Analysis of the Presence of Vapor in Residual Heat Removal System in Modes 314 Loss-of-Coolant Accident Conditions Using RELAP5

The Westinghouse Nuclear Safety Advisory Letter NSAL-09-8 investigated the possibility of presence of vapor in RHR （residual heat removal） system in modes 3/4 LOCA （loss-of-coolant accident） conditions. This concerns the Westinghouse standard three-loops plant for which the RHR is the low pressure part of the St （safety injection）. In some cases one or both RHR trains may become inoperable for SI function. As a response to this letter, Westinghouse Electric Belgium is providing RELAP5 analyzes for Westinghouse NSSS （nuclear steam supply system） European plants to assess the thermal hydraulic behavior of the RHR suction piping system for ECCS （emergency core cooling system） initiation events postulated to occur during startup/shutdown operations. Several concerns including condensation induced water hammer and voiding at the RHR pump have been investigated. As a conclusion, the analysis allowed to define the bounding hot leg temperature conditions under which both RHR trains remain safely operable. These bounding conditions are then implemented by the customer in their OPs （operating procedures） to achieve safe operations and successful accident management.

Experimental study of primary and secondary side coupling natural convection heat transfer characteristics of the passive residual heat removal system in AP1000

Passive residual heat removal heat exchanger(PRHR HX),which is a newly designed equipment in the advanced reactors of AP1000 and CAP1400,plays an important role in critical accidental conditions.The primary and secondary side coupling heat transfer characteristics of the passive residual heat removal system(PRHRS)determine the capacity to remove core decay heat during the accidents.Therefore,it is necessary to investigate the heat transfer characteristics and develop applicable heat transfer formulas for optimized design.In the present paper,an overall scaled-down natural circulation loop of PRHRS in AP1000,which comprises a scaleddown in-containment refueling water storage tank(IRWST)and PRHR HX models and a simulator of the reactor core,is built to simulate the natural circulation process in residual heat removal accidents.A series of experiments are conducted to study thermal-hydraulic behaviors in both sides of the miniaturized PRHR HX which is simulated by 12 symmetric arranged C-shape tubes.For the local PRHR HX heat transfer performance,traditional natural convection correlations for both the horizontal and vertical bundles are compared with the experimental data to validate their applicability for the specific heat transfer condition.Moreover,the revised natural convection heat transfer correlations based on the present experimental data are developed for PRHR HX vertical and lower horizontal bundles.This paper provides essential references for the PRHRS operation and further optimized design.

Study of Accident Progression in Unsealed WWER-1000/V320 Reactor during Maintenance

This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating state with unsealed primary circuit by removing the MCP head. The reference nuclear power plant is Unit 6 at Kozloduy NPP (KNPP) site. RELAP5/ MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of KNPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The plant modifications performed in frame of modernization program have been taken into account for the investigated conditions for the unsealed primary circuit. The most specific in this analysis compared to the analyses of NPP accidents at full power is the unavailability of some important safety systems. For the purpose of the present investigation two scenarios have been studied, involving a different number of safety systems with and without operator actions. The selected initiating event and scenarios are used in support of analytical validation of Emergency Operating Procedures (EOP) at low power and they are based on the suggestions of leading KNPP experts and are important in support of analytical validation of EOP at low power.

Application of Finite Difference Method to the Analysis of HTTR Reactor Cavity Natural Convection

An attempt is made to investigate the ability of the revised K FIX code to the solution to natural convection. The new experimental data of the benchmark problem of the IAEA coordinated research program (CRP 3) on heat transport and afterheat removal for GCRs under accident conditions provided by JAERI are used to calculate nitrogen natural convection in the pressurized vessel and air natural convection in the reactor cavity by using this revised code. Based on analysis, a refined mesh is used to solve the differential equations so as to get more detailed and more accurate result. The obtained velocity profiles are consistent with the result of TRIO EF code and the result of Bechtel laboratory. It can be drawn that the revised K FIX code can be used to solve this kind of problems.