Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

This study focuses on the in-vessel phase of severe accident management (SAM) strategy for a hypothetical 1000 MWe pressurized water reactor (PWR). To examine the effectiveness of SAM strategy, it is necessary to identify and assess epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is reactor coolant system (RCS) depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, sensitivity analyses are carried out to assess the impact of the cutoff porosity related to the flow area of core node (EPSCUT), the critical temperature (TCLMAX) and the minimum fraction of oxidized Zr (FZORUP) for cladding rupture, and the flag to divert gas flows in the core to the bypass channel (FGBYPA) on the core melting and relocation process. In this study, the effect of injection time on the integrity of RPV has been examined based on the quantification of the scenario and phenomenological uncertainties.     

Development of severe accident management guidance (SAMG) for the Canadian CANDU 6 nuclear power plants

This paper discusses the severe accident management guidance (SAMG) development process undertaken for the Canadian CANDU 6 nuclear power plants (NPPs); the customization process of the generic CANDU SAMG for the Point Lepreau NPP is presented. Examples of severe accident management (SAM) guidelines related to containment pressure control are included in this paper. This paper also provides an overview summary of the severe accident analysis program at Atomic Energy of Canada Limited (AECL) that complements the SAM guidelines development process for the CANDU 6 NPPs in Canada. These analyses provided insights into the accident progression and basis to develop the SAM guidelines.     

Development of a risk-informed accident diagnosis and prognosis system to support severe accident management

RISARD, risk-informed severe accident risk diagnosis system, is a computerized tool developed to improve a severe accident management (SAM) for a nuclear power plant and to effectively support the MCR and the TSC in executing the relevant SAM activities. In order to provide a diagnostic capability to a state of the plant and a prognostic capability for an anticipated accident progression, the system examines (a) a symptom-based diagnosis of a plant damage state (PDS) sequence in a risk-informing way and (b) a PDS sequence-based prognosis of key plant parameter behavior, through a prepared database (DB) containing plant-specific severe accident risk (SAR)-related information. For a given accident, the replicated use of these two processes makes it possible to obtain information about the functional states of the plant and containment safety systems expected at the time of a severe accident as well as future trend of the key plant parameters that are essentially required for taking the relevant SAM action, eventually leading to an answer about the best strategy for SAM. The foregoing concept for an accident diagnosis and prognosis can give the SAM practitioners more time to take action for mitigating the consequences of the potential accident scenarios since they are made in a simple, fast, and efficient way through a prepared SAR database and it is useful especially when the plant information available for SAM is incomplete and limited. The main purpose of this paper is to (a) introduce the concept of the RISARD system proposed to support SAM and its implementation through a prepared OPR1000 plant- and MAAP code-specific SAR database and (b) assess prediction capabilities of major events expected during the evolution of a severe accident through the system.     

A simple novel analysis procedure for IVR calculation in core-molten severe accident

In-Vessel Retention (IVR) of core melt is a key severe accident management strategy adopted by operating nuclear power plants and advanced light water reactors (ALWRs), AP600, AP1000, etc. External Reactor Vessel Cooling (ERVC), which involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris relocated to the vessel low head, is a novel severe accident management for IVR analysis. In present study, IVR analysis code in severe accident (IVRASA) has been proposed to evaluate the safety margin of IVR in AP600 with anticipative depressurization and reactor cavity flooding in severe accident. For, IVRASA, a simple novel analysis procedure has been developed for modeling the steady-state endpoint of core melt configurations. Furthermore, IVRASA was developed in a more general fashion so that it is applicable to compute various molten configurations such as UCSB Final Bounding State (FIBS). The results by IVRASA were consistent with those of the UCSB and INEEL. Benchmark calculations of UCSB-assumed FIBS indicate the applicability and accuracy of IVRASA and it could be applied to predict the thermal response of various molten configurations.     

Severe accident management strategy verification for VVER-1000 (V320) reactor

Severe accident analysis of a reactor is an important aspect in evaluation of source term. This in turn helps in emergency planning and Severe Accident Management (SAM). The use of the Severe Accident Management Guideline (SAMG) is required for accident situation which is not handled adequately through the use of Emergency Operating Procedures (EOP), thus leading to a partial or a total core melt. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). Initiation of SAMG for VVER-1000 is considered at two core exit temperatures viz. 650 °C as a desirable entry temperature and 980 °C as a backup action. Analyses have been carried out for VVER-1000 (V320) for verification of some of the strategies namely water injection in primary and secondary circuit. These strategies are analysed for a high and low pressure primary circuit transients. Station Black Out (SBO) is one such high pressure transient for which core heat can be removed by natural circulation of the primary circuit inventory by maintaining the secondary side inventory. This strategy has been verified where the feed water injection to secondary side of SG is considered from external power sources (e.g. mobile DG sets) as suggested in SAM guidelines. The second transient, a low pressure event is analysed for verification of the SG flooding and core flooding strategies. The analysis shows that SG flooding is not adequate to arrest the degradation of the core. In case of core flooding strategy, the analyses show that core flooding is not adequate to arrest the degradation of the core for the large break LOCA where as for small break LOCA the injections through available safety systems are adequate. The assessments are carried out with integral severe accident computer code ASTEC V1.3.     

CHF enhancement by vessel coating for external reactor vessel cooling

In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding the reactor cavity during a severe accident. As part of a joint Korean–United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the subscale boundary layer boiling (SBLB) facility at the Pennsylvania State University using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady-state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady-state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.     

严重事故管理导则入口条件研究

严重事故管理导则的入口是从电厂应急运行规程(EOP)向严重事故管理导则(SAMG)转换的条件,也是严重事故缓解行动的重要依据。本文选取失去四级电源导致的典型高压熔堆序列以及大破口失水事故(LLOCA)导致的典型低压熔堆序列,根据严重事故堆芯剧烈氧化机理,得出了燃料温度、氢气产生速率及产氢量、入口集管过冷度以及慢化剂液位的关系。结果表明入口集管过冷度小于0且持续十几分钟,同时慢化剂系统的状态指示慢化剂液位低于6 900mm,可以作为严重事故管理的入口条件。最后,阐述了目前电厂EOP向SAMG转换的机制,并提出了改进的意见。     

Analysis of safety margin of in-vessel retention for AP1000

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by most operating nuclear power plants and advanced light water reactors (ALWRs), AP600, AP1000, etc. External reactor vessel cooling (ERVC) is a novel severe accident management for IVR analysis. In present study, IVR analysis code in severe accident (IVRASA) has been developed to evaluate the safety margin of IVR in AP1000 with anticipative depressurization and reactor cavity flooding in severe accident. For, IVRASA, the point estimate procedure has been developed for modeling the steady-state endpoint of two core melt configurations: Configuration I and Configuration II. The results of benchmark calculations of AP600 by IVRASA were consistent with those of the UCSB and INEEL. Then, IVRASA is used to calculate the heat transfer process caused by two core melt configurations of AP1000. The results of calculations of Configuration I indicate that the heat flux remains below the critical heat flux (CHF), however, the sensitivity calculations show that the heat flux in the metallic layer could exceed the CHF because of the focusing effect due to the thin metallic layer. On the other hand, the results of calculations of Configuration II suggest that the thermal failure of the lower head at the bottom location is highly unlikely, but the heat flux in light metallic layer could be higher than that of base case due to the portion of metal partitioning into the lower head. This work also investigated the effect of the uncertainties of the CHF correlations on the analysis of IVR.     

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment.     

严重事故下大功率先进压水堆IVR-ERVC有效性分析

通过压力容器外部冷却(ERVC)以实现堆内熔融物滞留(IVR)作为反应堆严重事故缓解管理的一项重要举措一直以来广泛受到关注和研究。本文使用严重事故分析程序MELCOR,从瞬态角度对大型先进压水堆进行了IVR-ERVC相关研究。过程中重点关注了堆芯熔毁和重新定位,熔池形成、生长及其传热过程,并且对压力容器外部流动传热进行了分析。MELCOR计算所得下封头热流密度分布的瞬态结果与临界热流密度(CHF)比较和分析表明,1700 MWe大功率压水堆发生严重事故后在IVRERVC条件下能够保证压力容器的完整性,即,IVR-ERVC能够有效带出下封头熔融物的衰变热量,缓解严重事故后果。     

Severe accident mitigation strategy for the generation II PWRs in France – Some outcomes of the on-going periodic safety review of the French 1300 MWe PWR series

The 3rd Periodic Safety Review of the French 1300 MWe PWRs series includes some modifications to increase their robustness in case of a severe accident. Their review is based on both deterministic and probabilistic approaches, keeping in mind that severe accidents frequencies and radiological consequences should be as low as reasonably practicable, severe accidents management strategies should be as safe as possible and the robustness of equipment used for severe accident management should be ensured.Consequently, the IRSN level 2 probabilistic safety assessment (L2 PSA) studies for the 1300 MWe reactors have been used to re-assess the results of the utility's L2 PSA and rank them to identify the containment failure modes contributing the most to the global risk. This ranking helped the review of plant modifications.Regarding strategies for accident management, the EDF management of water in the reactor cavity during a severe accident for the 1300 MWe PWRs is presented as well as the IRSN position on this strategy: this is an example where the optimal severe accident management strategy choice is not so easy to define.Regarding the robustness of equipment used for severe accident management, the interest of a diversification or redundancy of the French emergency filtered containment venting opening is one example among many others.This paper presents the analysis conducted by IRSN during the 3rd periodic safety review of the French 1300 MWe PWRs. Future NPP upgrades to limit radioactive releases in case of containment filtered venting, to prevent containment venting and basemat melt-through are analysed in another framework (post-Fukushima and long-term operation projects).     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.     

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.     

The development and demonstration of integrated models for the evaluation of severe accident management strategies—SAMEM

This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the generic modelling framework was refined to enable a number of applications. Some recommendations have also been made regarding the applicability of these approaches to existing operating reactors and future reactors. The need for further research and development in the area of human reliability quantification was identified.     

Accomplishments and challenges of the severe accident research

This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research.     

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.     

Consequence analysis for nuclear reactors,Yongbyon

Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea.     

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.     

Detailed evaluation of coolant injection into the reactor vessel with RCS depressurization for high pressure sequences

A coolant injection into the reactor vessel with depressurization of the reactor coolant system (RCS) has been evaluated as part of the evaluation for a strategy of the severe accident management guidance (SAMG). Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feedwater (LOFW) accident in Optimized Power Reactor (OPR)1000 have been analyzed by using the SCDAP/RELAP5 computer code. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 s with indirect RCS depressurization by using one condenser dump valve (CDV) at 6  min after implementation of the SAMG prevents reactor vessel failure for the small break LOCA without SI. In this case, only one train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent reactor vessel failure. Only one train operation of the HPSI at 20,208 s with direct RCS depressurization by using two SDS valves at 40 min after an initial opening of the safety relief valve (SRV) prevents reactor vessel failure for the total LOFW.     

Development of severe accident management advisory and training simulator (SAMAT)

The most operator support systems including the training simulator have been developed to assist the operator and they cover from normal operation to emergency operation. For the severe accident, the overall architecture for severe accident management is being developed in some developed countries according to the development of severe accident management guidelines which are the skeleton of severe accident management architecture. In Korea, the severe accident management guideline for KSNP was recently developed and it is expected to be a central axis of logical flow for severe accident management. There are a lot of uncertainties in the severe accident phenomena and scenarios and one of the major issues for developing a operator support system for a severe accident is the reduction of these uncertainties. In this paper, the severe accident management advisory system with training simulator, SAMAT, is developed as all available information for a severe accident are re-organized and provided to the management staff in order to reduce the uncertainties. The developed system includes the graphical display for plant and equipment status, the previous research results by knowledge-base technique, and the expected plant behavior using the severe accident training simulator. The plant model used in this paper is oriented to severe accident phenomena and thus can simulate the plant behavior for a severe accident. Therefore, the developed system may make a central role of the information source for decision-making for a severe accident management, and will be used as the training simulator for severe accident management.