Uncertainty Evaluation of Anticipated Transient without Scram Plant Response in the Monju Reactor Considering Reactivity Coefficients within the Design Range

This paper describes the methods and results of an uncertainty evaluation of a significant plant response analysis of reactor trip failure events,specifically anticipated transients without scram in the Japanese prototype fast breeder reactor Monju.Unprotected loss of heat sink(ULOHS)has a relatively large contribution to the core damage frequency due to reactor trip failure.The uncertainty in the allowable time to core damage in this event has so far been estimated by considering the range of reactivity coefficients.There are some cases where it is considered that core damage will be avoided.Specifically,if the primary heat transport system(PHTS)pump inlet sodium temperature stays below 650℃for 1 h,the avoidance of core damage due to a ULOHS event is assumed.This is the temperature at which the probability of cavitation in the static pressure bearing begins to increase.In this study,a success scenario was investigated in two aspects:identification of influential input parameters and estimation of the probability of success.In the parameter identification,input parameters that satisfy the pump inlet temperature being below 650°C are clarified by treating the reactivity coefficients and reactor kinetics parameters as variables that can be taken to be within the design range.In the probability estimation,the results are fitted to a lognormal distribution function,from which the output variable was found to fall between 640 and 679℃with a probability of 90%,the probability of the temperature being 650℃or lower was 0.23,and the average and mode value was 659℃.

Numerical Simulations of Upper Plenum Thermal-Hydraulics of Monju Reactor Vessel Using High Resolution Mesh Models

In order to evaluate the effects of the FHs （flow holes） on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this model, it was mainly clear that in the 40% rated operational conditions, the shape of the FHs on the inner barrel did not change largely to the upper plenum thermal-hydraulics. The effect of the FHs on the honeycomb structure in the upper structure was also investigated in these calculations. The results indicated that the height of thermal stratification interface became lower than that evaluated from the test data.

Plant Dynamics Evaluation of a MONJU Ex-vessel Fuel Storage System during a Station Blackout

The prototype fast breeder reactor ＂MONJU＂ has an EVSS （ex-vessel fuel storage system） which consists mainly of an EVST （ex-vessel fuel storage tank） and an EVST sodium cooling system. EVST sodium cooling system consists of three independent loops. During the normal operation, the primary sodium in the EVST is circulated by natural convection and the secondary circulation in the EVST sodium cooling system is powered by electromagnetic pumps. When an SBO （station blackout） occurs, all the pumps and blowers are tripped. Therefore, it was necessary to evaluate the cooling ability by the natural circulation of sodium in the EVST sodium cooling system and air through the air cooler during the SBO. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an SBO were performed. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450 ~C. However, the structural integrity of the EVSS was maintained. The analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.

Unplanned Shutdown Frequency Prediction of FBR MONJU Using Fault Tree Analysis Method

In order to evaluate the operational reliability of Japanese FBR （fast breeder reactor） MONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicted using FTA （fault tree analysis） technique for the plant system model. The targeted devices are the following： PHTS （primary heat transport system）, SHTS （secondary heat transport system）, WS （water and steam system）, PPS （plant protection system） and PCS （plant control system）. In this paper, the frequency of automatic reactor trips was estimated by extracting and analyzing the important intermediate events and equipment failures covering all the derived fault trees of these systems. The analyses predicted 1.2/RY （reactor year） the value of unplanned shut down frequency by the internal factor of the system. The largest contributed event was function failure of SHTS accounting for 42.6% of total events followed by PHTS with 40.1%. The contribution factor of WS was only 4.4%.

日本文殊原型快堆堆芯出口腔室热分层现象数值模拟

本文利用商业CFD程序STAR-CCM＋，采用合理的网格生成技术及物理模型，对日本文殊原型快堆堆芯出口腔室建立近似1∶1的模型，模拟分析40％额定功率停堆过程中堆芯出口腔室的瞬态工况，获得腔室内较为完整的热分层进程。结果表明：停堆2 min后腔室内出现稳定热分层现象；10～21 min时热分层通过上升桶桶顶位置；10～140 min热分层处于上升筒顶端位置附近期间，腔室内流型不稳定；140 min后热分层完全处于上升桶顶，桶内流型稳定且接近于停堆前。模拟结果与实验数据对比表明，停堆初期4 min内两者符合较好，表明本文模拟方法适用于停堆工况堆芯出口腔室热分层进程模拟；之后模拟进程明显快于实验，分析其偏差主要来自模拟边界及结构与实际的差异。

日本文殊快堆紧急停堆后堆芯出口腔室瞬态工况模拟研究

基于日本文殊快堆停堆实验数据,完成了文殊快堆上升桶通流孔结构分别为直角、圆角下堆芯出口腔室内较完整的热分层进程模拟,并从热分层的形成、界面上升速度、温度梯度及通流孔钠流量比率等方面对热分层特点进行深入分析。结果表明,数值模拟结果与实验结果符合较好,在一定条件下,数值模拟可很好地预测钠冷快堆内整体热工水力行为。本文结果为建立一套用于预测钠堆内复杂瞬态工况的数值模拟方法积累了经验。

日本文殊快堆整体热腔室流场三维数值模拟

为研究日本文殊快堆一回路热腔室的热工水力特性,借鉴和消化国外快堆的设计经验,使用流体力学软件CFX对文殊快堆整体热腔室进行三维稳态数值模拟,得到其整体热腔室流场。文殊快堆全堆芯温度监测系统可为我国快堆小型化设计作技术准备。

Investigation of Thermal Expansion Model for Evaluation of Core Support Plate Reactivity in ATWS Event

Thermal expansion behavior was investigated in detail for evaluation of the core support plate expansion reactivity in the ULOHS(Unprotected Loss of Heat Sink)reactor trip failure event.The core support plate expansion reactivity plays an important role in the safety evaluation of the ULOHS event.In this paper,a possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype FBR(Fast Breeder Reactor)Monju.The reactor core expansion was simulated in a three-dimensional FEA(Finite Element Analysis)model of the RV(Reactor Vessel)considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model.It was found that the thermal expansion of the core was not restrained in the ULOHS event,although part of the core structure is mechanically restrained.