scholarly journals Status and Assessment Method of Nuclear Safety Analysis Software in China

2021 ◽  
Vol 9 ◽  
Author(s):  
Xinli Gao ◽  
Jianping Jing ◽  
Xiangzhen Han ◽  
Bin Jia ◽  
Xinlu Tian ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Evaluation Process ◽  
Safety Analysis ◽  
Assessment Method ◽  
Nuclear Safety ◽  
Analysis Software ◽  
Software Safety ◽  
Engineering Software ◽  
Software Applications

In recent years, China’s nuclear power industry has enjoyed a good momentum of development, and related companies have also developed many nuclear analysis software applications. However, as the National Nuclear Safety Administration (NNSA, Chinese nuclear regulatory institution) did not approve any software before 2018, all these software applications were not evaluated formally, so they have not yet been used in reactor safety analysis. In order to solve this problem, in 2018, the National Nuclear Safety Administration started to carry out an engineering applicability evaluation for software developed by Chinese companies. After several years of review, as the first approved Chinese domestic software, core physics analysis software PCM developed by the China General Nuclear Power Group officially passed the software safety evaluation of the China Nuclear Safety Administration. This study will present the basic situation of the development of China’s nuclear power engineering software and introduce the framework, methods, procedures, requirements, and other aspects of China’s software safety evaluation work. The evaluation process and evaluation key issues of PCM software will also be illustrated.

HFE Safety Evaluation in Angra-1 NPP for the Preparation and Inclusion of a Chapter 18 in the Plant Final Safety Analysis Report

2014 ◽  
Author(s):  
Pedro Trueba Alonso ◽  
Juan Carlos Valdivia ◽  
Luís Fernández Illobre ◽  
Mark Hulsmans
Keyword(s):  
Nuclear Power ◽  
Action Plan ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Nuclear Safety ◽  
Human Factors Engineering ◽  
The European Union ◽  
Commercial Operation ◽  
The Government ◽  
Set Up

Angra-1 Nuclear Power Station (Westinghouse PWR-600 MW, 2 loops) started commercial operation in 1985, being property of Eletronuclear, subsidiary of Eletrobras in Brazil. Angra-1 has been preparing the necessary measures to renew the operating license and to apply for a lifetime extension up to 60 years. Among the many activities to perform, there are some related to fulfilling the requirements of the Brazilian regulator, the CNEN. These include requirements related to Human Factors Engineering (HFE) that included the preparation of a Chapter 18 of HFE, to become part of the plant’s Final Safety Analysis Report (FSAR). In the framework of the Instrument for Nuclear Safety Cooperation (INSC), created and funded by the European Union (EU) to enhance nuclear safety world-wide, cooperation activities between the EU and the Government of Brazil were set up in 2009. One of the INSC projects funded was to support the Brazilian nuclear operator of Angra-1 in the field of HFE. In 2010, the implementation of the project was awarded to a consortium lead by Tecnatom for performing a HFE Safety Evaluation to the plant and to provide support for preparing this Chapter 18. For this Project a specific methodology was developed for the execution of the Safety Evaluation. The methodology has been developed for evaluating — from the HFE viewpoint — a plant in operation, from the beginning of commercial operation until nowadays, including the design modifications performed to date. The obtained results have been used for developing the aforementioned Chapter 18. The main results of the Project Execution have been: 1. The developed methodology has made it possible to perform a comprehensive HFE evaluation of Angra-1, including the analysis of Post-TMI requirements, the design included in the current FSAR, the existing Angra-1 procedures and the verification of the current Main Control Room. 2. Technical support has been provided to Angra-1 for the preparation of Chapter 18 of the FSAR, following the structure of NUREG-0711, and using the results of the HFE Safety Evaluation. 3. An Action Plan has been developed for identifying and addressing in the future all those deficiencies found during the HFE Safety Evaluation, as well as those activities that are the consequence of the new FSAR Chapter 18.

Studies on Safety Assessment Method for Operating License Extension of Nuclear Power Plants in China

2017 ◽  
Author(s):  
Zhilin Chen ◽  
Ping Huang ◽  
Chunhui Wang ◽  
Zhiyuan Chi ◽  
Fangjie Shi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Assessment Method ◽  
Nuclear Safety ◽  
Management Systems ◽  
Effect Evaluation ◽  
Environment Effect ◽  
Assessment Environment

It’s the trend to extend the operating license time, called Operating License Extension (OLE) in China, of nuclear power plants (NPPs) in the future. It needs to be adequately demonstrated by licensees and approved by the regulator to gain an extended license time, such as 20 years. The demonstration methods for OLE are different among countries due to the different management systems for NPPs. Safety assessment, environment effect evaluation and update of the final safety analysis report (FSAR) will be the main aspects during OLE demonstration of NPPs in China according to the technical policy issued by National Nuclear Safety Administration (NNSA). Technical methods for scoping and screening, aging management review and time-limited aging analyses, which are the main contents of safety assessment are established based on the technical policy drafted by NNSA and international experiences in order to assist the operators to implement the safety assessment for OLE of NPP.

Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

2011 ◽  
Author(s):  
S. Herstead ◽  
M. de Vos ◽  
S. Cook
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Safety Analysis ◽  
Reactor Power ◽  
Nuclear Safety ◽  
Regulatory Framework ◽  
Nuclear Industry ◽  
Plant Design ◽  
Regulatory Documents

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

IVR Model Study for Passive PWR

2012 ◽  
Author(s):  
Hong Xu ◽  
Peng Zhang ◽  
Zhiwei Zhou
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Sensitivity Study ◽  
Nuclear Safety ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Plant Design ◽  
Layer Model ◽  
Pressurized Water ◽  
Exterior Surface

1000-MWe scale Pressurized Water Reactor (PWR) is taking service or under construction all over the world, and larger scale plant is studied and developed for its more competitive economics. Not only design basic accidents are analyzed for nuclear safety, the severe accident must also be considered to meet with the increasing requirement of safety. In the “nuclear power plant design safety regulation” (HAF102) issued by Nation Nuclear Safety Administration (NNSA), aim at the preventing and mitigating of severe accident, the regulation bring forward new requirement, which required that during design phase, NPP should consider setting the preventing and mitigation measurement of severe accident as actually as possible. As an approach to prevent the curium from melting down the vessel and entering the containment when a postulated severe accident occurs, In-vessel retention (IVR) of molten core debris via water cooling of the external surface of the reactor vessel has been introduced into AP1000. External reactor vessel cooling (ERVC) is assumed to be achieved keeping exterior surface of vessel at 400K. It is known to all that different scenario and process results in different IVR molten model. As the core melt, different IVR model is formed at different time, such as two-layer model, three-layer model and four layer model. It is necessary to study the IVR model when severe accident process moves on. This paper studies two-layer and three-layer IVR models and find the features of the models. Based on this, sensitivity study of important parameters has also been analyzed. It is useful for us to understand the mechanism of the molten pool. This paper has some directive significance on future IVR strategy research and also provides theoretical support to safety evaluation of PWR plants.

Development of Self-Reliant Thermal Hydraulic and Safety Analysis Software for NPP Design

2018 ◽  
Author(s):  
Yu Liu ◽  
Jian Deng ◽  
Junjie Pan ◽  
Zongjian Lu
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Core Competencies ◽  
Theoretical Models ◽  
Safety Analysis ◽  
Analysis Model ◽  
Analysis Software ◽  
Research Software ◽  
Model Research ◽  
Technological Level

For the thermal hydraulic and safety analysis of nuclear reactor, a lot of theoretical models and engineering experience are mainly contained in the design software. The degree of self-reliance of software directly reflects the technological level and core competencies. According to the nuclear power plant (NPP) design requirements, self-reliant thermal hydraulic and safety analysis software, such as CORTH, TRANTH, PHYCA and etc., have been developed by Nuclear Power Institute of China (NPIC) with reasonable planning and scientific implementation. In this paper, the development process of the self-reliant software is reviewed, covering requirement analysis, model research, software design, coding, testing, verification and validation. And the main characteristics of self-reliant software were summarized. The successful development of thermal hydraulic and safety analysis software support the export of nuclear power units of China, and enhance the competitiveness.

Preliminary Safety Analysis of CSR1000

2013 ◽  
Author(s):  
Pan Wu ◽  
Junli Gou ◽  
Jianqiang Shan ◽  
Bo Zhang ◽  
Xiang Li
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Safety System ◽  
Sensitivity Analyses ◽  
Passive Safety ◽  
Loss Of Coolant Accident ◽  
Passive Safety System ◽  
Loss Of Coolant ◽  
Break Area

This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.

NPP I&C Safety Assessment by Aggregation of Formal Techniques

2018 ◽  
Author(s):  
Eugene Babeshko ◽  
Vyacheslav Kharchenko ◽  
Kostiantyn Leontiiev ◽  
Oleg Odarushchenko ◽  
Oleksiy Strjuk
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Failure Modes ◽  
Assessment Method ◽  
Assessment Methods ◽  
Assessment Process ◽  
Software Safety ◽  
Safety Level ◽  
Normative Documents

Safety assessment of nuclear power plant instrumentation and control systems (NPP I&Cs) is a complicated and resource consuming process that is required be done so as to ensure the required safety level and comply to normative regulations. A lot of work have been performed in the field of application of different assessment methods and techniques, modifying them and using their combinations so as to provide unified approach in comprehensive safety assessment. Anyway, performed research have shown there are still challenges to overcome, including rationale and choice of the safety assessment method, verification of assessment results, choosing and applying techniques that support safety assessment process, especially in the nuclear field. In our work we present developed framework that aggregates the most appropriate safety assessment methods typically used for NPP I&Cs. Key features that this framework provides are the formal descriptions of all required input information for every safety assessment method, possible data flows between methods, possible output information for every method. Such representation allows to obtain possible paths required to get necessary indicators, analyze the possibility to verify them by application of different methods that provide same indicators etc. During safety assessment of NPP I&Cs it is very important to address software due to its crucial role in I&C safety assurance. Relevant standards like IEC 60880 [1] and IEC 62138 [2] provide requirements for software related activities and supporting processes in the software safety lifecycle of computer-based I&C systems of nuclear power plants performing functions of safety category A, B and C, as defined by IEC 61226 [3]. Requirements and frameworks provided by IEC 60880 and IEC 62138 for the nuclear application sector correspond to IEC 61508, part 3 [4]. These standards define several types of safety related software and specify particular requirements for each software type. So as to verify software and confirm correspondence to required safety level, different techniques are suggested in normative documents. We share our experience obtained during software failure modes and effect analysis (software FMEA) and software fault insertion (software FIT) processes into FPGA-based platform, NPP I&C systems based on that platform, and RPCT, integrated development environment used by RPC Radiy and end users to design user application logic, specify hardware configuration etc. We apply software FIT to outputs of RPCT, considering source code, configuration files and firmware files. Finally, we provide a case study of application the developed safety assessment framework and software FMEA/FIT practices during practical assessment of FPGA-based NPP I&C system.

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