Severe Accident Management Guidance Research Articles

External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krško

The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krško applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant’s site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant’s post-Fukushima improvements and adequacy with plant’s procedures for severe accident mitigation. Conclusion is that NPP Krško could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance’s (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the lost core geometry.

Model-based and rule-based synthesis of operating procedures for planning severe accident management strategies

When severe accidents happen in nuclear power plants (NPPs), although there are candidate high-level actions (CHLAs) that can be selected from the severe accident management guidance (SAMG) to mitigate consequences, operators may still have cognitive challenges to identify available plant sources and then formulate executable action plans to implement the chosen actions. The difficulties faced by humans may be resolved by a decision support system taking advantages of computers’ knowledge representation and inferential capability.This article presents a systematic approach for synthesizing operating procedures that can achieve operational goals with currently available plant resources. Multilevel flow modeling (MFM), whose representation has been proved to be consistent with the general emergency procedure development, is used as knowledge basis for the action planning. A rule reasoning method is proposed to infer about actions from perspective of functions’ state transition. A model-based and rule-based operating procedure synthesis (OPS) system is developed with a rule engine. The OPS system is applied to elaborate existing high level strategies for dealing with station blackout (SBO) in a BWR plant as multiple executable action plans. The results shows some untypical ways of implementing available systems for achieving current safety goals, which may be considered in the further SAM. Moreover, what critical plant resources should be well prepared in advance is also indicated from the case study. Finally, limitation of MFM and the develop OPS system on planning SAM strategies are discussed.

Best-estimate analysis for a MSGTR accident of CANDU-6 plants using the MAAP-ISAAC code

In CANDU-6 type reactors, a multiple steam generator tube rupture (MSGTR) accident is characterized as a reactor building (RB) bypass scenario. Although the probability of an MSGTR accident in the CANDU-6 type Wolsong plants is very low, a direct release of radioactive nuclides to the environment can cause severe radiation exposure to residents around the plants. For this reason, a best-estimate analysis has been carried out based on the currently available best-practice knowledge in the Wolsong plants. The present study is divided into three steps. First, the existing parameter file of generic CANDU-6 plants for the MAAP-ISAAC 4.03 code was revised to incorporate its design-specific features for the Wolsong plants. A comparative analysis between the existing and modified parameter files showed that the progress of a severe accident was delayed or mitigated in the case of the modified parameter file. Secondly, variables related to the steam generator decontamination factors (DFs) were improved. It was confirmed that applying the steam generator DFs based on the ARTIST test significantly reduced the release fraction of each nuclide element to the environment. Finally, severe accident management guidance (SAMG) developed for the Wolsong plants was used to manage the progression of severe accidents and to limit the release of fission products during a MSGTR accident. A mitigation action of closing the main steam safety valves (MSSVs) after meeting a SAMG entry condition can significantly reduce the amount of released fission products compared with simply supplying feedwater to a broken steam generator (SG). To meet the domestic regulatory requirement for the release of Cs-137 to the environment (100 TBq), it was identified that an operator should supply feedwater to the broken SG within 1.5 h after meeting the SAMG entry conditions. The best-estimate analysis methodology and relevant accident management strategy for the mitigation of severe accidents, which have been developed through the study, can be applied to all similar CANDU plants and provide valuable insights into Level 2 probabilistic safety assessment and development of relevant SAMG.

Best-estimate severe accident and source term analysis for an ISLOCA scenario of a CANDU-6 plant using the MAAP-ISAAC code

An interfacing system LOCA (ISLOCA) is a bypass accident that results in the release of reactor coolant directly to the outside of a reactor building through a break in part of the low-pressure boundary of the reactor coolant system. Though the core damage frequency resulting from an ISLOCA scenario in the CANDU-6 type Wolsong nuclear power plant is very low, a direct release of radioactive nuclides to the environment could lead to severe radiation exposure among residents near the plant, and cannot be overlooked. For this reason, a quantitative evaluation of the amount of radioactive nuclides to be released to the environment was carried out by applying a best-estimate methodology to the MAAP-ISAAC code11MAAP-ISAAC is a CANDU version of MAAP owned by the Electric Power Research Institute, and a relevant accident management strategy for it was investigated in terms of a deterministic analysis. As part of this, an auxiliary building was additionally modeled to simulate deposition effects in the auxiliary building as well as pool scrubbing effects at the submerged break location following an ISLOCA. The decontamination effect by deposition and so on within the piping following an ISLOCA was evaluated. The analysis results showed that the decontamination factor for other nuclide groups excluding inert gas, Sb and Te2 nuclide groups was about 2.5. The analysis with consideration of all these decontamination effects was identified to yield a 78.4% reduction of Cs released to the environment compared with previous conservative analysis, which did not consider the decontamination effect within the ISLOCA piping and in the auxiliary building. In the case of the Wolsong CANDU plant, prior to external cooling water injection to the reactor vault (which is considered to be a Severe Accident Management Guidance (SAMG) accident management strategy), the amount of Cs-137 released to the environment exceeds the domestic regulatory requirement (100 TBq). In contrast, a sensitivity analysis in which an operator closes the test valve installed in a medium-pressure emergency core cooling system piping path and relieves the pressure of a reactor building using a containment filtered venting system (CFVS), met the aforementioned limitation for the amount of Cs-137 when an operator closes the test valve within 1 h following the accident and operates the CFVS properly (i.e., according to its operating procedure). This means that the appropriate accident management strategy for an ISLOCA scenario in the Wolsong CANDU plant can sufficiently prevent severe radiation exposure around the plant. Moreover, it is expected that these best-estimate methodologies developed from this study can be applied to all CANDU-6 plants.

Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism.In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel.The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).

Modelling issues related to molten pool behaviour in case of In-Vessel Retention strategy

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and retain the core melt in the reactor pressure vessel. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactors below 500MWe, like VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000.The demonstration of IVR feasibility for high power reactors requires using less conservative models as the margins are reduced.In this paper modelling issues related to molten pool behaviour in case of IVR are highlighted based on sensitivity studies performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. In particular, the following physical processes are addressed:-Heat transfer coefficients in thin light metal layer and in heavy metal layer.-Kinetics of corium stratification and thermal exchanges between oxide and metal phases during inversion of stratification.-Relocation of steel coming from ablated vessel wall.-Characteristics of oxide crust.Analyses based on a simple configuration of corium in the lower head allow identifying potential impact of each modelling assumption on the heat flux distribution along the vessel wall, especially in transient situations. Based on this study, the most critical areas where the knowledge is too limited or where the current analytical methods should be improved are identified as being (i) the duration of situations with thin metal layer on the top of the oxide, associated to the knowledge of the kinetics of corium stratification and (ii) the possible super-heat of the metal layer at stratification inversion coupled with the risk of oxide crust failure.This work is performed in the frame of the European project IVMR (In-Vessel Melt Retention) coordinated by IRSN. This project has been launched in 2015 and gathers 27 organizations. Its main objective is the evaluation of feasibility of IVR strategy for LWR (PWR, VVER, BWR) of total power 1000 MWe or higher.

Strategy validation of severe accident management guidance for Qinshan-I NPP under typical severe accident sequence

Strategy validation of severe accident management guidance for Qinshan-I NPP under typical severe accident sequence

Development of safety injection flow map associated with target depressurization for effective severe accident management of OPR1000

ABSTRACTIf any severe accident occurs, application of the Severe Accident Management Guidance (SAMG) is initiated by the Technical Support Center (TSC). In order to provide advisory information to the TSC, required safety injection flow rate for maintaining the coolability of the reactor core has been suggested in terms of the depressurization pressure. In this study, mechanistic development of the safety injection flow map was performed by post-processing the core exit temperature (CET) data from MELCOR simulation. In addition, effect of oxidation during the core degradation was incorporated by including simulation data of core water level decrease rate. Using the CET increase rate and core water level decrease rate, safety injection flow maps required for removing the decay and oxidation heat and finally for maintaining the coolability of the reactor core were developed. Three initiating events of small break loss of coolant accidents without safety injection, station black out, and total loss of feed water were considered. Reactor coolant system depressurization pressure targeting the suggested injection flow achievable with one or two high pressure safety injections was included in the map. This study contributes on improving the current SAMG by providing more practical and mechanistic information to manage the severe accidents.

Passive depressurization accident management strategy for boiling water reactors

According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

严重事故时蒸汽发生器传热管蠕变断裂风险评估

严重事故时，一回路将会出现蒸汽自然循环冷凝回流现象；高温蒸汽不断进入热管段，波动管和蒸汽发生器传热管，最终可能导致蒸汽发生器传热管发生蠕变断裂失效。本文首先使用严重事故分析计算严重事故时传热管蠕变断裂参数和传热管断裂的概率。然后使用简化二级概率安全分析模型评估传热管蠕变断裂导致的大量早期释放频率。最后评价一回路卸压和恢复二次侧给水等严重事故缓解措施对防止蠕变断裂的有效性。可以发现在中压和高压的情况下，一回路卸压和恢复二次侧给水能够有效地防止传热管蠕变断裂，降低大量早期释放频率。 Under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg, surge line and steam generator tubes, SG tubes integrity could be threatened by creep rupture, particularly if cracks are present in the tube walls. In this study, the first step we perform thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step we apply the creep rupture models to test the potential for the degraded SG to rup-ture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in simplified Level-2 PSA model, and the overall LERF (Large and Early Release Frequency) risk due to the Induced SGTR was calculated. In the final step, imple-mentation of severe accident management guidance, such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the severe accident process under the condition of high and medium pressure, and reduce LERF effectively.

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.

Human performance evaluation: The procedures of ultimate response guideline for nuclear power plants

In the nuclear accident which occurred in Japan on March 11, 2011, several units of Fukushima conventional BWR experienced a total loss of power and water supply triggered by a heavy earthquake and subsequent Tsunami which were outside design models. In the past, when an accident occurred, operators in nuclear power plants (NPP) followed emergency operating procedures (EOPs) or severe accident management guidance (SAMG). However, EOP and SAMG are symptom-based procedures to cope with severe transients and accidents, depending on real-time operational parameters. Ultimate response guidelines (URG), a plant specific interim remedy action plan, was developed to manage accidents caused by compound disasters which exceed design models. The URG guides the plant operators’ conduct of reactor depressurization, core cooling water injection, and containment venting. This study adopts NUREG/CR-6883 (Standardized Plant Analysis Risk Human Reliability Analysis, SPAR-H) to evaluate human error probabilities (HEPs) of action and diagnosis in the current URG procedures. We found the human reliability of URG procedures analyzed by SPAR-H is about 85% (depending on different decision makers). Upgrading the training level of staff or enhancing plant managers ability to decide whether to execute URG will enhance the human reliability of URG procedures.

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.

Investigation of nuclear power plant behaviour at low power and cold conditions during an overpressurization in primary circuit

This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behaviour at low power and cold conditions during an overpressurization in the primary circuit. The reference nuclear power plant for this analysis is Unit 6 at Kozloduy NPP (KNPP) site. The systems and equipment of the KNPP, Unit 6 operate according to the design requirements for the corresponding level of reactor power. In this paper is also presented an analysis of the additionally installed equipment for situations such as cold overpressurization. This equipment was mounted during the Modernization program in KNPP. In the paper is analysed the response of the new equipment together with other NPP available safety systems during the postulated transient in shutdown state. For the purpose of this analysis a RELAP5/MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of Kozloduy NPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The low power and cold conditions and the modifications after the modernization program, have been taken into account.The main purpose of this analysis is to estimate the parameters of the monitored plant which are used to identify symptoms that are used by operators to identify the plant’s state and the Critical Safety Function (CSF). The analysis is also used to define the timing needed to reach the following stages during the progression of processes in the reactor system:•Reaching the saturated temperature at the outlet of the assembly.•Beginning of reactor core uncover.•Heating up of fuel.•Defining the transition time between Emergency Operating Procedures (EOPs) and Severe Accident Management Guidance (SAMG) at temperature of 923.15K.•Restoring of water level in the core.•Defining the CSF “Integrity” status and the time of its loss.The results of the thermal–hydraulic analysis have been used to assist KNPP specialists in analytical validation of EOPs at low power and cold conditions.The performed analysis is based on a previously used bounding approach in analytical validation of Symptom Based EOPs (SB EOPs). The presented thermal–hydraulic calculations of the accident scenarios involve the loss of Critical Safety Function (CSF) “Integrity” for WWER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP). During the analysis is also discussed the behaviour of CSF “Core cooling”. The most specific in this analysis compared to the analyses of NPP accidents at full power is the lack of some important safety systems due to NPP regulations. Based on this approach a list of scenarios has been performed, involving a different number of safety systems with or without operator actions. The other specific characteristic of the accidents in NPP at low power and cold conditions is that even though all process are progressing significantly slower compared to full power, the operator actions should always bring the reactor system under safety conditions.

The Analysis of Severe Accident Induced Steam Generator Tube Rupture and LERF Risk

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg and SG tubes. The first step performs thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step applies the creep rupture models to test the potential for the degraded SG to rupture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in LERF (Large and Early Release Frequency) model, and the overall LERF risk due to the Induced SGTR was calculated. In the final step, implementation of severe accident management guidance (SAMG), such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the result of induced SGTR and LERF effectively.

Effect of SAMG entry condition on operator action time for severe accident mitigation

The effect of SAMG (Severe Accident Management Guidance) entry condition on operator action time for prevention of reactor vessel failure in the OPR1000 (Optimized Power Reactor 1000) was evaluated with a comparison to other conditions using the SCDAP/RELAP5/MOD3 computer code in detail. Dominant severe accident sequences, such as a station blackout, a total loss of feedwater, a small and medium break loss of coolant accidents without a safety injection were evaluated in this study. The results showed that a core exit temperature of 480 °C for the SAMG entry condition was too early, but a core exit temperature of 1000 °C was too late for operator action time to prevent reactor vessel failure. The SAMG entry condition of a core exit temperature of 650 °C for the OPR1000 was suitable for the operator action timing point in order to mitigate a severe accident.

SAMEX: A severe accident management support expert

A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R&D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.

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.

The Impact of Severe Accident Management Guidance on the Frequency of Containment Failure Modes of a Boiling Water Reactor

Probabilistic safety assessment (PSA) employs a systematic approach to estimate the risk associated with the operation of nuclear power plants (NPPs). Severe accident management guidance (SAMG), which delineates the mitigation actions of core meltdown accidents of NPPs, is developed to support operators and staffs in the technical support centers during the emergency responses of core melt accidents. Proper execution of SAMG could lower the failure probability of containment and reduce the amount of radionuclides released to the environment during the accident. It can be expected that the implementation of SAMG will reduce the risk of NPPs. However, SAMG is not available when most of the conventional level-2 PSA analyses are performed. In the present study, the mitigation actions of SAMG are incorporated into the level-2 PSA model of the ChinShan Nuclear Power Station of the Taiwan Power Company. The NPP analyzed employs a General Electric-designed boiling water reactor-4 with Mark I containment.The effectiveness of the mitigation actions specified in SAMG to terminate the progression of the accident is verified and validated using the MAAP4 code. The containment system event trees and containment phenomenological event trees of the level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The Human Cognitive Reliability (HCR) and Technique for Human Error Rate Prediction (THERP) models are used to quantify the human error probability (HEP) of all the actions in the level-2 PSA model. The MAAP4 code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the total frequency of accident progression beyond vessel failure is reduced by 41% and the change in the probability of containment staying intact is not very significant because of the implementation of SAMG. After SAMG implementation, the frequency of containment early failure is reduced by 69.9%. The frequency of suppression pool venting is increased by 77.9%. The changes in the frequency of other containment failure modes are relatively insignificant. The most important human action is specified in Guideline RC/F of Severe Accident Guideline-1, i.e., In-Vessel Injection to Arrest Core Damage.

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.