FUELPOOL: A Computer Program to Model CANDU Spent Fuel Pool Severe Accident Progression and Consequences

2021 ◽  
Author(s):  
Yong Mann Song ◽  
Jong Yeob Jung ◽  
Sunil Nijhawan
Keyword(s):  
Computer Program ◽  
Severe Accident ◽  
Spent Fuel ◽  
Spent Fuel Pool

FUELPOOL: A Computer Program to Model CANDU Spent Fuel Pool Severe Accident Progression and Consequences

2020 ◽  
Author(s):  
Yong Mann Song ◽  
Jong Yeob Jung ◽  
Sunil Nijhawan
Keyword(s):  
Water Layer ◽  
Daily Basis ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Fluid Loss ◽  
Spent Fuel ◽  
On Line ◽  
Additional Water ◽  
Spent Fuel Pool ◽  
Tennis Court

Abstract CANDU PHWR spent fuel pools (SFPs), smaller than a tennis court, contain up to 38,000 or more (49,000 in Wolsong)fuel bundles in geometries not replicated in any other power reactor. Therefore, the phenomenological issues, accident progression pathways and effectiveness of mitigative actions are somewhat different. This requires a dedicated approach in progression and consequence assessments of potential accidents and development of mitigation measures. The SFPs house densely packed fuel bundles stacked in about a hundred vertical stainless steel tray towers, each containing 24 spent fuel bundles in each of the 16 or more (19 in Wolsong) horizontal fish basket style steel trays. Some of theupto 10 year worth of the on-line refuelled bundles in the SFP are at relatively high decay powers as fuel trays are prepped for the towers in near daily basis. In addition, there is a provision (see Figure 1) that a full core of bundles 20 days after being at full power can be transferred to the spent fuel bay at any time. About 4.5m of additional water layer on top of the tray towers provide radiation protection and a healthy margin to small rate of fluid loss.

Probabilistic Assessment of Countermeasures for Loss of Ultimate Heat Sink to Spent Fuel Pool

2017 ◽  
Author(s):  
Bumpei Fujioka ◽  
Naoki Hirokawa ◽  
Daisuke Taniguchi
Keyword(s):  
Heat Sink ◽  
Nuclear Power ◽  
Water System ◽  
Cooling Water ◽  
Severe Accident ◽  
Probabilistic Assessment ◽  
Spent Fuel ◽  
The Past ◽  
Plant Safety ◽  
Spent Fuel Pool

In the Fukushima Dai-ichi nuclear power station, Loss of Ultimate Heat Sink (LUHS) was caused by the great east japan earthquake and the subsequent tsunami [1]. It resulted in severe accident in three units. In that time, fuel damage in Spent Fuel Pool (SFP) were prevented by the various countermeasures such as makeup by pump truck and recovery of injection systems /cooling water system. In the past, Probabilistic Safety Assessment (PSA) has been developed with a focus on the reactor. After the accident, it has been acknowledged that SFP PSA is important to enhance the plant safety. In this study, probabilistic assessment is performed to suggest countermeasures for LUHS to SFP.

Study on Hydrogen Risk of Spent Fuel Compartment Induced by Containment Venting

2017 ◽  
Author(s):  
Yabing Li ◽  
Xuewu Cao
Keyword(s):  
Control System ◽  
Heat Removal ◽  
Compartment Model ◽  
Severe Accident ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Spent Fuel Pool ◽  
Heat Removal System ◽  
And Control ◽  
Removal System

Hydrogen risk in the spent fuel compartment becomes a matter of concern after the Fukushima accident. However, researches are mainly focused on the hydrogen generated by spent fuels due to lack of cooling. As a severe accident management strategy, one of the containment venting paths is to vent the containment through the normal residual heat removal system (RNS) to the spent fuel compartment, which will cause hydrogen build up in it. Therefore, the hydrogen risk induced by containment venting for the spent fuel compartment is studied for advanced passive PWR in this paper. The spent fuel pool compartment model is built and analyzed with integral accident analysis code couple with the containment analysis. Hydrogen risk in the spent fuel pool compartment is evaluated combining with containment venting. Since the containment venting is mainly implemented in two different strategies, containment depressurization and control hydrogen flammability, these two strategies are analyzed in this paper to evaluated the hydrogen risk in the spent fuel compartment. Result shows that there will not be significate hydrogen built up with the hydrogen control system available in the containment. However, if the hydrogen control system is not available, venting into the spent fuel pool compartment will cause a certain level of hydrogen risk there. Besides, suggestions are made for containment venting strategy considering hydrogen risk in spent fuel pool compartment.

Analyses of Severe Accident Sequences in the Spent Fuel Pool of the VVER-1000 Type of Reactor With MELCOR 1.8.6

2014 ◽  
Author(s):  
Miroslav Kotouč
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Geometry ◽  
Steel Plates ◽  
Severe Accident ◽  
Spent Fuel ◽  
Concrete Walls ◽  
Nuclear Disaster ◽  
Spent Fuel Pool ◽  
Fukushima Nuclear Disaster

Since the Fukushima nuclear disaster in 2011, much attention has been paid to investigation of severe accidents (SA) progression in spent fuel pools (SFP) of various types of nuclear power plants (NPP). In Czech Republic, 4 VVER-440 and 2 VVER-1000 types of reactors (at the Dukovany and Temelin NPPs, respectively) are currently under operation. In order to enhance their safety, especially with respect to station black-out (SBO) events, numerical analyses have been carried out evaluating the risks associated with accidents occurring also in the SFP. The present paper deals with analyses of 2 postulated scenarios (loss of cooling and loss of coolant) and is mainly focused on the input deck preparation for the integral, lumped parameter (LP) code for SA analyses — MELCOR 1.8.6. Emphasis is put on description of correct implementation of the complex geometry of the SFP, consisting of 3 distinct pools separated by concrete walls (lined with steel plates) in which fuel assemblies (FA) are stored in an absorber grid (rack). In the description of the prepared numerical model, light is shed on the encountered modeling issues, and corresponding hints and solutions are proposed in order to provide guidance for preparing adequate models for various types of SFP designs. Finally, some of the most characteristic results are presented for both postulated scenarios.

Analysis of a Loss of Heat Removal Accident in a PWR Spent Fuel Pool

2014 ◽  
Author(s):  
Xiaoli Wu ◽  
Yapei Zhang ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Flow Rate ◽  
Evaporation Rate ◽  
Heat Removal ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Reference Case ◽  
Makeup Water ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents a study on the consequences of loss of heat removal accident in the spent fuel pool of a typical pressurized water reactor using the Modular Accident Analysis Program (MAAP5) code. Analysis of uncompensated loss of water due to the loss of heat removal with initial pool water level of 12.2 m (designated as a reference case) has been performed. The analyses cover a broad spectrum of severe accident in the spent fuel pool. Those consequences such as overheating of uncovered fuel assemblies, oxidation of zirconium and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission product are also analyzed in this paper. Furthermore, as important mitigation measures, the effects of makeup water in SFP on the accident progressions have also been investigated based on the events of spent fuels uncovery. The results showed that spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate higher than evaporation rate defined in the reference case. Although spent fuel assemblies partly exposed due to a mass flow rate of makeup water smaller than the average evaporation rate, continuous steam cooling and radiation heat transfer might maintain the spent fuels coolability as the actual evaporation was balanced by the makeup in a period of time of the order of several days. However, larger makeup rate should be guaranteed to ensure long-term safety of SFP.

The Importance of Reliable Indications for Accident Management

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
Bill T. Williamson
Keyword(s):  
Severe Accident ◽  
Spent Fuel ◽  
Critical Importance ◽  
Dc Power ◽  
The Status ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
And Control ◽  
Parameter Values ◽  
Support Guidelines

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. There is evidence that the failure of key instrumentation to provide reliable information to the control room licensed operators contributed to the severity of the accident at both TMI and Fuskushima Daiichi. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data and yet have to make urgent decisions. While progress in these areas has been made since TMI-2, the accident at Fukushima suggests there may still be some potential for further improvement in critical plant instrumentation. As a result, several approaches are being employed to provide better information to emergency response personnel during a severe accident. The first approach being taken by the PWROG and BWROG is the identification of methods to obtain information related to key plant parameters when there is a loss of dc power for instrumentation and control. The FLEX guidance in NEI 12-06 requires that reliable instrumentation be available to ensure core, containment and spent fuel pool cooling is maintained for the beyond design basis events for which FLEX was intended. For the most part, this instrumentation that is important for FLEX is the same instrumentation that is used for diagnosis of severe accident conditions and challenges to fission product barriers. Generic FLEX Support Guidelines have been developed to provide a uniform basis for plants to meet the NEI 12-06 requirements that includes methods to obtain key parameter values in the event of a loss of all dc instrument power. The PWROG and the BWROG have also taken a complimentary approach to provide Technical Support Guidance (TSG) for instrumentation during a severe accident. This approach identifies the primary instrumentation as well as alternate instrumentation and other tools to validate the indications from the primary instrumentation. The validation consists of: a) comparing the primary instrument indications to the alternate instrumentation, b) comparing instrument indications to related instrumentation, c) comparing instrument indications and trends to expected trends based on the accident progression and actions already implemented, and d) comparing instrument indications to information in calculational aids.

Severe accident code-to-code comparison for two accident scenarios in a spent fuel pool

2018 ◽  
Vol 120 ◽  
pp. 880-887 ◽  
Author(s):  
O. Coindreau ◽  
B. Jäckel ◽  
F. Rocchi ◽  
F. Alcaro ◽  
D. Angelova ◽  
...  
Keyword(s):  
Severe Accident ◽  
Spent Fuel ◽  
Code Comparison ◽  
Spent Fuel Pool ◽  
Accident Scenarios

scholarly journals SAFEST Roadmap for Corium Experimental Research in Europe

2016 ◽  
Author(s):  
Christophe Journeau ◽  
Viviane Bouyer ◽  
Nathalie Cassiaut-Louis ◽  
Pascal Fouquart ◽  
Pascal Piluso ◽  
...  
Keyword(s):  
Experimental Research ◽  
Research Priorities ◽  
Large Mass ◽  
Severe Accident ◽  
Stress Tests ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Global Vision ◽  
Spent Fuel Pool ◽  
Experimental Facilities

SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on corium experimental research has been written to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. It is based on the research priorities determined by SARNET SARP group as well as those from the recently formulated in the NUGENIA Roadmap for severe accidents and the recently published NUGENIA Global Vision report. It also takes into account issues identified in the analysis of the European stress tests and from the interpretation of the Fukushima accident. 19 relevant issues related to corium have been selected during these prioritization efforts. These issues have been compared to a survey of the European corium experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. It shows a few lacks in EU corium infrastructures, especially in the domains of core late reflooding impact on source term, Reactor Pressure Vessel failure and corium release, Spent Fuel Pool accidents, as well as the need for a large mass (100–500 kg) prototypic corium facility.

Research and Development of Validation and Drill System for Full Scope of Severe Accident Management Guideline

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Zhifei Yang ◽  
Yali Chen ◽  
Hu Luo
Keyword(s):  
Research And Development ◽  
Computer Code ◽  
Severe Accident ◽  
Accident Analysis ◽  
Spent Fuel ◽  
Accident Management ◽  
Analysis Computer ◽  
Spent Fuel Pool ◽  
Accident Scenarios ◽  
Drill System

To respond to the urgent needs of verification, training, and drill for full scope severe accident management guidelines (FSSAMG) among nuclear regulators, utilities, and research institutes, the FSSAMG verification and drill system is developed. The FSSAMG includes comprehensive scenarios under power condition, shutdown condition, spent fuel pool (SFP) condition, and refueling conditions. This article summarized the research and development of validation and drill system for FSSAMG by using the severe accident analysis computer code modular accident analysis program 5 (MAAP5). Realistic accident scenarios can be verified and exercised in the developed system to support FSSAMG training, drill, examination, and verification.

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