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

During 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.     

Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima.     

Effectiveness and adverse effects of reactor coolant system depressurization strategy with various severe accident management guidance entry conditions for OPR1000

Severe accident analysis for Korean OPR1000 with MELCOR 1.8.6 was performed by adapting a mitigation strategy under different entry conditions of Severe Accident Management Guidance (SAMG). The analysis was focused on the effectiveness of the mitigation strategy and its adverse effects. Four core exit temperatures (CETs) were selected as SAMG entry conditions, and Small Break Loss of Coolant Accident (SBLOCA), Station Blackout (SBO), and Total Loss of Feed Water (TLOFW) were selected as postulated scenarios that may propagate into severe accidents. In order to delay reactor pressure vessel (RPV) failure, entering the SAMG when the CET reached 923 K, 923 K, and 753 K resulted in the best results for SBLOCA, SBO, and TLOFW scenarios, respectively. This implies that using event-based diagnosis for severe accidents may be more beneficial than using symptom-based diagnosis. There is no significant difference among selected SAMG entry conditions in light of the operator's available action time before the RPV failure. Potential vulnerability of the RPV due to hydrogen generation was analyzed to investigate the foreseeable adverse effects that act against the accident mitigation strategies. For the SBLOCA cases, mitigation cases generated more hydrogen than the base case. However, the amount of hydrogen generated was similar between the base and mitigation cases for SBO and TLOFW. Hydrogen concentrations of containment were less than 5% before RPV failure for most cases.     

Study on severe accident mitigation measures for the development of PWR SAMG

In the development of the Severe Accident Management Guidelines （SAMG）, it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident （LOCA）, Steam Generator Tube Rupture （SGTR）, Station Blackout （SBO）, and Anticipated Transients without Scram （ATWS） in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management （SAM） such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.     

Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences.