scholarly journals The Fukushima accident was preventable

2015 ◽  
Vol 373 (2053) ◽  
pp. 20140379 ◽  
Author(s):  
Costas Synolakis ◽  
Utku Kânoğlu
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Hazard Analysis ◽  
Severe Accident ◽  
Regulatory Structure ◽  
Fukushima Accident ◽  
Critical Facilities ◽  
Safety Studies ◽  
New Research ◽  
Large Earthquakes

The 11 March 2011 tsunami was probably the fourth largest in the past 100 years and killed over 15 000 people. The magnitude of the design tsunami triggering earthquake affecting this region of Japan had been grossly underestimated, and the tsunami hit the Fukushima Dai-ichi nuclear power plant (NPP), causing the third most severe accident in an NPP ever. Interestingly, while the Onagawa NPP was also hit by a tsunami of approximately the same height as Dai-ichi, it survived the event ‘remarkably undamaged’. We explain what has been referred to as the cascade of engineering and regulatory failures that led to the Fukushima disaster. One, insufficient attention had been given to evidence of large tsunamis inundating the region earlier, to Japanese research suggestive that large earthquakes could occur anywhere along a subduction zone, and to new research on mega-thrusts since Boxing Day 2004. Two, there were unexplainably different design conditions for NPPs at close distances from each other. Three, the hazard analysis to calculate the maximum probable tsunami at Dai-ichi appeared to have had methodological mistakes, which almost nobody experienced in tsunami engineering would have made. Four, there were substantial inadequacies in the Japan nuclear regulatory structure. The Fukushima accident was preventable, if international best practices and standards had been followed, if there had been international reviews, and had common sense prevailed in the interpretation of pre-existing geological and hydrodynamic findings. Formal standards are needed for evaluating the tsunami vulnerability of NPPs, for specific training of engineers and scientists who perform tsunami computations for emergency preparedness or critical facilities, as well as for regulators who review safety studies.

scholarly journals Estimation of Dietary Intake of Radionuclides and Effectiveness of Regulation after the Fukushima Accident and in Virtual Nuclear Power Plant Accident Scenarios

2018 ◽  
Vol 15 (8) ◽  
pp. 1589 ◽  
Author(s):  
Michio Murakami ◽  
Takao Nirasawa ◽  
Takao Yoshikane ◽  
Keisuke Sueki ◽  
Kimikazu Sasa ◽  
...  
Keyword(s):  
Cost Effectiveness ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Dietary Intake ◽  
Emergency Preparedness ◽  
Nuclear Power ◽  
Food Distribution ◽  
Fukushima Accident ◽  
Accident Scenarios ◽  
Effectiveness Of Regulation

Evaluation of radiation exposure from diet is necessary under the assumption of a virtual accident as a part of emergency preparedness. Here, we developed a model with complete consideration of the regional food trade using deposition data simulated by a transport model, and estimated the dietary intake of radionuclides and the effectiveness of regulation (e.g., restrictions on the distribution of foods) after the Fukushima accident and in virtual accident scenarios. We also evaluated the dilution factors (i.e., ratios of contaminated foods to consumed foods) and cost-effectiveness of regulation as basic information for setting regulatory values. The doses estimated under actual emission conditions were generally consistent with those observed in food-duplicate and market-basket surveys within a factor of three. Regulation of restricted food distribution resulted in reductions in the doses of 54–65% in the nearest large city to the nuclear power plant. The dilution factors under actual emission conditions were 4.4% for radioiodine and 2.7% for radiocesium, which are ~20 times lower than those used in the Japanese provisional regulation values after the Fukushima accident. Strict regulation worsened the cost-effectiveness for both radionuclides. This study highlights the significance and utility of the developed model for a risk analysis of emergency preparedness and regulation.

Nuclear Power Safety Development and Accident Emergency in China

2012 ◽  
Vol 512-515 ◽  
pp. 2509-2514
Author(s):  
Zi Ying Jiang ◽  
Fan Yu
Keyword(s):  
Emergency Response ◽  
Emergency Preparedness ◽  
Nuclear Power ◽  
Power Development ◽  
Fukushima Accident ◽  
Nuclear Power Development ◽  
History Of ◽  
Potential Risks ◽  
Power Safety ◽  
Safety First

Nuclear power is clean, safe, but not zero risk, which has been evidenced by the history of nuclear power development. Nuclear accident emergency response is the final barrier of depth defense to reduce the potential risks that may arise from nuclear power development, which must be enhanced. The accident emergency preparedness in China and China responses to Fukushima accident are presented. Learning lessons from past nuclear power accidents (the Three Mile Island, Chernobyl and Fukushima), China would be keeping confidence in nuclear power development and advancing further improvement of emergency response capabilities to insist on the safety-first principle for nuclear power development.

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Public Domain ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Critical Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Radiological Aspects of Public Protection at Early Release Stage of NPP Severe Accident

2020 ◽  
pp. 62-73
Author(s):  
V. Bogorad ◽  
O. Slepchenko ◽  
T. Lytvynska ◽  
D. Bielykh ◽  
I. Kalyta ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Emergency Preparedness ◽  
Nuclear Power ◽  
Severe Accident ◽  
Modern World ◽  
Plant Design ◽  
Early Release ◽  
Safety Standards ◽  
General Physical

In the modern world, nuclear energy is one of the most economically feasible sources of energy. Although related risks are of a great public interest, they are quite obvious and predictable. The steady risk reduction trend is based on two main areas: improvement of nuclear power plant design basis and enhancement of emergency preparedness at all levels. Such a trend is maintained by more strict requirements for licensees from states and international competent authorities. Recently, practical elimination of an early radioactive release at a nuclear power plant is one of the most significant requirements for NPP safety. The paper covers issues focusing on how this requirement will be implemented within national radiation safety standards, namely within the regulations of the third group on intervention in the case of a radiological emergency, and whether Ukraine is ready to implement this standard for operating NPPs. The paper addresses the following issues: determination of an early release, available interval, theoretical grounds applied to assess an early release and public exposure doses. All the results presented in the paper are of an evaluative nature. The attention is focused not on specific features of a particular power unit or settlement, but on general physical principles of modeling the source term and protective properties of the premises.

Global nuclear industry views: challenges arising from the evolution of the optimisation principle in radiological protection

2012 ◽  
Vol 41 (3-4) ◽  
pp. 275-281 ◽  
Author(s):  
S. Saint-Pierre
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
Hereditary Diseases ◽  
Severe Accident ◽  
High Price ◽  
Radiological Protection ◽  
Operational Parameters ◽  
The Public ◽  
Risk Of Cancer

Over the last few decades, the steady progress achieved in reducing planned exposures of both workers and the public has been admirable in the nuclear sector. However, the disproportionate focus on tiny public exposures and radioactive discharges associated with normal operations came at a high price, and the quasi-denial of a risk of major accident and related weaknesses in emergency preparedness and response came at an even higher price. Fukushima has unfortunately taught us that radiological protection (RP) for emergency and post-emergency situations can be much more than a simple evacuation that lasts 24–48 h, with people returning safely to their homes soon afterwards. On optimisation of emergency and post-emergency exposures, the only ‘show in town’ in terms of international RP policy improvements has been the issuance of the 2007 Recommendations of the International Commission on Radiological Protection (ICRP). However, no matter how genuine these improvements are, they have not been ‘road tested’ on the practical reality of severe accidents. Post-Fukushima, there is a compelling case to review the practical adequacy of key RP notions such as optimisation, evacuation, sheltering, and reference levels for workers and the public, and to amend these notions with a view to making the international RP system more useful in the event of a severe accident. On optimisation of planned exposures, the reality is that, nowadays, margins for further reductions of public doses in the nuclear sector are very small, and the smaller the dose, the greater the extra effort needed to reduce the dose further. If sufficient caution is not exercised in the use of RP notions such as dose constraints, there is a real risk of challenging nuclear power technologies beyond safety reasons. For nuclear new build, it is the optimisation of key operational parameters of nuclear power technologies (not RP) that is of paramount importance to improve their overall efficiency. In pursuing further improvements in the international RP system, it should be clearly borne in mind that the system is generally based on protection against the risk of cancer and hereditary diseases. The system also protects against deterministic non-cancer effects on tissues and organs. In seeking refinements of such protective notions, ICRP is invited to pay increased attention to the fact that a continued balance must be struck between beneficial activities that cause exposures and protection. The global nuclear industry is committed to help overcome these key RP issues as part of the RP community's upcoming international deliberations towards a more efficient international RP system.

Numerical Two-Phase Flows Simulation and Analysis of the Evolution of the Local Hydrogen Concentration in a PWR Nuclear Containment in the Event of a Severe Accident

2016 ◽  
Author(s):  
Alexandre Zanchetti ◽  
Mickael Hassanaly ◽  
Hervé Cordier ◽  
Antonio Sanna ◽  
Namane Mechitoua ◽  
...  
Keyword(s):  
Hydrogen Concentration ◽  
Nuclear Power ◽  
Cfd Simulation ◽  
Severe Accident ◽  
Water Spray ◽  
Hydrogen Distribution ◽  
Two Phase ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Spray System

The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a water cooled reactor building. Some mitigation means against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and hydrogen combustion are necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on CFD simulation of hydrogen distribution in penalized conditions. After dealing with cases for which the water spray system was assumed to be unavailable, and so treated with single-phase CFD code [1] [2], the present paper content is now about simulation and analysis of the local hydrogen concentration in the case of a severe accident for which the water spray system is available. Numerical developments of a multi-phase CFD code (Neptune_CFD) and code validation lead to consistent simulations. The numerical simulation performed by EDF confirms the favorable safety impact of water spray on pressure and temperature for a LOCA scenario occurring on a 1300 MWe Pressurized Water Reactor. Nevertheless, CFD results show that the activation of the spray system before hydrogen injection gives greater hydrogen concentration. So, in the future, to better assess hydrogen risk, EDF will perform computations at CFD taking into account the interaction between combustion and water sprays.

CAE—The Spanish Emergency Support Center: A Centralized and Shared Emergency Support Service for Beyond Design Basis Events

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Rafael J. Caro
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Power Plants ◽  
Stress Test ◽  
Support Service ◽  
Fukushima Accident ◽  
Design Basis ◽  
Emergency Preparedness And Response ◽  
Response Organization ◽  
Center Service

The Spanish nuclear power generation industry proposed the development of an external Emergency Support Center as one of the measures to strengthen the nuclear safety and Emergency Preparedness and Response, as a consequence of the stress test developed after the Fukushima accident. The CAE project was carried out to define and establish a centralized service composed of intervention equipment and specialized personnel in the framework of an Emergency Support Center shared among all the Spanish nuclear power plants (NPPs). The emergency support service aims to strengthen the NPP emergency capabilities, by integrating with the Emergency Response Organization (ERO) of plants. This service successfully developed preoperational tests at each one of the Spanish NPPs in 2014. With these tests, the development of the different aspects that make up the Emergency Support Center service, at every Spanish NPP, was validated in four stages: (1) CAE mobilization in less than 24 hrs, (2) equipment deployment to its final location in the plant, (3) checking the connections with NPP’s interfaces, and (4) functionality of the equipment. The CAE proved in this way its capability to provide support to the Spanish NPPs, to strengthen their already strong characteristics in EP&R, to face these new extended damage/beyond design basis scenarios.

Utilization of NEK full scope simulator for plant operation optimization, nuclear education and engineering in 20 years

2021 ◽  
Author(s):  
Matjaž Žvar ◽  
Tomaž Žagar
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Impact Analysis ◽  
Training Session ◽  
Nuclear Safety ◽  
Severe Accident ◽  
Electricity Production ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water

Abstract This paper gives an impact analysis of utilization of NPP full scope simulator on operation parameters, training and education in nuclear power plant Krško. The Slovenian Nuclear Safety Administration issued their simulator decree to NEK in April 1995. The first training session on the simulator was performed in April 17th 2000 and since then the simulator has been used on daily bases to improve operator knowledges, skills and performances. At the time, this was the first full scope simulator with the capability to simulate Beyond design basis accidents (severe accidents). The ability to simulate core meltdown and containment breach made it very suitable for emergency preparedness drills. After the 2017 simulator upgrade, fuel meltdown in the spent fuel pool can be simulated using the Modular Accident Analysis Program – MAAP5. This capability is still unique for full scope simulators even today. The simulator is also used for pre-testing of plant modifications before their implementation on site or for just-in-time training for infrequent performed evolutions or for procedure development and testing. The Pressurized Water Reactor Owners Group (PWROG) used the NEK simulator in 2018 to develop the new set of the Severe Accident Management Guidelines, incorporated with a completely new usage approach. In all of these years, the simulator has been actively participating in the increased reliability and stability of the electricity production and in achieving NEK's vision to be a worldwide leader in nuclear safety and excellence.

Innovative Practice of Severe Accidents Measures in CAP1400

2017 ◽  
Author(s):  
Likai Fang ◽  
Xin Liu ◽  
Guobao Shi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Power Sources ◽  
Severe Accidents ◽  
Accident Management ◽  
Accident Conditions

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.

scholarly journals Simulation of a Station Blackout Accident for the SMART Using the CINEMA Code

2020 ◽  
Vol 8 ◽  
Author(s):  
Hyoung Tae Kim ◽  
Jin Ho Song ◽  
Rae-Joon Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Hydrogen Generation ◽  
Alternate Current ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Active Safety Systems ◽  
Station Blackout ◽  
Safety Systems

SMART is a small-sized integral type PWR containing major components within a single reactor pressure vessel. Advanced design features implemented into SMART have been proven or qualified through experience, testing, or analysis according to the applicable approved standards. After Fukushima accident, a rising attention is posed on the strategy to cope with a Station Blackout (SBO) accident, which is one of the representative severe accidents related to the nuclear power plants. The SBO is initiated by a loss of all offsite power with a concurrent failure of both emergency diesel generators. With no alternate current power source, most of the active safety systems that perform safety functions are not available. The purpose of SBO analysis in this paper is to show that the integrity of the containment can be maintained during a SBO accident in the SMART (System-integrated Modular Advanced ReacTor). Therefore, the accident sequence during a SBO accident was simulated using the CINEMA-SMART (Code for INtegrated severe accidEnt Management and Analysis-SMART) code to evaluate the transient scenario inside the reactor vessel after an initiating event, core heating and melting by core uncovery, relocation of debris, reactor vessel failure, discharge of molten core, and pressurization of the containment. It is shown that the integrity of the containment can be maintained during a SBO accident in the SMART reactor. It has to be mentioned that the assumptions used in this analysis are extremely conservative that the passive safety systems of PSIS and PRHRS were not credited. In addition, as ANS73 decay heat with 1.2 multiplier was used in this analysis, actual progression of the accident would be much slow and amount of hydrogen generation will be much less.

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