Yoshio Kamishima

Development of Advanced Loop-Type Fast Reactor in Japan

This paper describes the current status of the design study and the related research and development for an advanced loop-type fast reactor: the Japan sodium-cooled fast reactor (JSFR). First, the development targets and the major design requirements are established. Then, a cost-down approach in which developing innovative technologies is key to being competitive with another future energy source is discussed. Here, the development status of several innovative technologies such as a two-loop cooling system, reliable reactor system, simplified fuel-handling system, passive reactor shutdown system, mitigation measure against a core disruptive accident, and minor actinide-bearing oxide fuel core is described. Last, a review of JSFR development and the demonstration plan for the innovative technologies are discussed.

Seismic Isolation Design for JSFR

This paper describes the seismic design of Japan Sodium-Cooled Fast Reactor (JSFR), which includes the seismic condition, the seismic isolation system, and the seismic evaluation of the primary components. Since the design seismic loading is set out severely than ever since The Niigata-ken Chuetsu-oki Earthquake in 2007, an advanced seismic isolation system is aimed to reduce the seismic force loaded on the primary components of JSFR to be less than that of the previous seismic isolation system. The advanced seismic isolation system is developed by optimizing the performance based on the previous seismic isolation system considering the natural frequency of the primary components. The laminated rubber bearings thicker than the previous ones and oil dampers are adopted for the advanced seismic isolation system of SFR. The seismic evaluation of nuclear reactor components applying the advanced seismic isolation system is performed and its feasibility is confirmed.

Severe External Hazard on Hypothetical JSFR in 2010

Abstract Severe external hazards on the Japan Sodium-cooled Fast Reactor (JSFR) have been analyzed and evaluated. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. Integrity of the major components has been confirmed covering recent earthquake conditions. In the case of a tsunami, the seawater pumps for the component cooling water system (CCWS) could be damaged by the tsunami, since they are located at sea level. In the JSFR design with full natural convection decay heat removal systems (DHRSs) and an air-cooling emergency gas turbine, safety-grade components are independent of CCWS, and loss of CCWS does not affect reactor cooling. As a conservative case, hypothetical station blackout (SBO) has been evaluated. In the case of SBO, decay heat is removed by natural convection DHRS, but control of the air cooler (AC) damper is lost after the battery power is out. The analysis has revealed that freezing at one of three ACs could happen due to loss of automatic control of AC dampers. However, the time margin to protected loss of heat sink is evaluated to be >10 days. Manual control of the AC damper is also investigated. Transient analyses show that the AC dampers can be controlled manually adopting a simple operation procedure with sufficient operation time. Decay heat cooling in the case of collapse in all air stacks of AC has been evaluated. The result shows that decay heat could be removed maintaining air paths in two of three ACs by accident management. In conclusion, JSFR in the 2010 design version has enough external hazard toughness mainly thanks to passive safety features and a seismic isolation system.

Current Status of Conceptual Design Study Toward the Demonstration Reactor of JSFR

JAEA is now conducting “Fast Reactor Cycle Technology Development (FaCT)” project for commercialization before 2050s. A demonstration reactor for Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since FY2007 to determine referential reactor specifications for the next stage of design work of licensing and construction study. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. In this paper, the current status of the conceptual design study for the demonstration reactor plant is summarized.Copyright

Probabilistic fracture mechanics analysis for pipe considering dispersion of seismic loading

Studies on probabilistic fracture mechanics (PFM) have been performed for the quantitative integrity assessment of the pipes, which are used in nuclear power plants. The seismic load is one of the dominant loads for the failure assessment of the pipes. Its probabilistic dispersion, however, was not taken into account in the past PFM analysis. Authors have developed PFM analysis code, which accommodates the dispersion of the seismic load and performed parametric PFM analyses using it. The seismic stress has more effect on the break probability, but not for the leak probability. The earthquake, whose occurrence probability is less than 10 -5 /year has little effect on the break probability. The break probability is affected by the dispersion of the stress due to earthquake rather than the seismic hazard curve.

Study on Minimum Wall Thickness Requirement for Seismic Buckling of Reactor Vessel Based on System Based Code Concept

The minimum wall thickness required to prevent seismic buckling of a reactor vessel (RV) in a fast reactor is derived using the system based code (SBC) concept. One of the key features of SBC concept is margin optimization; to implement this concept, the reliability design method is employed, and the target reliability for seismic buckling of the RV is derived from nuclear plant safety goals. Input data for reliability evaluation, such as distribution type, mean value, and standard deviation of random variables, are also prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Minimum wall thickness required to achieve the target reliability is evaluated, and is found to be less than that determined from a conventional deterministic design method. Furthermore, the influence of each random variable on the evaluation is investigated, and it is found that the seismic load has a significant impact.

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (1) Overview

This paper is the first one of the series of four papers that describe ongoing activities in the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) on the elaboration of the System Based Code (SBC) concept aiming at its application to nuclear structural codes and standards. This paper includes a brief introduction to the SBC concept and the technical features of structural evaluation methodologies that are being developed for use in the SBC concept; Load and Resistance Factor Design based reliability assessment methods and the JSME guidelines for reliability evaluation of static components of sodium-cooled fast reactors. Also described is the ongoing collaboration of JSME and ASME at the Joint Task Group for System Based Code established in the ASME Boiler and Pressure Vessel Code Committee which is developing alternative rules for ASME B&PV Code Section XI Division 3, inservice inspection requirements for liquid metal reactor components.Copyright

Development of Limit State Design for Fast Reactor by System Based Code

Limit state design was newly developed in order to apply the margin exchange which is one of the innovative concepts of the System Based Code (SBC). It was shown that limit state design method is applicable to plant design instead of current design criteria. In this report, working example of a reactor vessel of a Fast Reactor subject to thermal load is conducted to demonstrate this concept. As the result allowable stress was increased by changing the acceptance criteria from current design criteria to limit state design criteria.Copyright

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (3) Guidelines on Structural Reliability Evaluation for FBR

This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the task force for the system based code in the Japan society of mechanical engineers in order to contribute to reducing differences in evaluated structural reliability by evaluators. They consist of five chapters, which are “General rules”, “Reliability evaluation”, “Failure scenario setting”, “Modeling”, and “Failure probability calculation”, respectively. In the chapter of “Reliability evaluation”, the general procedures of reliability evaluation are explained. Detailed procedures at each step are explained in the following chapters in the guidelines. Evaluation procedures for accumulation damage and crack propagation due to creep-fatigue interaction which is a typical degradation phenomenon in FBR are also provided in the appendix of the guidelines. In addition, some examples of random variables are prepared as standard input data for structural reliability evaluation.© 2014 ASME

Study on Minimum Wall Thickness Requirement of Reactor Vessel of Fast Reactor for Seismic Buckling by System Based Code

In this paper, minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling is discussed on the basis of the System Based Code (SBC) concept. One of key concepts of SBC is the margin optimization. To implement this concept, reliability design method is employed, and the target reliability for seismic buckling of reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation such as distribution type, mean value and standard deviation of random variable are prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Wall thickness needed to achieve the target reliability is evaluated, and as a result, it is shown that the minimum wall thickness can be reduced from that required by a deterministic design method.Copyright