Regulatory Actions That Hinder Development of Effective Risk Reduction Measures by the Nuclear Industry for Enhanced Severe Accident Prevention and Mitigation Measures After Fukushima

Author(s):  
Sunil Nijhawan

The official report of The Fukushima Nuclear Accident Independent Investigation Commission concluded that “The TEPCO Fukushima Nuclear Power Plant accident was the result of collusion between the government, the regulators and TEPCO, and the lack of governance by said parties. They effectively betrayed the nation’s right to be safe from nuclear accidents. Therefore, we conclude that the accident was clearly ‘manmade.’ We believe that the root causes were the organizational and regulatory systems that supported faulty rationales for decisions and actions, rather than issues relating to the competency of any specific individual.” This wakeup call for the nuclear power utilities should require a public review of their relationship with of regulators. However, severe accident related risk reduction is a relatively uncharted territory and given the apparent lack of in-house technical expertise, the regulators are heavily relying on the qualitative and ‘hand waving’ arguments being presented by the utilities inherently disinterested in further investments they are not required to make under original license conditions. As a result, it has accelerated further deterioration of the safety culture and emboldened many within the regulatory staff to undertake or support otherwise questionable decisions in support of the utilities that prefer status quo. Case in point is the Canadian Nuclear Safety Commission (CNSC) which mostly accepts any and all requests by the nuclear power industry. After Fukushima, the CNSC took a year to publish a set of ‘Action Items’ for the Canadian Nuclear industry to prepare plans over 3 years and then accepted most if not all submissions that in many cases barely addressed the already watered down recommendations. In some cases the solutions proposed by the industry were economically expedient but technically flawed; and some could even be considered dangerous. CNSC also published a study on consequences of a severe accident with a source term that was limited to the desirable safety goal (100 TBq of Cs-137), which coincidently years later matched the utility ‘calculations’, but orders of magnitude smaller than predicted by independent evaluations. As a result, some well publicized conclusions on the benign nature of consequences of a CANDU severe accident were made and the local and provincial agencies that actually are supposed to prepare off-site emergency measures were left with an incorrect picture of what havoc a severe accident can cause otherwise. CNSC then published a much publicized video highlighting the available operator actions to terminate the accident early and later a report outlining the accident progression for a severe accident without operator action with conclusions that were immediately technically suspect from a variety of aspects. The aim was to claim that a severe core damage accident has no unfavorable off-site consequences. The regulator effectively, in this case, comes across as a promoter for the industry it is legislated to regulate. The paper outlines examples of actions being taken by the regulators that hinder development of effective risk reduction measures by the industry which otherwise would be forced to undertake them if the regulators had not stepped on the plate to bat for them. They vary from letters to editors to silence any safety concerns raised by the public, muzzling of its own staff, trying to silence external specialists who question their wisdom on to blatant disregard for any intervention by public they are required to entertain by law but are accustomed to factually ignore or belittle. The paper also outlines a number of examples of actions that an independent regulator would undertake to reduce the risk and enhance the safety culture. The nuclear regulatory regimes work well generally but in cases where it does not, the results can be disastrous as evident from the events in Japan and as is building up in Canada. The paper also summarizes the disparities between the number of Regulatory Actions instituted by the CNSC against small companies that use nuclear substances for industrial applications and almost none actions against the nuclear power plant utilities it regularly grants a pass in spite of the larger risk their operations pose to public.

Author(s):  
Vasilij V. Begun ◽  
Sergij V. Begun ◽  
Olena O. Kilina

The necessity of safety analysis methods and probable scenarios of accidents teaching in the education of experts for nuclear industry in Ukraine has been realised only after the Chernobyl accident. We developed the content of the first educational course in probabilistic safety analysis in 1995 based on the experience of the countries having developed nuclear power, the USA first of all, and on the training course of the Idaho National Laboratory. After this in 1996 the new course in probabilistic safety analysis of nuclear power plant (NPP) was adopted at our university. The new educational course in safety for students was developed and adopted in 2009 educational year - “Safety culture at nuclear installations of Ukraine”. Education and training in safety culture in higher educational institutions and in the nuclear power plants is a part of the general modern process of maintenance of safety, it is recommended by IAEA standards. The principles of safety culture are taken as a basis of the modern concept of safety of nuclear power plants. This work has received a positive appreciation from the management of departments of safety and training of the personnel of operating organization National Nuclear Energy Generating Company Energoatom (NNEGC Energoatom) and from other leaders of nuclear industry. The content of this educational course was discussed at the international scientific conferences on safety culture in 2008 and 2010, and was preliminary printed in the professional journal «Nuclear and radiation safety». The purposes of education have been defined as a survey, generalizing course on safety of the NPP with an allocation of safety issues on the foreground. Practical questions of the equipment and NPP systems work, their interaction in emergencies and the role of the human-operator are studied. The procedure of failure analysis at NPP is studied. Students analyze equipment work, root and direct causes of incidents. Methods of estimation of safety conditions based on observable operational indicators are studied. Parameters, variables and indicators of safety culture are studied. As a result of gained experience we have come to the conclusion about high advisability of educational courses in safety for students. Specially formed knowledge and education in the field of safety from a student’s bench are the basis of safety culture of the future nuclear industry expert.


Author(s):  
Gary Park

The nuclear industry is a pretty dynamic industry, in that it is always on the move, changing every time we turn around. For that very reason, there is a need to keep up with the industry by providing changes to American Society of Mechanical Engineering Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” There have been many changes over the last three years. This paper addresses a few of those, but gives a feel for the number of changes from the 2000 Addenda to the 2003 Addenda, there have been a total of approximately 56 changes. Of those changes, 11 were in the repair/replacement requirements, 19 in the inspection requirements, 4 in the evaluation requirements, 18 in the nondestructive examination requirements, and 4 in the administrative requirements. The paper classifies the changes as “Technically Significant,” “Significant,” “Non-Significant,” or “Editorial.” The paper addresses only a few of those changes that were “Technically Significant.” The paper also includes some of the activities that the ASME Section XI Subcommittee is currently working on.


2014 ◽  
Vol 989-994 ◽  
pp. 2097-2100
Author(s):  
Zheng Zhang ◽  
Hai Bo He ◽  
Hao Liang Lu

In order to satisfy the calculation requirements of nuclear power plant operating in different conditions, the integration and combination of reactor core computation modules have been proposed. By writing logical language instructions, and then read by interpreter, the engineering designers can make grammatical analysis, lexical analysis, semantic analysis and information extraction. In Linux system environment, the interpreter can fulfill computational tasks based on the actual operating parameters of nuclear power plant. The comparison results indicate that the calculated results obtained by the interpreter language are correct. Therefore, it also demonstrates that the interpreter language is valid.


2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Kevin Fernández-Cosials ◽  
Gonzalo Jiménez ◽  
César Serrano ◽  
Luisa Ibáñez ◽  
Ángel Peinado

During a severe accident (SA) in a nuclear power plant (NPP), there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system (FCVS) can be used to retain iodine and aerosols radioactive releases coming from the containment atmosphere. However, during a SA, large quantities of hydrogen can also be generated. Hydrogen reacts violently with oxygen and its combustion could impair systems, components, or structures. For this reason, to protect the integrity of the FCVS against hydrogen explosions, an inertization system is found necessary. This system should create an inert atmosphere previous to any containment venting that impedes the contact of hydrogen and oxygen. In this paper, the inertization system for Cofrentes NPP is presented. It consists of a nitrogen injection located in three different points. A computational model of the FCVS as well as the inertization system has been created. The results show that if the nitrogen sweeps and the containment venting are properly synchronized, the hydrogen risk could be reduced to a minimum and therefore, the integrity of the FCVS would be preserved.


2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Kwame Gyamfi ◽  
Sylvester Attakorah Birikorang ◽  
Emmanuel Ampomah-Amoako ◽  
John Justice Fletcher

Abstract Atmospheric dispersion modeling and radiation dose calculation have been performed for a generic 1000 MW water-water energy reactor (VVER-1000) assuming a hypothetical loss of coolant accident (LOCA). Atmospheric dispersion code, International Radiological Assessment System (InterRAS), was employed to estimate the radiological consequences of a severe accident at a proposed nuclear power plant (NPP) site. The total effective dose equivalent (TEDE) and the ground deposition were calculated for various atmospheric stability classes, A to F, with the site-specific averaged meteorological conditions. From the analysis, 3.7×10−1 Sv was estimated as the maximum TEDE corresponding to a downwind distance of 0.1 km within the dominating atmospheric stability class (class A) of the proposed site. The intervention distance for evacuation (50 mSv) and sheltering (10 mSv) were estimated for different stability classes at different distances. The intervention area for evacuation ended at 0.5 km and that for sheltering at 1.5 km. The results from the study show that designated area for public occupancy will not be affected since the estimated doses were below the annual regulatory limits of 1 mSv.


Author(s):  
Frank Kretzschmar

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.


Author(s):  
Jean-Jacques Grenouillet

Nowadays, decommissioning of nuclear power plants has become a key issue for nuclear industry in Europe. The phasing out of nuclear energy in Germany, Belgium and Sweden, as well as the early closure of nuclear units in applicant countries in the frame of EU enlargement, has largely contributed to consider decommissioning as the next challenge to face. The situation is slightly different in France where nuclear energy is still considered as a safe, cost-effective and environment friendly energy source. Electricite´ de France (EDF) is working on the development of a new generation of reactor to replace the existing one and erection of a new nuclear power plant could start in the next few years. Nevertheless, to achieve this objective, it will be necessary to get the support of political decision-makers and the acceptance of public opinion. Due to the growing concern of these stakeholders for environmental issues, their support can only be obtained if it is possible to demonstrate that nuclear energy industry will not leave behind unsolved issues that will be a burden to the next generations. In this context decommissioning of the first generation of EDF NPPs constitutes a prerequisite for the erection of a new type of nuclear power plant. This paper will present the programme defined by EDF for the decommissioning of its nine already shutdown reactors (Fig. 1). The reasons of the recent evolution of EDF decommissioning strategy will be explained and the key issues that will contribute to the successful implementation of this programme will be addressed. Finally, what has been achieved on sites so far and major planned activities will be described.


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