Post-Fukushima Assessment of the AP1000® Plant

2014 ◽  
Author(s):  
Terry L. Schulz ◽  
Thomas A. Kindred ◽  
Bryan N. Friedman ◽  
John E. Glavin ◽  
Adam D. Malinowski ◽  
...  
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Fuel Oil ◽  
Passive Safety ◽  
Final Report ◽  
Power Station ◽  
Pressurized Water ◽  
Passive Coping

The AP1000 plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, and safety with reduced plant costs. The AP1000 passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at the Fukushima Daiichi nuclear power station without compromising core and containment integrity. Without AC power, using passive safety technology, the AP1000 plant provides cooling for the core, containment and spent fuel pool for more than 3 days without the need for operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. With the re-supply of fuel oil the coping time may be extended for an indefinite time. Connections for a few, small, easily transportable components provide a diverse backup means of extending passive system operation after the first 3 days. As a result, the AP1000 design provides very robust protection of public safety and the utility investment. Following the accident at the Fukushima Dai-ichi nuclear power station in Japan, several initiatives were launched worldwide to assess the lessons learned. These include, but are not limited to, the European Nuclear Safety Regulators Group (ENSREG) stress tests, the Office for Nuclear Regulation (ONR) Final Report, the International Atomic Energy Agency (IAEA) Expert Mission Report, and the U.S. NRC Near-Term Task Force Recommendations. The AP1000 design has been assessed against these initiatives and lessons learned. The purpose of this paper is to describe: • How the accident at the Fukushima Dai-ichi nuclear power station was evaluated and translated into conclusions and recommendations for nuclear power plants worldwide • How the AP1000 plant was evaluated in light of the recommendations resulting from the various post-Fukushima assessments • The key conclusions resulting from the post-Fukushima evaluation of the AP1000 design

Incorporation of the Radioactive Interinfluence in Multi-Unit Seismic PRA

2018 ◽  
Author(s):  
Shuhei Matsunaka ◽  
Chikahiro Sato ◽  
Manabu Watanabe
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Common Factor ◽  
Radioactive Material ◽  
Power Station ◽  
Core Damage ◽  
Accident Management ◽  
Multiple Units

Kashiwazaki-Kariwa nuclear power station of TEPCO is the largest nuclear power station in the world, and it has seven nuclear power plants. As the experience at Fukushima Daiichi nuclear power station accident in March 2011 involving concurrent core damage at multiple units, it is considered that the risk derived from hazards of Earthquakes and Tsunamis is relatively significant in Japan, and these events have a high likelihood of damaging multiple units simultaneously. Therefore, it is very important to grasp the multi-unit specific risk. Although there are some unique accident scenarios of Multi-Unit PRA, this paper focuses on the influence of radioactive materials released outside the containment vessel on the accident management of the adjacent unit. The events including core damage and loss of containment function should be considered as the causes of the release of radioactive substances, and operator’s operation or the like should be considered as objects to be adversely affected by them. It is necessary to incorporate that into PRA to confirm the effect on risk. It is very difficult in terms of the maturity of evaluation method and the calculation load to accurately incorporate consequences derived from time series of various events and complicated interaction into PRA model. Therefore, as the first step in evaluating the risk of influence of radioactive material release on the accident management, some streamlining efforts are implemented according to the purpose. For example, Kashiwazaki-Kariwa unit 6 and unit 7 were set as the target units for model simplification. We also assume the earthquake as the initiating event due to the strong common factor for multi units. Whether or not to be operable in the adjacent plant is set conservatively based on deterministic evaluation. PRA taking into consideration the radiation influence by multi-unit accident is compared with normal PRA. Some kind of Core Damage Frequency (CDF) such as CDF1 (Core Damage Frequency at which the damage of one or the other of two unit occur), CDF2 (CDF at which the damage of both of units occur) and CDFTOTAL (CDF at which the damage of one or more units occur: CDF1 + CDF2) are quantified, and the degree of this issue is provided. Although the change of CDFTOTAL was insignificant, the necessity of further study was shown from the viewpoint of the amount and timing of radioactive substance released due to an approximately 1.5-fold increase in CDF2.

Lessons Learned in Implementation of Guidelines for Quantifying Support System Initiating Event Frequencies

2008 ◽  
Author(s):  
Rupert A. Weston ◽  
Ashley J. Mossa
Keyword(s):  
Support System ◽  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Common Cause Failures ◽  
Areas Of Concern ◽  
Pilot Implementation ◽  
American Society

The pilot implementation results for Regulatory Guide 1.200 identified four probabilistic risk assessment (PRA) technical elements that required additional guidance. One of these elements involved the use of fault tree technique to quantify the frequencies of support system initiating events (SSIEs). To address this technical element, guidelines were developed by the Electric Power Research Institute (EPRI) to provide a common industry approach for addressing the identification and quantification of SSIEs. The EPRI guidelines were issued as an interim report to allow trial use and pilot implementation by the industry prior to finalizing the guidelines. These interim guidelines provide an industry-consensus approach for addressing areas of concern in the development of support system initiating event models to ensure that the associated supporting requirements of the American Society of Mechanical Engineers (ASME) PRA Standard for internally initiated events are satisfied. A Pressurized Water Reactor Owners Group (PWROG) pilot implementation of the EPRI interim guidelines was conducted to determine whether the pilot participants have adequately addressed all areas of concern in the development of SSIE models. To determine this, a SSIE model currently used was selected by each of four the pilot participants and subject to detail review to demonstrate whether these models meet the expectations of the EPRI interim guidelines. The EPRI interim guidelines identified the areas of concern to be addressed in using fault tree technique to develop and quantify SSIE models. The guidelines addressed several areas of concern including the treatment of passive failures, the assignment of an appropriate mission time for primary and secondary failures, treatment of common cause failures (CCFs) between running and standby equipment, and consideration of all combinations of CCFs. The PWROG pilot implementation of the interim guidelines summarized the lessons learned and provided feedback to EPRI for consideration in finalizing the guidelines. In addition to the compilation of lessons learned, the PWROG implementation of the EPRI interim guidelines identified existing practices used to develop fault tree models for quantifying SSIE frequencies. Such practices did not necessarily follow a common approach and did not fully meet the expectations of the interim guidelines. Detailed reviews of the SSIE models currently in use at nuclear power plants (NPPs) for the pilot participants demonstrated that the elements of evaluation described in the interim guidelines were not addressed consistently among the PWROG pilot participants. Recommended improvements were identified and incorporated in the SSIE models to meet the expectations of the EPRI interim guidelines. The re-quantification of SSIE frequencies based on the recommended improvements, demonstrated that by not adequately addressing all elements in the evaluation, the SSIE frequency may be under-estimated.

The Use of Robots to Respond to Nuclear Accidents: Applying the Lessons of the Past to the Fukushima Daiichi Nuclear Power Station

2020 ◽  
Vol 4 (1) ◽  
Author(s):  
Yasuyoshi Yokokohji
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Autonomous Systems ◽  
Lessons Learned ◽  
Annual Review ◽  
Publication Date ◽  
Nuclear Accidents ◽  
Power Station ◽  
Fukushima Daiichi ◽  
Further Development

It has been 10 years since the Fukushima Daiichi Nuclear Power Station (NPS) accident. This article begins by discussing the robots used during the responses to the Three Mile Island and Chernobyl nuclear accidents. It then reviews the robots used to respond to the Fukushima Daiichi NPS accident, while considering the lessons learned from the previous accidents. Such discussions will hopefully lead to the further development of robots for decommissioning the Fukushima Daiichi NPS. Expected final online publication date for the Annual Review of Control, Robotics, and Autonomous Systems, Volume 4 is May 3, 2021. Please see http://www.annualreviews.org/page/journal/pubdates for revised estimates.

Incefa-Plus Project: Lessons Learned From the Project Data and Impact on Existing Fatigue Assessment Procedures

2020 ◽  
Vol 142 (6) ◽  
Author(s):  
Sam Cuvilliez ◽  
Alec McLennan ◽  
Kevin Mottershead ◽  
Jonathan Mann ◽  
Matthias Bruchhausen
Keyword(s):  
Surface Finish ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Fatigue Assessment ◽  
Horizon 2020 ◽  
Project Data ◽  
Assessment Procedures

Abstract This work focuses on the analysis of the data generated during the INCEFA+ project (INcreasing safety in nuclear power plants (NPPs) by Covering gaps in Environmental Fatigue Assessment, a five-year project supported by the European Commission Horizon 2020 program). More specifically, this paper discusses how the outcome of this analysis can be used to evaluate existing fatigue assessment procedures that incorporate environmental effects in a similar way to NUREG/CR-6909. A key difference between these approaches and the NUREG/CR-6909 is the reduction of conservatisms resulting from the joint implementation of the adjustment subfactor related to surface finish effect (as quantified in the design air curve derivation) and a Fen penalization factor for fatigue assessment of a location subjected to a pressurized water reactor (PWR) primary environment. The analysis presented in this paper indicates that the adjustment (sub-) factor on life associated with the effect of surface finish in air (as described in the derivation of the design air curve in NUREG/CR-6909) leads to substantial conservatisms when it is used to predict fatigue lifetimes in PWR environments for rough specimens. The corresponding margins can be explicitly quantified against the design air curve used for environmentally assisted fatigue (EAF) assessment, but may also depend on the environmental correction Fen factor expression that is used to take environmental effects into account.

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