Engineering Assessment Approach for the Leningrad Nuclear Power Plant In-Depth Safety Assessment

2000 ◽  
Author(s):  
Garill Coles ◽  
Sam McKay ◽  
Jon Young ◽  
Yuri Skok
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Safety Analysis ◽  
Regulatory Compliance ◽  
Regulatory Requirements ◽  
Related Function ◽  
Assessment Approach ◽  
Leningrad Nuclear Power Plant

Abstract Engineering assessment that supports the safety basis for a reactor plant operating license is defined as: “An assessment of a system to determine its adequacy to successfully perform its safety-related function(s) when required.” The approach to engineering assessment of systems at the Leningrad Nuclear Power Plant (LNPP), as part of its in-depth safety assessment (ISA), is unique. The content and format of engineering assessments for western Safety Analysis Reports (SARs) have evolved over time and current requirements are somewhat scattered in the governing documents (USNRC, 1978). Many regulatory guides and requirements (western or eastern) have not kept up with changes in safety analysis technology. Performance of the ISA for LNPP affords the opportunity to rethink the approach to engineering assessments, and to incorporate current methods and latest technology in safety analysis. As an example, for many systems, information about system reliability obtained from a modem Probabilistic Safety Assessment is more comprehensive than that from a Single Failure Analysis as prescribed in SAR content and format guides. Overall, the engineering assessment of LNPP systems looks at five major assessment elements: 1) assessment of regulatory compliance, 2) assessment of operability, 3) assessment of vulnerability, 4) assessment of environmental qualifications, and 5) assessment of reliability. By reorganizing the approach to meeting regulatory requirements, and by looking at engineering assessment in various ways, information can be obtained that goes beyond simply demonstrating regulatory compliance to more fully supporting the safety basis for a plant operating license.

Research and Software Implementation of Seismic Probabilistic Safety Analysis Quantification Method

2017 ◽  
Author(s):  
Wei Gao ◽  
Guofeng Tang ◽  
Jingyu Zhang ◽  
Qinfang Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Fragility Curves ◽  
Safety Analysis ◽  
Fukushima Accident ◽  
Binary Decision ◽  
Plant Level ◽  
Accurate Quantification ◽  
Probabilistic Safety

Seismic risk of nuclear power plant has drawn increasing attention after Fukushima accident. An intensive study has been carried out in this paper, including sampling of component and structure fragility based on Monte Carlo method, fragility analysis on system or plant level, convolution of seismic hazard curves and fragility curves. To derive more accurate quantification results, the binary decision diagram (BDD) algorithm was introduced into the quantification process, which effectively reduces the deficiency of the conventional method on coping with large probability events and negated logic. Seismic Probabilistic Safety Analysis (PSA/PRA) quantification software was developed based on algorithms discussed in this paper. Tests and application has been made for this software with a specific nuclear power plant seismic PSA model. The results show that this software is effective on seismic PSA quantification.

A Quantitative Approach for Reliability Evaluation of Safety I&C Systems in Nuclear Power Plants

2017 ◽  
Author(s):  
Liu Dongxu ◽  
Xu Dongling ◽  
Zhang Shuhui ◽  
Hu Xiaoying
Keyword(s):  
Probability Theory ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Boolean Algebra ◽  
Nuclear Power ◽  
Power Plants ◽  
Block Diagram ◽  
Reliability Assessment ◽  
Simple Approach ◽  
Assessment Approach

The probability that the safety I&C system fails to actuate or advertently actuates RT or ESF functions, in part, essentially determines whether a nuclear power plant could operate safely and efficiently. Since more conservative assumptions and simplifications are introduced during the analysis, this paper achieves solid results by performing the modeling and calculation based on a relatively simple approach, the reliability block diagram (RBD) method. A typical safety I&C platform structure is involved in the model presented in this paper. From the perspective of conservation and simplicity, some assumptions are adopted in this paper. A group of formulas is derived in this paper based on Boolean algebra, probability theory, basic reliability concepts and equations, to facilitate the calculations of probabilities that the safety I&C system fails to actuate or advertently actuates RT or ESF functions. All the inputs of the analysis and calculation in this paper, which includes the I&C platform structure, the constitution of the hardware modules, and reliability data, are referenced to the nuclear power plant universal database where applicable. Although the conclusion drawn in the paper doesn’t apply to the I&C platform assessment for a specific plant, the method of modeling and process of analysis provides an illustration of an alternative quantitative reliability assessment approach for a typical safety I&C system installed in the nuclear power plant.

scholarly journals Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2013 ◽  
Vol 479-480 ◽  
pp. 543-547
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Wan Yun Li ◽  
Shao Wen Chen ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Injection ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

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