Safety Depressurization System Research Articles

Time uncertainty analysis method for level 2 human reliability analysis of severe accident management strategies

This paper proposes an extended time uncertainty analysis approach in Level 2 human reliability analysis (HRA) considering severe accident management (SAM) strategies. The method is a time-based model that classifies two time distribution functions—time required and time available—to calculate human failure probabilities from delayed action when implementing SAM strategies. The time required function can be obtained by the combination of four time factors: 1) time for diagnosis and decision by the technical support center (TSC) for a given strategy, 2) time for strategy implementation mainly by the local emergency response organization (ERO), 3) time to verify the effectiveness of the strategy and 4) time for portable equipment transport and installation. This function can vary depending on the given scenario and includes a summation of lognormal distributions and a choice regarding shifting the distribution. The time available function can be obtained via thermal-hydraulic code simulation (MAAP 5.03). The proposed approach was applied to assess SAM strategies that use portable equipment and safety depressurization system valves in a total loss of component cooling water event that could cause reactor vessel failure. The results from the proposed method are more realistic (i.e., not conservative) than other existing methods in evaluating SAM strategies involving the use of portable equipment.

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

ABSTRACTDuring a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.

영광3,4호기 안전감압계통 추가설비 설계최적화를 위한 시스템엔지니어링 적용연구

The purpose of this paper is to present the results of reengineering of the YGN 3&4 (Yonggwang Nuclear Power Plant, Units 3&4) SDS (Safety Depressurization System) retrofit design and to make recommendations for the improvement in design and design procedure implementing the Systems Engineering (SE) process. YGN 3&4 is a basic model for OPR1000 (the Korean standard 1000 MWe plant). The basic model, herein, represents the reference plant for the OPR1000 development. In the middle of the YGN 3&4 construction, the Korean Nuclear Regulatory Body requested a retrofit of this plant with a means to rapidly depressurize the plant in conformance with a severe accident mitigation requirement. For the reengineering of the SDS in YGN 3&4, V-model and functional and physical architectures have been developed. A SE decision making method has been used for the selection of SDS valves. Finally, recommendations have been made to improve OPR1000 design for the improved operation and enhanced safety.

Dynamic sequence analysis for feed-and-bleed operation in an OPR1000

A feed-and-bleed (F&B) operation in an OPR1000 provides residual heat removal when the secondary cooling system is not available. Operators may hesitate to initiate F&B operation if a clear cue is not provided because its initiation will result in the release of radioactive coolant into the containment structure. An analysis of the necessity and effects of F&B operation was performed systematically. The plant conditions requiring F&B operation were categorized as type of transients. The factors which affect the necessity and effects of F&B operation are water inventory in the primary and secondary cooling systems, break size and timing of the break-up in the case of loss-of-coolant accident, the availability of the safety injection system and safety depressurization system, and timing of the F&B operation initiation. The boundary of each state in an OPR1000 was identified and the effects of factors were investigated to support the operator’s decision to initiate F&B operation under various reactor conditions. A thermohydraulic analysis was performed using the MARS code to determine this success boundary. This study is expected to provide a useful guideline for the initiation of F&B operation for operators. In particular, in the case of a combined accident including a secondary cooling system failure, the provision of clear information will play a critical role in the operators’ decision to initiate F&B operation.

Validation of RCS depressurization strategy and core coolability map for independent scenarios of SBLOCA, SBO, and TLOFW

Severe accident analysis for Small Break (SB), Middle Break (MB), and Large Break (LB) Loss-Of-Coolant Accident (LOCA), Station Black Out (SBO), Total Loss-Of-Feed-Water (TLOFW) was performed and effectiveness of Reactor Coolant System (RCS) depressurization strategies of OPR1000 was analyzed using MELCOR 1.8.6 code. Required injection flow rate has been derived using Core Exit Temperature (CET) information obtained from MELCOR calculation and a simple model and corresponding coolability map have been suggested to assist effective operator action. The depressurization strategies using secondary Atmospheric Dump Valve (ADV) for SBLOCA, pressurizer Safety Depressurization System (SDS) for SBO and TLOFW were introduced in 5 min since the initiation of Severe Accident Management Guidance (SAMG). Respective mitigation strategy employed leads to significant delay of the reactor pressure vessel failure and RCS pressure at Reactor Pressure Vessel (RPV) failure was lower than the SAMG target pressure of 2.86 MPa. Thus, possibility of High Pressure Melt Ejection (HPME) and impair of containment building is expected to avoid effectively. Using CET information obtained from MELCOR calculation, a simple model and a coolability map for the required injection flow rate were developed for recovery of core coolability. It is suggested that the coolability map based on MELCOR calculation results may provide decisive and intuitive information to operators for more effective safety management.

APR에서 사용하는 SDS 특성 분석에 관한 연구

안전감압시스템(SDS)의 주요 기능은 시스템의 압력이 운전 제한치를 초과하는 비정상적인 상황이 일어날 경우에 빠른 시간 내에 안정적으로 발전소의 압력을 감압시켜 발전소를 정상화하는데 그 목적이 있기 때문에 이러한 기능을 만족시키기 위한 안전감압시스템의 설계 인자가 무엇인가를 찾고자 하였다. 시스템의 압력을 조절하는 가압기 상단부분의 안전 감압밸브를 열어 증기를 배출함으로 시스템의 압력을 낮추는데 증기 배출을 위한 배관 설계에 있어서 배관내의 과도 유동 상태를 실제 발전소 여건에 맞도록 전산 프로그램을 개발하여 안전감압시스템의 배관에서의 증기의 과도 유동 해석을 수행하였고 그 결과를 토대로 안전감압시스템의 배관 및 배관 지지대 설계에 활용하고자 하였다. The purpose of this study analysis is evaluation of Safety Depressurization System function on PWR plant. The Safety Depressurization System is a system which releases steam from the pressurizer of Reactor Coolant System(RCS) and depressurizes RCS. There is a similar system in Boiling Water Reactor(BWR) type plant but the SDS with In-Containment Refueling Water Storage System(IRWST) is a new system to Pressurized Water Reactor(PWR). The current worldwide trend in a new PWR design is to implement a SDS with IRWST because of the advantages in installing IRWST. There have been also trials to install a SDS with IRWST to newly ordered domestic plant was that the domestic technical level for analysis and design of the system was not sufficient to do work.

Evaluation of the RCS depressurization strategy for the high pressure sequences by using SCDAP/RELAP5

As part of the evaluation for a severe accident management strategy, a reactor coolant system (RCS) depressurization in optimized power reactor (OPR)1000 has been evaluated by using the SCDAP/RELAP5 computer code. An indirect RCS depressurization by a secondary depressurization by using a feed and bleed operation has been estimated for a small break loss of coolant accident (LOCA) without a safety injection (SI). Also, a direct RCS depressurization by using the safety depressurization system (SDS) has been estimated for the total loss of feed water (LOFW). The SCDAP/RELAP5 results have shown that the secondary feed and bleed operation can depressurize the RCS, but it cannot depressurize the RCS sufficiently enough. For this reason, a greater direct RCS depressurization by using the SDS is necessary for the 1.35 in. break LOCA without SI. A proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time from 7.5 to 10.7 h. An opening of two SDS valves can depressurize the RCS sufficiently enough and the proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time of approximately 5 h for the total LOFW. An opening of one SDS valve cannot depressurize the RCS sufficiently enough.

Liquid entrainment and off-take through the break at the top of a vessel

In order to determine the bleed capacity of the safety depressurization system (SDS) of Advanced Power Reactor 1400 (APR1400) in the case of total loss of feed water (TLOFW), we performed an experimental study of liquid entrainment and liquid off-take from the swelled two-phase mixture surface in a vessel. A total of 220 experimental data on the entrainment and off-take are obtained using a test vessel with a height of 2.0 m and an inner diameter of 0.3 m and a top break with a diameter of 0.05 m. Two-phase mixture levels are measured by an ultrasonic sensor within ±1.77% with respect to the visual level data. Droplet entrainment data are obtained with and without the top break and are compared with the existing pool entrainment data. The present droplet entrainment data have higher values than the existing pool entrainment data due to (a) the pulling toward the break of the liquid deentrained on the top wall of the vessel and (b) gas acceleration in the vicinity of the break. In the present experiment, droplet entrainment, E fg, strongly depends upon j g * / h * and is proportional to the seventh power of j g * / h * in the same way as the off-take data. The empirical correlation for the onset of off-take is developed in terms of the Froude number ( Fr g) at the break and the non-dimensional inception height ( h b/ d). This correlation shows agreement with the present experimental data within ±15%. The present off-take quality data show agreement with Schrock's off-take quality correlation with the rms error of 15.8%.

Design options for safety depressurization system

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature, a safety depressurization system (SDS) to rapidly depressurize the primary system in case of events beyond the design basis. Two design approaches are considered for the SDS design. The use of bleed valves similar to the ABB-CE System 80+ is design option 1, while in design option 2, the French Sebim valve is considered to provide the combined function of overpressure protection and rapid depressurization. In this paper, thermal hydraulic analysis using a best-estimate version of CEFLASH-4AS/REM is performed for a total loss of feedwater (TLOFW) event to investigate the feasibility of those two design options. For each design option, various feed and bleed (F and B) procedures are investigated for a TLOFW event. For design option 1, the required bleed capacity is determined from the CEFLASH-4AS/REM simulation according to the EPRI Advanced Light Water Reactor (ALWR) requirements. The analysis results demonstrate that the TLOFW event can be mitigated in a proper manner with a sufficient margin using design option 1. For design option 2, the operator action times for initiating the F and B are investigated by varying the number of Sebim valves and high pressure safety injection (HPSI) pumps. If the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible.