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.

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.

FLEX Loss of Instrumentation Guidance for PWRs Enhances Severe Accident Diagnostics

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
James Lynde ◽  
Steven Pierson
Keyword(s):  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Spent Fuel ◽  
Pressurized Water ◽  
The Core ◽  
Design Basis ◽  
External Events ◽  
Accident Conditions ◽  
Specific Implementation ◽  
Support Guidelines

The industry response to the Nuclear Regulatory Commission (NRC) Order EA-12-049 is based on a set of Diverse and Flexible Coping Strategies (commonly referred to as FLEX) for beyond design basis external events as described in NEI 12-06. The Pressurized Water Reactors Owners Group (PWROG) developed generic guidance for response to these Beyond Design Basis External Events (BDBEE), called FLEX Support Guidelines (FSGs). These guidelines are referenced from the plant Emergency Operating Procedures (EOPs) when it is determined that an event exhibits certain beyond design basis characteristics such as an Extended Loss of all AC Power (ELAP). These generic FLEX Support Guidelines provide a uniform basis for all PWRs to implement the FLEX guidance in NEI 12-06 that was endorsed by the NRC to maintain core, containment and spent fuel cooling. The PWROG generic FSGs include guidance in FSG-7, “Loss of Vital Instrumentation or Control Power” for obtaining information for key plant parameters in an ELAP event. The key parameters were selected based on industry guidance and plant specific implementation. This set of key parameters will allow the licensed operators to have vital instrumentation to safely shutdown the core and maintain the core in a shutdown condition, including core, containment and spent fuel pool cooling. These parameters are used in the EOPs as well as the FSGs that are designed to mitigate a beyond design basis event. The requirements of NEI 12-06, as implemented through the FSGs, enhance both availability and reliability of instrumentation by requiring diverse methods of providing DC power for instrumentation and control as well as protection of instrumentation from the beyond design basis event. The subsequent implementation of this guidance at the Byron Station has proven to also be beneficial for diagnosis of severe accident conditions (where core cooling could not be maintained). The same parameter values that are needed to verify core, containment and spent fuel cooling prior to core damage are also needed to diagnose severe accident conditions. Guidance provided within FSG-7, as implemented at the Byron Station, contains several layers of diverse methods to obtain parametric values for key variables that can be especially useful when the environmental qualification is exceeded for the primary instrumentation that provides this information. The methods range from the use of self-powered portable monitoring equipment to the use of local mechanical instrumentation. The FSG-7 guidance is referenced from the Byron Severe Accident Management Guidance (SAMG) to either obtain parameter information during a severe accident or to validate the information that is available from the primary instrumentation.

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.

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.

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.

Evaluation of Alternate Water Injection Cooling for Accident Conditions in Spent Fuel Pool Using MAAP Code

2019 ◽  
Vol 2019 (0) ◽  
pp. S08106
Author(s):  
Satoshi NISHIMURA ◽  
Masaaki SATAKE ◽  
Yoshihisa NISHI ◽  
Yoshiyuki KAJI ◽  
Yoshiyuki NEMOTO
Keyword(s):  
Water Injection ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Accident Conditions

Performance of Spent Fuel Transportation Casks in Severe Tunnel Fires

2006 ◽  
Author(s):  
Earl P. Easton ◽  
Christopher S. Bajwa ◽  
Robert Lewis
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Radioactive Material ◽  
Severe Accident ◽  
Spent Fuel ◽  
Performance Study ◽  
Fire Scenario ◽  
Tunnel Fire ◽  
Accident Conditions

As part of the Nuclear Regulatory Commission's (NRC) overall review of the performance of transportation casks under severe accident conditions, the NRC has undertaken a number of initiatives, including the Package Performance Study (PPS), described in USNRC Package Performance Study Test Protocols, NUREG-1768, which will test full size transportation casks in a severe accident, as well as an examination of the Baltimore tunnel fire of 2001. The final PPS test plan is currently under development by the NRC's Office of Research. The NRC, working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of three different spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of these analyses are described in detail in Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario, NUREG/CR-6886, published in draft for comment in November 2005.

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