Behavior of the Spent Fuel Pool and Dose Rate Calculations for the CPR1000 Reactor Under Severe Accident Conditions Using the Integral Code MELCOR

2013 ◽  
Author(s):  
Klaus Mueller ◽  
Moses Yeung ◽  
Justin Byard ◽  
Zhen Xun Peng ◽  
Jun Tao ◽  
...  
Keyword(s):  
Fission Product ◽  
Dose Rate ◽  
Flow Rate ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Pure Hydrogen ◽  
Spent Fuel ◽  
Product Release ◽  
Spent Fuel Pool ◽  
Accident Conditions

The behavior of the spent fuel pool and the fission product release and transport for the CPR1000 reactor under severe accident conditions was analyzed using the integral severe accident code MELCOR. In the investigated accident scenario a total failure of the pump of the spent fuel cooling system was assumed. Furthermore, it is assumed that accident management fails to bring water into the spent fuel pool using mobile pumps or due to the non-recovery of the cooling pump. The grace time available for measures in order to avoid significant fission product release to the environment is determined. The calculated hydrogen mass flow rate due to clad oxidation and the steam flow rate from the spent fuel pool to the compartment above the spent fuel pool serve as boundary conditions for the three dimensional fluid dynamics code GASFLOW to assess possible hydrogen combustion or detonation in the compartment. Using this spent fuel pool MELCOR model the dose submerged in air or water can be determined. The calculated gamma dose rate in a specific compartment can be used for equipment qualification and compartment accessibility assessment. It was found that after four days the fuel assemblies are significantly heated-up and ten hours later the fission products are released as well as a significant amount of hydrogen is produced. A preliminary GASFLOW analysis shows by assuming an air atmosphere in the fuel building, that the risk of a hydrogen combustion or detonation is high. In late state of the accident a convection flow of pure hydrogen is established in spent fuel pool region. It was shown, that the flow conditions strongly influence the fission product transport behavior and consequently the dose rates in the compartment above the spent fuel pool.

Sensitivity Analysis of RCS Depressurization Strategy Under a Postulated SGTR Accident in OPR1000

2018 ◽  
Author(s):  
Wonjun Choi ◽  
Taeseok Kim ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Sung Joong Kim
Keyword(s):  
Sensitivity Analysis ◽  
Fission Product ◽  
Steam Generator ◽  
Flow Rate ◽  
Sensitivity Study ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Feed Water ◽  
Relief Valve ◽  
Product Release

Steam generator tube rupture (SGTR) accident is one of important accident that has high probability of resulting in severe accidents. As a bypass scenario, fission product can be directly released to the environment during the SGTR accident. Thus, the severe accident by SGTR should be carefully managed by severe accident management guidance (SAMG). In Korea, SAMG for optimized power reactor 1000 (OPR1000) has been developed in 1999 and used to mitigate the severe accident of OPR1000 with seven mitigation strategies. Among the mitigation strategies, ‘Depressurization of reactor coolant system (RCS)’ is one of the most powerful strategies to reduce direct release of the fission product. To reduce the RCS pressure, indirect depressurization using steam generator is generally recommended. However, depending on the RCS condition, the indirect depressurization can be ineffective to reduce the RCS pressure. In this case, direct depressurization using pilot operated relief valve (PORV) should be performed as a second plan. From this point of view, sensitivity study of RCS depressurization was performed to investigate priority of depressurization in this study. The severe accident scenario initiated by SGTR accident was selected from probabilistic safety assessment (PSA) level 1 report and simulated using MELCOR 2.1. For the mitigation strategy, various timing of depressurization, the number of opening valves and flow rate of feed water were applied to simulate the possible depressurization strategies during the severe accident. The MELCOR code simulation shows that if depressurization was performed at 30 minutes after SAMG entrance, the direct depressurization was more efficient to reduce the RCS pressure and the fission product release. Therefore, it was recommended to use direct depressurization rather than indirect depressurization in certain time. The sensitivity of flow rate of feed water and different number of opening valves were insignificant for progress of the accident and fission product release. In conclusion, operators should select the way of depressurization to reduce the RCS pressure and the fission product release during the SGTR accident, considering the condition of the plant such as accident progress and availability of safety features. To suggest more proper information for depressurization, more sensitivity analysis and detailed thermal-hydraulic analysis should be performed for the future work.

Fission Product Release from Fuel under Severe Accident Conditions

1993 ◽  
Vol 101 (3) ◽  
pp. 270-281 ◽  
Author(s):  
Richard R. Hobbins ◽  
David A. Petti ◽  
Donald L. Hagrman
Keyword(s):  
Fission Product ◽  
Severe Accident ◽  
Product Release ◽  
Accident Conditions

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.

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

2007 ◽  
Vol 44 (11) ◽  
pp. 1428-1435 ◽  
Author(s):  
Tamotsu KUDO ◽  
Mitsuko KIDA ◽  
Takehiko NAKAMURA ◽  
Fumihisa NAGASE ◽  
Toyoshi FUKETA
Keyword(s):  
Fission Product ◽  
Severe Accident ◽  
Fuel Oxidation ◽  
Product Release ◽  
Accident Conditions

Calculation of Evaporation From Fukushima NPP Spent Fuel Pools

2018 ◽  
Author(s):  
Mirza M. Shah
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Spent Fuel ◽  
Decay Heat ◽  
Wide Range ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Safety Systems ◽  
Good Agreement

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and post-accident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The post-accident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25 % lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

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