Kenichi Kurisaka

Technological Feasibility of Two-Loop Cooling System in JSFR

An advanced large-scale sodium-cooled fast reactor named JSFR adopts an innovative two-loop cooling system. This cooling system design raises major technological issues: hydraulic and structural integrity due to the increase in one-loop coolant flow rate, safety design against the break or failure in one-loop piping, and ensuring the reliability of the decay heat removal system (DHRS). The present paper describes the investigation of the piping structural integrity due to flow-induced vibration using a 1/3-scale hot-leg piping test. The structural integrity of the hot-leg piping in the JSFR design has been confirmed by a flow-induced vibration analytical methodology, verified with the experimental data. Additional experimental results have revealed that hydraulic issues including gas entrainment and vortex cavitation could be prevented by some design measures. By applying appropriate safety design, the two-loop system has been confirmed to be valid against the break or failure in one-loop piping by a safety evaluation in this study. The DHRS with natural circulation is designed in conformity with the two-loop system by introducing adequate safety designs. In this paper, the validity of this DHRS is given by a probabilistic safety assessment and safety evaluation.

ATWS Frequency Evaluation of FBR Monju

MONJU is a sodium-cooled, loop-type prototype fast breeder reactor which can supply 280MW of electricity. The frequency of anticipated transient without scram (ATWS) of MONJU that used to be based on conservative assumption was re-evaluated by appropriately providing the event sequence from the result of plant response analysis. As a result, detailed evaluation of the ATWS frequency according to the result of the plant response analysis and reviewing event sequence scenarios have enabled realistic ATWS frequency estimation.Copyright

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

ABSTRACT The accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10−5 and 1×10−6 per plant operating year, respectively, which were selected based on the IAEAs safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.

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.

Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences.

Assessment of FBR MONJU Accident Management Reliability in Causing Reactor Trips

This paper describes a method and application of quantitatively evaluating Accident Management (AM) reliability upon a reactor trip failure for the MONJU fast breeder reactor using a PSA technique. The present method comprises an allowable time estimation that is based on plant transient response analysis using the Super-COPD code that was developed for use in best estimates of the plant dynamics of MONJU and in estimating failure probability of operators actions in AMs within the allowable time based on time records obtained from simulator training. Application of this method to MONJU resulted in the estimation that the allowable time for an unprotected loss-of-heat sink event would be more than the longest observed time of 326 s. The corresponding operation failure probability would be less than 0.1 even after taking the uncertainty into consideration. Combining this with a level 1 PSA revealed that the total frequency of core damage accompanying a reactor trip failure at MONJU could be decreased by at least 50 percent due to the reactor trip AM.

Development of Probabilistic Risk Assessment Methodology of Decay Heat Removal Function Against Combination Hazard of Low Temperature and Snow for Sodium-Cooled Fast Reactors

A probabilistic risk assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is aimed at developing a PRA methodology for the combination of low temperature and snow for a sodium-cooled fast reactor which uses the ambient air as its ultimate heat sink to remove decay heat under accident conditions. The annual exceedance probabilities of low temperature and of snow can be statistically estimated based on the meteorological records of temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., a clogged intake and/or outtake for a DHRS and for an emergency diesel generator, an unopenable door on necessary access routes due to accumulated snow, failure of intake filters due to accumulated snow, and possibility of water freezing in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around intakes and outtakes caused by loss of the access routes, this study has investigated effects of electric heaters installed around the intakes and outtakes as an additional countermeasure. By using the annual exceedance probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence caused by a snow and low temperature combination hazard is the failure of the electric heaters and the loss of the access routes for snow removal due to low temperature and snowfall which last for a day, and daily snowfall depth of 2 m/day.

Updating of local blockage frequency in the reactor core of SFR and PRA on consequent severe accident in Monju

ABSTRACT Fuel subassemblies of sodium-cooled fast reactors (SFRs) are densely arranged and have high power densities. Therefore, the local fault has been considered as one of the possible initiating events of severe accidents. In the conventional analyses for the license of Japanese prototype SFR (Monju), according to the local fault evaluation under the condition of one sub-channel flow blockage in the analyses of design basis accident (DBA), it was confirmed that the pin failures were limited locally without severe core damage. In addition, local flow blockage of 66% central planar in the subassembly was historically investigated as one of the beyond-DBAs. However, it became clear that these deterministic analyses were not based on a realistic assumption by experimental studies. Therefore, probabilistic risk assessment on local fault which was initiated from local flow blockage was performed reflecting the state-of-the-art knowledge in this study. As a result, damage propagation from local fault caused by local flow blockage in Monju can be negligible compared with the core damage due to anticipated transient without scram or protected loss of heat sink in the viewpoint of both frequency and consequence.

Development of Probabilistic Risk Assessment Methodology Against Volcanic Eruption for Sodium-Cooled Fast Reactors

The objective of this paper is to develop a probabilistic risk assessment (PRA) methodology against volcanic eruption for decay heat removal function of sodium-cooled fast reactors (SFRs). In the v...

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