Issues of Insurance of Civil Liability for Nuclear Damage From Nuclear Low Power Plants

2014 ◽  
Author(s):  
Vladimir F. Demin ◽  
Vyacheslav P. Kuznetsov
Keyword(s):  
Life Cycle ◽  
Nuclear Power ◽  
Power Plants ◽  
Civil Liability ◽  
Severe Accident ◽  
Insurance Costs ◽  
Development Direction ◽  
Expert Analysis ◽  
Nuclear Damage ◽  
Total Life

In the frame of economic analysis of nuclear power (NP) with SMRs in their total life cycle the expert analysis of the problems associated with the civil liability for nuclear damage from SMRs on the example of transportable nuclear power units (TNUs) was performed. Purpose of the analysis is as follows: • Assessment of NP’ safety and economy changes in its development based on TNUs with KLT-40S and partly with RITM-200M reactor units. • Work out of recommendations on this development’ direction in terms of the insurance approach justification and amount of compensation for nuclear damage. The following aspects were considered in the analysis: 1. National and international approaches and practice of nuclear insurance. 2. Specific features of TNUs and differences from large NPPs basing on example of the design of floating power unit FPU “Academician Lomonosov” with KLT-40S reactors. 3. Assessment of severe accident consequences during TNU’ life cycle. 4. Analysis of insurances’ approaches and assessment of possible insurance costs.

Issues of Safety and Civil Liability Insurance for Nuclear Damage from Small Nuclear Power Plants

2019 ◽  
Vol 64 (6) ◽  
pp. 31-36
Author(s):  
V. Demin ◽  
A. Golosnaya ◽  
S. Korolev ◽  
V. Kuznetsov ◽  
V. Makarov ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Civil Liability ◽  
Third Parties ◽  
Liability Insurance ◽  
Financial Costs ◽  
Design Basis ◽  
Nuclear Damage

Purpose: To study the possibility of achieving assured safety for the environment and public in all modes of operation of small nuclear power plants (SNPP) and providing real civil liability insurance for nuclear risks at reasonable financial costs. Material and methods: Particular attention on small nuclear power plants is driven by regional development, local communities and productions, which are not covered by centralized transport and energy supply. The peculiar properties and benefits of energy production at SNPP are considered, including: the possibility of locating in remote regions; the short construction period and the modular structure of SNPP; availability of potential to improve safety and reliability; reducing the size of the sanitary protection zone up to the boundaries of the technological site; the reality of liability insurance (full financial responsibility of the operator) for nuclear damage to third parties caused by an accident at SNPP at reasonable financial costs; industrial serial production; ability to move the entire nuclear power plants with small modular reactors in the assembled form, etc. A comparative analysis of the technical characteristics of the SNPP and a conventional nuclear power plant from a safety perspective is made. Results: The results of the SNPP safety analysis performed on the basis of the design documentation of the floating nuclear power plant “Akademik Lomonosov” is presented, with particular attention to assessing the consequences of design and beyond design basis accidents, in terms of probabilistic safety analysis and assessment of the maximum possible damage to third parties. The maximum possible damage to third parties from severe accidents is estimated to be about 0.5 billion RUR, which is hundreds of times less than damage from a catastrophic accident at a conventional NPP. Estimated costs for insurance of damage to third parties from an accident at SNPP will not exceed 1 kopeck/kWh. Possible approaches to civil liability insurance for nuclear risks and aspects of legal support are considered. Conclusions: The results of the analysis allow to conclude that it is possible to provide in the future: the achievement of practically assured safety of the SNPP for the environment and the public in normal operation and possible design and beyond design basis accidents; real civil liability insurance for nuclear risks of SNPP at reasonable financial costs.

scholarly journals A decision-support approach to severe accident management in nuclear power plants

2020 ◽  
pp. 1-12
Author(s):  
Marko Bohanec ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
Klemen Debelak ◽  
Luka Štrubelj
Keyword(s):  
Decision Support ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems

2018 ◽  
Author(s):  
P. Papadopoulos ◽  
T. Lind ◽  
H.-M. Prasser
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
Tube Bundle ◽  
Droplet Size Distribution ◽  
Severe Accident ◽  
Design Element ◽  
Venturi Scrubber

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.

scholarly journals Spreading of Excellence in SARNET Network on Severe Accidents: The Education and Training Programme

2012 ◽  
Vol 2012 ◽  
pp. 1-9 ◽  
Author(s):  
Sandro Paci ◽  
Jean-Pierre Van Dorsselaere
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Training Programme ◽  
Education And Training ◽  
Severe Accident ◽  
Research Network ◽  
Safety Issues ◽  
Severe Accidents ◽  
Management Procedures ◽  
And Training

The SARNET2 (severe accidents Research NETwork of Excellence) project started in April 2009 for 4 years in the 7th Framework Programme (FP7) of the European Commission (EC), following a similar first project in FP6. Forty-seven organisations from 24 countries network their capacities of research in the severe accident (SA) field inside SARNET to resolve the most important remaining uncertainties and safety issues on SA in water-cooled nuclear power plants (NPPs). The network includes a large majority of the European actors involved in SA research plus a few non-European relevant ones. The “Education and Training” programme in SARNET is a series of actions foreseen in this network for the “spreading of excellence.” It is focused on raising the competence level of Master and Ph.D. students and young researchers engaged in SA research and on organizing information/training courses for NPP staff or regulatory authorities (but also for researchers) interested in SA management procedures.

Effective Software Verification and Validation Approach for Nuclear Power Plant Digital Instrumentation and Control Systems

2010 ◽  
Author(s):  
Steve Yang ◽  
Jun Ding ◽  
Huifang Miao ◽  
Jianxiang Zheng
Keyword(s):  
Life Cycle ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Software Verification ◽  
Verification And Validation ◽  
Common Cause ◽  
Digital Instrumentation ◽  
And Control ◽  
Software Verification And Validation

All 1000 MW nuclear power plants currently in construction or projected to-be-built in China will use the digital instrumentation and control (I&C) systems. Safety and reliability are the ultimate concern for the digital I&C systems. To obtain high confidence in the safety of digital I&C systems, rigorous software verification and validation (V&V) life-cycle methodologies are necessary. The V&V life-cycle process ensures that the requirements of the system and software are correct, complete, and traceable; that the requirements at the end of each life-cycle phase fulfill the requirements imposed by the previous phase; and the final product meets the user-specified requirements. The V&V process is best illustrated via the so-called V-model. This paper describes the V-model in detail by some examples. Through the examples demonstration, it is shown that the process detailed in the V-model is consistent with the IEEE Std 1012-1998, which is endorsed by the US Regulatory Guide 1.168-2004. The examples show that the V-model process detailed in this paper provides an effective V&V approach for digital I&C systems used in nuclear power plants. Additionally, in order to obtain a qualitative mathematical description of the V-model, we study its topological structure in graph theory. This study confirms the rationality of the V-model. Finally, the V&V approach affording protection against common-cause failure from design deficiencies, and manufacturing errors is explored. We conclude that rigorous V&V activities using the V-model are creditable in reducing the risk of common-cause failures.

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

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