in ²³⁸Pu fraction of generated plutonium but less effect on other even-mass-number plutonium isotopes. However, it is important that MA compositions should be adequately controlled to satisfy both the proliferation resistance requirements and the decay heat constraints for fuel handling.     

Advanced evaluation of thermal ratchetting of FBR components

A new prediction method for the thermal ratchetting of a cylinder subjected to an axially moving temperature distribution is proposed in this paper. This ratchetting is quite different from the conventional Bree-type ratchetting, and an advanced evaluation method has been required in the structural design of FBR components. The proposed method considers the work hardening of actual materials for FBR components. Firstly the basic scheme of the prediction method is shown, and secondly the application procedure to the actual design is shown. Predicted results by using this method coincide well with experimental results, when compared with the case by using the previous method.     

Method of Core Thermal Calculation for Multichannel on Sodium Cooled FBR

A few concrete conditions for the existence of the time eigenvalue of fast multiplying systems are derived from the integral transport equation. In order that the real time eigenvalue may exist, there are limitations to the composition and the size of the systems. It may be considered, with very good approximation, that a necessary condition for the existence of the real time eigenvalue is that the infinite multiplication constant is not less then unity. With respect to the size of the system, if half the minimum chord length is not less than three times the inverse of {Σ _s (ν)+ νΣ _f (ν)}_min, the system is not too small for the existence of the real time eigenvalue. These conditions can be applied to the planning and the analysis of pulsed neutron experiments on a fast multiplying system.     

氧化物燃料快堆的事故瞬态分析

利用在核电厂动态仿真器ＤＳＮＰ上开发完成的仿真程序ＯＸＳＹＳ，分析计算了氧化物燃料钠冷快堆ＣＲＢＲＰ在超功率和失流事故瞬态下的响应特性，所得结果与国外系统分析程序ＳＳＣ－Ｌ、ＦＰＲＥ－Ⅱ的相应计算结果符合较好。     

Correlations among FBR core characteristics for various fuel compositions

In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations.     

Compatibility of FBR structural materials with supercritical carbon dioxide

A key problem in the application of a supercritical carbon dioxide (CO₂) turbine cycle to a fast breeder reactor is the corrosion of structural materials brought about by supercritical CO₂ at high temperatures. In this study, long-term (8000 h) compatibility tests on candidate materials, two high-chromium martensitic steels (12Cr- and 9Cr-steels) and an austenitic stainless steel (316FR), were performed at 400-600 °C in supercritical CO₂ pressurized at 20 MPa, and corrosion allowances for the steels were proposed for application to preliminary reactor design.Although high temperature oxidation was measured in all steels, the behavior differed greatly. For martensitic steels, weight gain exhibited parabolic growth as exposure time increased at each temperature. Neither exfoliation of the oxide nor the breakage was observed during the 8000 h of exposure. The corrosion behavior was equivalent to that seen in supercritical CO₂ at 10 MPa, and it was confirmed that no effects of CO₂ pressure were present under the CO₂ turbine cycle operation conditions. Based on the results, corrosion allowances for temperature-dependant parabolic growth were proposed. For 316FR steel, weight gain was significantly lower than that of martensitic steels, with a maximum value of 6.2 g/m² at 600 °C for 8000 h. Since no dependency of temperature and immersion time on weight gain such as the martensitic steels was noted, corrosion allowances proportional to time was proposed. Estimated corrosion allowances for the martensitic and austenitic steels were 380 μm and 220 μm, respectively, for reactors, whose design life is rated at 60 years.     

快堆堆芯组件抗震分析方法研究

快堆堆芯组件的抗震安全评价是国家核安全局审评的重要内容,也是中国实验快堆(CEFR)取得装料许可证的必要条件之一.本文使用有限元分析软件CAST3M[1],首先对堆芯单排组件在空气中及液钠中进行抗震响应分析,经比较可以看出,考虑流固耦合作用对计算结果的影响非常显著.在此基础上,对液钠及组件进行单独建模,考虑流固耦合的作...