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atw Vol. 62 (2017) | Issue 8/9 ı August/September No. Event Remarks ENVIRONMENT AND SAFETY 540 1 TLOFW Start 2 Reactor/TBN/RCP trip SG WR17.03MPa 6 Start of bleed No.5+637sec 7 SIP injection signal SIP injection start measurement has an uncertainty, but the values are acceptable for analysis. Overall results show that SPACE code have a reasonable agreement with target values in steady state analysis. The steady-state results are established for conducting a null transient calculation. 4 Results and discussions 4.1 Sequence of events The simulation scenario is OECD- ATLAS A3.1 test, which is a TLOFW with additional failures. It is assumed that 2 out of 4 safety injection pump (SIP) and half of the pilot-operated safety relief valve (POSRV) are unavailable to simulate additional failures during entire test period. However, passive components such as the safety injection tank (SIT) and the main steam safety valve (MSSV) are assumed to be fully available. Note that the supply of auxiliary feedwater is unavailable under the TLOFW event. The TLOFW with additional failures is initiated by terminating the P PZR
atw Vol. 62 (2017) | Issue 8/9 ı August/September | | Fig. 5. Fluid temperature of the system. primary coolant inventory was discharged through the POSRV and the SIP is actuated by the low pressurizer pressure signal (12.47 MPa). When the primary pressure decreases until 4.0 Mpa, the SIT is injected in to the RCS. Figures 3 and 4 present the collapsed water level of the pressurizer, the reactor pressure vessel, and the steam generators. The steam generator level drastically decreases after the MSSV operation. After the steam generator dry-out, heat-up of the primary system leads to an increase of the pressurizer water level. When the feed and bleed operation is actuated, the pressurizer water level rapidly increases by swelling effect, however it decreases due to an inventory loss through the POSRV. The pressurizer water level is recovered again by the compensation of inventory through the SIP injection. Figure 5 shows the behavior of the core inlet and outlet temperature. The core inlet and outlet temperature increase after the level of steam generator decreases, and decrease after POSRV and SIP operation. Because there is a difference in the flow rate between the pumps in the experimental data, the SIP flow rate at each pump is controlled according to the experimental data, as shown in Figure 6. Based on the results of code comparison, the transient behavior of SPACE code shows good agreement with results of OECD-ATLAS A3.1 test for overall major thermal-hydraulic parameters. It means that SPACE code has sufficient capabilities to simulate the transient thermal-hydraulic behavior of typical beyond design basis accident. | | Fig. 6. Mass flow rate of SIP. 5 Conclusion In order to assess the capability of SPACE code of the beyond design basis accident, the total loss of feedwater accident with additional failure was performed and compared with OECD-ATLAS A3.1 test. In order to assess the capability of SPACE code, partial failure was modeled, and compared with results of OECD-ATLAS A3.1 results. The partial failure of the safety injection pumps (SIPs) and the pilot-operated safety relief valves (POSRVs) of pressurizer were selected as additional failures. Partial failure of the s and POSRVs, only 2 SIPs out of 4 SIPs were assumed to be operated and the nozzle at the upstream of the POSRV was replaced with one of 50 % discharging capacity. The transient calculations were conducted to investigate the major thermal-hydraulic phenomena of TLOFW accident. When main and auxiliary feedwater was terminated, the steam generators were dried out due to the MSSVs operation. Because the heat removal capacity of the steam generators decreased, the pressure of the primary system started to increase up to the opening point of POSRV. The feed and bleed operation was conducted as a mitigation measure with reduced capability of the POSRV and SIP. Based on the results and comparison, it is observed that SPACE code shows good agreement with results of OECD-ATLAS A3.1 test for overall major thermal-hydraulic parameters, and it has capability to simulate the beyond design basis accident in terms of the safety analysis. Acknowledgments This work was supported by the Nuclear Research & Development of the Korea Institute of Energy Technology and Planning (KETEP) grant funded by the Korea government Ministry of Trade, Industry and Energy. (No. 20161510101840) References [1] C. Song et al., ATLAS program for advanced thermal-hydraulic safety research, Nuclear Engineering and Design, Vol. 294, pp. 242–261, 2015. [2] Kwon, Y.M. et al., Comparative simulation of feed and bleed operation during the total loss of feedwater event by RELAP5:MOD3 and CEFLASH-4AS:REM computer codes, Nuclear Technology, Vol. 112, pp. 181–193, 1995. [3] Kwon, Y.M., Song, J.H., Feasibility of long term feed and bleed operation for total loss of feedwater event, Journal of Korean Nuclear Society, Vol. 28 (3), pp. 257–264, 1996. [4] Park, R.J. et al., Detailed evaluation of coolant injection into the reactor vessel with RCS depressurization for high pressure sequences, Nuclear Engineering and Design, Vol. 239, pp. 2484–2490, 2009. [5] Pochard, R. et al., Analysis of a feed and bleed procedure sensitivity study performed with the SIPACT simulator on a French 900 MWe NPP, Nuclear Engineering, Des. 215, pp. 1–14, 2002. [6] Reventós, F. et al., Analysis of the feed & bleed procedure for the Ascó NPP first approach study for operation support, Nuclear Engineering, Des. 237, pp. 2006–2013, 2007. [7] S. J. Ha et al., Development of the SPACE Code for Nuclear Power Plants, Nuclear Engineering & Technology, Vol. 43, No. 1, pp. 45, 2011. ENVIRONMENT AND SAFETY 541 Environment and Safety Simulation of Total Loss of Feed Water in ATLAS Test Facility Using SPACE Code ı Minhee Kim and Seyun Kim
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