증기발생기튜브누설에 대한 원전시뮬레이터 및 검증코드 비교

Abstract

A nuclear power plant simulator is a training device for operator license examination, requalification, and verification of emergency operating procedures under the circumstance similar to practical plants. Licensees in the U.S. are required to submit the results of the simulator performance test on the basis of the NRC (Nuclear Regulatory Commission) form 474 once every four years in accordance with 10CFR55.45 (b). In case of the Korea, licensee makes its own simulator performance tests on each training center, but independent verification of performance test results is not performed by any organization. The pressurizer (PZR) pressure/level and steam generator (SG) pressure/level main parameters associated with the SGTR were replicated in simulator and computer code. The PZR pressure obtained with the simulator in the initial accident decreased relatively a little. The reason is that the PZR back-up heaters were turned on to maintain the pressure of the PZR by the pressurizer pressure control system (PPCS). It was estimated that the increasing SG level from RCS leakage could increase SG pressure with slow feedwater control system (FWCS) and turbine control system (TCS) response in the simulator. The broken SG pressure by RELAP5/MOD3 code decreases until reactor and turbine trip as the RELAP5/MOD3 code could not adopt the FWCS and TCS in detail. For that reason, the SG pressures between the simulator and RELAP5/MOD3 code show different direction until reactor trip. After the reactor trip, the SG level of two methods is in direct opposition to each other. The opening period of the main steam safety valve (MSSV) in the simulator was relatively late, compared with that of RELAP5/MOD3 code. Because of this, the simulator SG inventory released to the atmosphere is smaller than the SG level increase by the ruptured tube. On the other hand, loss of feedwater flow and relatively long time to the MSADV stuck open could effect SG level decrease in the computer code. By comparing the changes in pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behaviors of the SG pressure and level that are important parameters to diagnose the accident were a little different. It was because RELAP5/MOD3 code did not reflect the major control systems in detail, such as FWCS, TCS and PPCS. The different behaviors of the SG level and pressure in this study need an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future.