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.

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.

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.

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.

Progress in Decommissioning of Ignalina NPP Unit 1

2002 ◽  
Author(s):  
Darius Ancius ◽  
Rimantas Krenevicius ◽  
Saulius Kutas ◽  
Michel Chouha
Keyword(s):  
Waste Management ◽  
Nuclear Power ◽  
Technical Assistance ◽  
Technical Support ◽  
Nuclear Safety ◽  
International Standards ◽  
Energy Strategy ◽  
Safety Level ◽  
The Government ◽  
Support Fund

The aim of the paper is to present the Lithuanian legal framework regarding the nuclear safety in Decommissioning and Waste Management, and the progress in the Decommissioning Programme of the unit 1 of Ignalina Nuclear Power Plant (INPP). INPP is the only nuclear plant in Lithuania. It comprises two RBMK-1500 reactors. After Lithuania has restored its independence, responsibility for Ignalina NPP was transferred to the Republic of Lithuania. To ensure the control of the Nuclear Safety in Lithuania, The State Nuclear Power Safety Inspectorate (VATESI) was created on 18 October 1991, by a resolution of the Lithuanian Government. Significant work has been performed over the last decade, aiming at upgrading the safety level of the Ignalina NPP with reference to the International standards. On 5 October 1999 the Seimas (Parliament) adopted the National Energy Strategy: • It has been decided that unit 1 of Ignalina NPP will be closed down before 2005, • The conditions and precise final date of the decommissioning of Unit 2 will be stated in the updated National Energy strategy in 2004. On 20–21 June 2000, the International Donors’ Conference for the Decommissioning of Ignalina NPP took place in Vilnius. More than 200 Millions Euro were pledged of which 165 M€ funded directly from the European Union’s budget, as financial support to the Decommissioning projects. The Decommissioning Program encompasses legal, organizational, financial and technical means including the social and economical impacts in the region of Ignalina. The Program is financed from International Support Fund, State budget, National Decommissioning Fund of Ignalina NPP and other funds. Decommissioning of Ignalina NPP is subject to VATESI license according to the Law on Nuclear Energy. The Government established the licensing procedure in the so-called “Procedure for licensing of Nuclear Activities”; and the document “General Requirements for Decommissioning of the Ignalina NPP” has been issued by VATESI. A very important issue is the technical support to VATESI and the Lithuanian TSO’s (Technical Support Organisations) in their activities within the licensing process related to the Decommissioning of INPP. This includes regulatory assistance in the preparation of decommissioning and radioactive waste management regulatory documents, and technical assistance in the review of the safety case presented by the operator. The Institute for Radioprotection and Nuclear Safety (IRSN, France) and the French Nuclear Safety Authority (DSIN) as well as Swedish International Project (SIP) are providing their support to VATESI in these areas.

Recent Developments of Corporate Governance in the European Union and their Impact on the German Legal System

2004 ◽  
Vol 5 (10) ◽  
pp. 1275-1294
Author(s):  
Athanasios Kouloridas ◽  
Jens von Lackum
Keyword(s):  
Corporate Governance ◽  
Action Plan ◽  
Company Law ◽  
Public Consultation ◽  
Public Confidence ◽  
The European Union ◽  
Legal Form ◽  
Recent Developments ◽  
Set Up ◽  
The Eu

The collapses of several US-businesses like those of Enron and Worldcom and a number of scandals in the EU – in the recent past that of Parmalat – have strongly affected public confidence in the operation and governance of large entities trading their shares in organized capital markets. The European Commission reacted by issuing the Action Plan on Modernizing Company Law and Enhancing Corporate Governance in the EU on 21 May 2003. The Action Plan contains measures which the Commission wants to implement over the short term (until 2005), medium term (until 2008) and long term (until 2010). The key issues set up in the Action Plan concern corporate governance, capital maintenance, recapitalization as well as decreasing capital, groups of companies, international corporate restructuring and the introduction of a new legal form of incorporation. The fact that the big rating agencies have begun to rate the corporate governance performances of major companies, can well be seen as a further indicator that good corporate governance has an important concern for managers, shareholders and for policy makers. As part of the Action Plan, the Commission has recently launched consultations on board responsibilities and improving financial and corporate governance information, on directors’ remuneration and on the role of (independent) non-executive or supervisory directors. In the light of these recent consultations and the results of the public consultation on the Action Plan, this Article offers an overview and assessment of the corporate governance measures planned at Community level.

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.

scholarly journals The concept of sustainable development in the mediterranean : emergence and recurrence

1999 ◽  
Vol 54 (2) ◽  
pp. 105-112 ◽  
Author(s):  
L. Chabason
Keyword(s):  
Sustainable Development ◽  
Action Plan ◽  
Ecological Impact ◽  
United Nations Environment Programme ◽  
Sustainable Urban Development ◽  
The European Union ◽  
First Results ◽  
The Mediterranean ◽  
Set Up ◽  
Barcelona Convention

Abstract. The history ofthe relationship between man and nature, since time immemorial, sets the scene for studying issues related to sustainable development. Concepts of «carrying capacity» and «ecological impact» are not new, as is illustrated by the example of the use of water resources in Ancient Greece. The Mediterranean region is particularly sensitive to such problems, and the protection of the Mediterranean sea was one of the first results to emerge from the Stockholm Conference in 1972. The United Nations Environment Programme (UNEP), born after the Conference, gave birth to the Mediterranean Action Plan, leading to the Barcelona Convention linking together twenty coastal states, as well as the European Union. In this framework, the Blue Plan was set up and designed to study the impacts on the environment of development and population growth. Several possible scenarios were set up, providing background material for the Earth Summit in Rio. The period between 1990 – 1995 saw the newly set up Mediterranean Commission dealing with issues relating to water (management, pollution), tourism (colonisation of natural sites, pressure on the environment) and sustainable management of coastal regions. Other points were also raised recently, such as sustainable urban development. However, this institutionalisation of environmental problems that happen at both the national and international levels, should not lead to a compromise Statement achieving nothing concrete. Indeed, the maintenance of environmental achievements and the definition of new concepts should allow sustainable development to move forward.

Safety Evaluation of the HTTR-IS Nuclear Hydrogen Production System

2010 ◽  
Author(s):  
Hiroyuki Sato ◽  
Hirofumi Ohashi ◽  
Yujiro Tazawa ◽  
Nariaki Sakaba ◽  
Yukio Tachibana
Keyword(s):  
Hydrogen Production ◽  
Production System ◽  
Evaluation Method ◽  
Cooling System ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Fuel Temperature ◽  
System Pressure ◽  
Production Of Hydrogen ◽  
The Government

Establishment of a safety evaluation method is one of the key issues for the nuclear hydrogen production demonstration since fundamental differences in the safety philosophy between nuclear plants and chemical plants exist. In the present study, a practical safety evaluation method, which enables to design, construct and operate hydrogen production plants under conventional chemical plant standards, is proposed. An event identification for the HTTR-IS nuclear hydrogen production system is conducted in order to select abnormal events which would change the scenario and quantitative results of the evaluation items from the existing HTTR safety evaluation. In addition, a safety analysis is performed for the identified events. The results of safety analysis for the indentified five Anticipated Operational Occurrences (AOOs) and three ACciDents (ACDs) show that evaluation items such as a primary cooling system pressure, temperatures of heat transfer tubes at pressure boundary, etc., do not exceed the acceptance criteria during the scenario. In addition, the increase of peak fuel temperature is small in the most severe case, and therefore the reactor core was not damaged and cooled sufficiently. These results will contribute to the safety review from the government and demonstration of the nuclear production of hydrogen.

DEVELOPMENT OF 222RN IN AIR SECONDARY STANDARD IN THE SLOVAK REPUBLIC

2020 ◽  
Vol 191 (2) ◽  
pp. 160-165
Author(s):  
Pavol Blahušiak ◽  
Matej Krivošík ◽  
Jarmila Slučiak ◽  
Andrej Javorník ◽  
Michaela Zálešáková ◽  
...  
Keyword(s):  
European Union ◽  
Slovak Republic ◽  
Action Plan ◽  
Secondary Standard ◽  
The European Union ◽  
Technical Parameters ◽  
Council Directive ◽  
Set Up ◽  
Definition Of

Abstract Slovak Institute of Metrology received in 2016 funding for realisation of a set up of a radon chamber with AlphaGUARD as a secondary standard of 222Rn in air as one part of the project. This secondary standard will serve to provide the traceability for laboratories that deal with 222Rn measurements in environmental samples. This project is a response to the European Union legislation and provides metrological support for the development and implementation of the national radon action plan, which the member states of the European Union are committed to fulfil in accordance with Council Directive 2013/59/Euratom. During development of the radon chamber, the determination of its basic technical parameters, such as the exact determination of the container volume and the area of the inner walls of the radon chamber, many tightness tests of the chamber, the definition of homogeneous radon atmosphere parameters and bilateral comparisons, were realised.

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.

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