ML20258A177

NUREG/IA-0519,Rev.1, Survey of Member Countries' Nuclear Power Plant Fire Protection Regulations by the Oecd Nuclear Energy Agency (NEA) Fire Incidents Records Exchange (Fire) Database Project - Topical Report No. 2

ML20258A177
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Issue date:	09/30/2020
From:	Thomas Aird, Nick Melly, Markhenry Salley Office of Nuclear Regulatory Research, Organization for Economic Co-operation and Development (OECD), Nuclear Energy Agency
To:
Malone, Tina
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NUREG/IA-0519, Rev. 1
Download: ML20258A177 (240)

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NUREG/IA-0519, Rev. 1 International Agreement Report Survey of Member Countries Nuclear Power Plant Fire Protection Regulations by the OECD Nuclear Energy Agency (NEA) Fire Incidents Records Exchange (FIRE) Database Project - Topical Report No. 2 Organisation for Economic Co-operation and Development (OECD)

Nuclear Energy Agency (NEA)

Committee on the Safety of Nuclear Installations (CSNI)

Paris, France Nicholas B. Melly, NRC FIRE Database Representative Thomas H. Aird, P.E., Editor Mark Henry Salley, P.E., Project Manager Division of Risk Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Manuscript Completed: September 2020 Date Published: September 2020 Published by U.S. Nuclear Regulatory Commission

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NUREG/IA-0519, Rev. 1 International Agreement Report Survey of Member Countries Nuclear Power Plant Fire Protection Regulations by the OECD Nuclear Energy Agency (NEA) Fire Incidents Records Exchange (FIRE) Database Project - Topical Report No. 2 Organisation for Economic Co-operation and Development (OECD)

Nuclear Energy Agency (NEA)

Committee on the Safety of Nuclear Installations (CSNI)

Paris, France Nicholas B. Melly, NRC FIRE Database Representative Thomas H. Aird, P.E., Editor Mark Henry Salley, P.E., Project Manager Division of Risk Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Manuscript Completed: September 2020 Date Published: September 2020 Published by U.S. Nuclear Regulatory Commission

ABSTRACT Nations establish and authorize agencies with the responsibility to protect the health and safety of the public and the environment by licensing and regulating nuclear power plants (NPP). Fires have been shown to be a major risk to NPP safety. Fire protection regulations built on defense-in-depth principles have been established in each country to minimize this risk. The purpose of this report is to collect and share the fire protection regulations and strategies used in different countries to ensure reactor safety. The report is built by each member country assembling their major fire protection regulations with all the regulations translated in the common language of English. In addition, an international trend exists for regulations to evolve from prescriptive requirements to risk-informed performance-based requirements. This report includes that information where applicable. The completed report now provides a single reference where countries can review and contrast their NPP fire protection regulations with other member countries. Through international cooperation in projects such as this research effort, each member country may discover new insights and ideas for their NPP fire protection regulations.

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FOREWORD This work was completed as a part of the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Fire Incidents Records Exchange Database Project. Each member country was responsible for assembling (and translating to English where necessary) their section on that countrys nuclear power plant fire protection regulations. The U.S. Nuclear Regulatory Commission (NRC) volunteered to complete the final preparation of this report for publication as a NUREG/International Agreement Report (IA).

The contents of this report should not be viewed as an official NRC endorsement or any other member countrys endorsement of the results or observations in this report. In addition, this report should not be viewed as binding the NRC or any member country in its rulemaking, licensing, or adjudicatory process.

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TABLE OF CONTENTS ABSTRACT ................................................................................................................... iii FOREWORD ................................................................................................................... v TABLE OF CONTENTS................................................................................................ vii LIST OF FIGURES......................................................................................................... xi LIST OF TABLES .......................................................................................................... xi EXECUTIVE

SUMMARY

............................................................................................. xiii ACKNOWLEDGMENTS ............................................................................................... xv ABBREVIATIONS AND ACRONYMS ........................................................................ xvii 1 INTRODUCTION ........................................................................................................ 1 1.1 Previous Topical Reports ..............................................................................................2 2 SCOPE AND OBJECTIVES ....................................................................................... 3 2.1 Description of the Work .................................................................................................3 2.1.1 Format/Structure................................................................................................3 2.1.2 Member Country Survey Questionnaire .............................................................3 3 FIRE PROTECTION AND POST-FIRE SAFE SHUTDOWN REGULATIONS IN MEMBER COUNTRIES ......................................................................................... 5 3.1 Belgium .........................................................................................................................5 3.1.1 Existing Reactors...............................................................................................5 3.1.2 New Reactors ....................................................................................................8 3.1.3 Supplementary Information ................................................................................8 3.2 Canada .......................................................................................................................10 3.2.1 Existing Reactors.............................................................................................13 3.2.2 New Reactors ..................................................................................................13 3.2.3 Supplementary Information ..............................................................................14 3.3 Czech Republic ...........................................................................................................15 3.3.1 Existing Reactors.............................................................................................15 3.3.2 New Reactors ..................................................................................................16 3.3.3 Supplementary Information ..............................................................................16 3.4 Finland ........................................................................................................................20 3.4.1 Existing Reactors.............................................................................................21 3.4.2 New Reactors ..................................................................................................22 3.4.3 Supplementary Information ..............................................................................24 3.5 France.........................................................................................................................26 3.5.1 Existing Reactors.............................................................................................26 3.5.2 New Reactors ..................................................................................................36 3.5.3 Supplementary Information ..............................................................................36 3.6 Germany .....................................................................................................................37 vii

3.6.1 Existing Reactors.............................................................................................37 3.6.2 New Reactors ..................................................................................................40 3.6.3 Supplementary Information ..............................................................................40 3.7 Japan ..........................................................................................................................46 3.7.1 Existing Reactors.............................................................................................46 3.7.2 New Reactors ..................................................................................................47 3.7.3 Supplementary Information ..............................................................................47 3.8 Korea ..........................................................................................................................51 3.8.1 Existing Reactors.............................................................................................51 3.8.2 New Reactors ..................................................................................................53 3.8.3 Supplementary Information ..............................................................................54 3.9 The Netherlands..........................................................................................................56 3.9.1 Existing Reactors.............................................................................................56 3.9.2 New Reactors ..................................................................................................57 3.9.3 Supplementary Information ..............................................................................57 3.10 Spain...........................................................................................................................59 3.10.1 Existing Reactors.............................................................................................59 3.10.2 New Reactors ..................................................................................................61 3.10.3 Supplementary Information ..............................................................................61 3.11 Sweden .......................................................................................................................64 3.11.1 Existing Reactors.............................................................................................64 3.11.2 New Reactors ..................................................................................................68 3.11.3 Supplementary Information ..............................................................................68 3.12 Switzerland .................................................................................................................72 3.12.1 Existing Reactors.............................................................................................72 3.12.2 New Reactors ..................................................................................................73 3.12.3 Supplementary Information ..............................................................................73 3.13 United Kingdom...........................................................................................................75 3.13.1 Existing Reactors.............................................................................................75 3.13.2 New Reactors ..................................................................................................79 3.13.3 Supplementary Information ..............................................................................80 3.14 United States of America.............................................................................................81 3.14.1 Existing Reactors.............................................................................................81 3.14.2 New Reactors ................................................................................................101 3.14.3 Supplementary Information ............................................................................ 102 4 REGULATORY QUESTIONAIRE........................................................................... 103 4.1 General Comparisons ...............................................................................................103 4.2 Regulation Questionnaire Form.................................................................................103 4.2.1 Questions about the use of FIRE PSA (PRA) in FIRE member countries ...... 103 4.2.2 High Level Questionnaire Summary Results .................................................. 103 4.3 Survey Responses ....................................................................................................104 4.3.1 Belgium .........................................................................................................104 4.3.2 Canada ..........................................................................................................105 4.3.3 Czech Republic .............................................................................................106 4.3.4 Finland ..........................................................................................................107 4.3.5 France ...........................................................................................................108 4.3.6 Germany........................................................................................................109 4.3.7 Japan ............................................................................................................110 4.3.8 Korea .............................................................................................................111 4.3.9 The Netherlands ............................................................................................111 viii

4.3.10 Spain .............................................................................................................112 4.3.11 Sweden .........................................................................................................112 4.3.12 Switzerland ....................................................................................................113 4.3.13 United Kingdom .............................................................................................114 4.3.14 United States of America ...............................................................................116 5 CONCLUSIONS AND RECOMMENDATIONS ...................................................... 121 5.1 General Conclusions .................................................................................................121 5.2 Recommendations to the FIRE Database Project ..................................................... 121 5.3 Recommendations to CSNI and CNRA ..................................................................... 122 6 REFERENCES ....................................................................................................... 123 APPENDIX A GERMAN NUCLEAR FIRE AND EXPLOSION STANDARDS KTA 2101.1-3 AND KTA 2103 .......................................................... A-1 ix

LIST OF FIGURES Figure 1 Key Elements of the CNSC's Regulatory Framework ............................................. 10 Figure 2 The Five Levels of Defense in Depth with Respect to Fire Protection..................... 12 Figure 3 Overview of the Regulatory Framework in Finland .................................................21 Figure 4 French Regulatory Architecture ..............................................................................26 Figure 5 Nuclear Regulatory Framework in Germany...........................................................37 Figure 6 Overview of the Regulatory Framework for Fire Protection in Korea ...................... 51 Figure 7 Hierarchical Pyramid of Nuclear Regulation in Spain ............................................. 62 Figure 8 How Conventional and Nuclear Specific Fire Related Requirements can be Combined to Result in a Facility Adapted Fire Protection ....................................... 66 Figure 9 Need for Fire Compartments and Evacuation Routes when all Equipment in a Fire Compartment are Assumed to Fail During Fire - Sweden ................................ 67 Figure 10 Need for Fire Compartments and Evacuation Routes when Analytical Fire Hazard Analysis can Demonstrate that Protection Measures are Sufficient to Prevent the Failure of Redundant Items Important to Safety - Sweden .................. 68 Figure 11 Regulatory Framework for Fire Protection in Switzerland ....................................... 72 LIST OF TABLES Table 1 Fire Resistance Class Requirement Related to Fire Load - Sweden ........................ 68 Table 2 Summary of the U.S. Operating NPPs Fire Protection Regulations .......................... 85 Table 3 Summary of U.S. Regulations for Pre-1979, Post-1979, and NFPA 805 Plants ........ 86 Table 4 Comparison of U.S. NRC Fire Protection Program Features .................................... 86 Table 5 U.S. 10 CFR 50 Appendix R Fire Damage ...............................................................91 Table 6 FIRE Member Countries - High Level Questionnaire Summary Results ................. 103 xi

EXECUTIVE

SUMMARY

Operating experience from nuclear installations worldwide has shown that fires can occur throughout the NPPs lifetime, and they can pose a significant impact on plant operation. Fires that have occurred in nuclear installations worldwide demonstrate that fires in certain locations of a NPP may impair the required functions of structures, systems and components important to safety, and in particular, redundant safety systems.

Fires in NPPs have the potential to induce initiating events (e.g., plant transients) that cause failures of equipment necessary to mitigate them, and to adversely affect, directly or indirectly, the barriers for prevention of the release of radioactive materials. Fires can also simultaneously challenge more than one level of defense-in-depth.

Many countries employ various defense-in-depth approaches for ensuring nuclear safety and that the plant can be safely shut down in the event of a fire including:

• Minimizing the occurrence of fires and explosions.

• Minimizing combustible loading.

• Maximizing the plant ability to detect, control, and extinguish the fires that do occur.

• Ensuring that plant operations have redundancy and diversity to enable a safe shutdown of the reactor despite having a fire.

• Minimizing the risk of releases to the environment.

Two primary types of methodologies for assessing fire safety at nuclear power plants include:

• Deterministic (or prescriptive) requirements, which dictate the level of protection to ensure reactor shutdown systems will survive an assumed serious fire, and

• Risk-informed and performance-based requirements, which considers risk insights as well as other factors to better focus attention and resources on design and operational issues to match the levels of protection according to their importance to safety.

Many member countries ensure that the requirements of the country specific regulation are met, and the fire protection codes and standards are followed by implementing a fire protection program for all operational stages of the NPPs. The fire protection program demonstrates that fire protection measures are implemented in a controlled, coordinated and effective manner, in order to minimize both the probability of occurrence and the consequences of fire.

This Topical Report presents the overall structure related to fire safety regulations for the fourteen FIRE member countries and depicts the structure of that countrys regulatory process.

It provides the baseline information and references to detailed, country specific documentation materials.

In addition to the regulatory information, each member country has also answered a survey related to the implementation of risk-informed, performance-based fire protection in the national regulations and associated challenges for investigating on a country-to-country basis if, and in particular how, probabilistic versus deterministic requirements are implemented.

One general conclusion from the survey questionnaire is that several countries are using a combination of prescriptive and risk-informed approaches to fulfil their fire safety regulatory xiii

framework. The general trend among member countries is moving towards using Fire PSA 1 as 0F a tool for gaining risk insights regardless of regulatory framework guidelines. This trend supports the conclusions from several investigations on the level of maturity for Fire PSA, which more recently (cf. [NEA-19]) is one of the areas of interest of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) Working Group on Risk Assessment (WGRISK).

1 In this report the abbreviations PRA (Probabilistic Risk Assessment) and PSA (Probabilistic Safety Assessment) are used synonymously.

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ACKNOWLEDGMENTS This activity was performed as a part of the Fire Incidents Records Exchange Database Project.

This report would not be possible without the individual contributions from the member countries. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Secretariat wishes to thank the experts listed below, who provided valuable time and considerable knowledge towards the development of this NUREG/International Agreement Report:

Charles Fourneau (Belgium) Abderrazzaq Bounagui Frantisek Stvan (Canada) (Czech Republic)

Matti Lehto (Finland) Jean-Pierre Cayla (France) Heinz-Peter Berg (Germany)

Alexandra Iancu (Germany) Marina Rwekamp (Germany) Hajime Kabashima (Japan)

Sang Kyu Lee (Korea) Jong-Seuk Park (Korea) Laima Kuriene (The Netherlands)

Eunate Armananzas Albaizar Miguel Angel Jimenez Garcia Anna Hggstrm (Sweden)

(Spain) (Spain)

Christian Karlsson (Sweden) Ralph Nyman (Sweden) Dominik Hermann (Switzerland)

Annette Ramezanian Simon Thompson J. S. Hyslop (United States)

(Switzerland) (United Kingdom)

Nicholas Melly (United States)

The authors would like to thank David Stroup from the NRC Office of Nuclear Regulatory Research (RES), Naeem Iqbal, Charles Moulton, and Jennifer Whitman from NRCs Office of Nuclear Regulatory Regulation (NRR) for their suggestions, comments, and review on an early draft of this report.

The authors would like to thank the OECD/NEA administrators, Markus Beilmann, Neil Blundell, and Olli Nevander for their exceptional support and oversight on this project.

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ABBREVIATIONS AND ACRONYMS AEC Atomic Energy Commission (United States)

AGR Advanced Gas Cooled Reactor AHJ authority having jurisdiction ALARA(P) as low as reasonably achievable (practicable)

ANS American Nuclear Society ANVS Autoriteit Nucleaire Veiligheid en Stralingsbescherming (The Netherlands)

APCSB Auxiliary and Power Conversion Systems Branch ASME American Society of Mechanical Engineers ASN Autorité de Sûreté Nucléaire (France)

Bel V Belgian TSO, part of the FANC BEST Belgian Stress Test BfE Bundesamt für kerntechnische Entsorgungssicherheit (Germany); Federal Office for Nuclear Waste Mangement BfS Bundesamt für Strahlenschutz (Germany); Federal Office for Radiation Protection BMU Bundesministerium für Umwelt, Naturschutz und nukleare Sicherheit (Germany); Federal Ministry for the Environment, Nature Protection and Nuclear Safety BTP Branch Technical Position CCR code compliance review CDF core damage frequency CNRA Committee on Nuclear Regulatory Activities (NEA)

CNSC Canadian Nuclear Safety Commission CP construction permit CRIEPI Central Research Institute of the Electric Power Industry (Japan)

CSA Canadian Standard Association CSN Consejo de Seguridad Nuclear (Spain)

CSNI Committee on the Safety of Nuclear Installations (NEA)

DBA design basis accident DEC design extension conditions DSR Dutch Safety Requirements EA environmental assessment xvii

ENSI Eidgenssisches Nuklearsicherheitsinspektorat (Switzerland)

EPRI Electric Power Research Institute EOP emergency operating procedures FANC Federal Agency for Nuclear Control FDF fuel damage frequency FHA fire hazard analysis FIRE Fire Incidents Records Exchange (NEA)

FPP fire protection program FSSA fire safe shutdown analysis GDA generic design assessment GRS Gesellschaft für Anlagen- und Reaktorsicherheit (Germany)

HEAF high energy arcing fault HSK Hauptabteilung für die Sicherheit der Kernanlagen (Switzerland)

HSWA Health and Safety at Work Act HTGR High-Temperature Gas-cooled Reactor I&C instrument and control IAEA International Atomic Energy Agency IAR International Agreement Report IRR Ionizing Radiation Regulations IRSN Institut de Radioprotection et de Sûreté Nucléaire (France)

KAERI Korea Atomic Energy Research Institute KFD Kernfysische Dienst (The Netherlands)

KINS Korea Institute of Nuclear Safety KTA Kerntechnischer Ausschuss (Germany)

LAR license amendment request LC License Condition LERF large early release frequency LOCA loss of coolant accident LOOP loss of offsite power MCR main control room MSO multiple spurious operation NBCC National Building Code of Canada xviii

NSCA Nuclear Safety and Control Act NEA Nuclear Energy Agency NEI Nuclear Energy Institute NFCC National Fire Code of Canada NFPA National Fire Protection Association NIA Nuclear Installations Act NPP nuclear power plant NRA Nuclear Regulation Authority (Japan)

NRC Nuclear Regulatory Commission (USA)

NRR Office of Nuclear Reactor Regulation (NRC USA)

NRRC Nuclear Risk Research Center (Japan)

NSSC Nuclear Safety and Security Commission (Korea)

NSCA Nuclear Safety and Control Act OECD Organisation for Economic Co-operation and Development OL operating license ONR Office for Nuclear Regulation (United Kingdom)

PIE postulated initiating event PORV power-operated relief valve PRA Probabilistic Risk Assessment PRG project review group PROC power reactor operating license (Canada)

PSA Probabilistic Safety Assessment PSR periodic safety review PWR Pressurized Water Reactor RB regulatory body RD royal decree REPPIR Radiation (Emergency Preparedness and Public Information) Regulations 2001 RES Office of Nuclear Regulatory Research (NRC USA)

RGP relevant good practice RHWG Reactor Harmonization Working Group SAP Safety Assessment Principle SFAIRP so far as is reasonably practicable SONS State Office for Nuclear Safety (SÚJB in Czech) xix

SSC structures, systems and components SSM Strlskerhetsmyndigheten (Swedish Radiation Safety Authority)

STUK Radiation and Nuclear Safety Authority (Finland)

TAG Technical Assessment Guide TEA The Energy Act TSO Technical Safety Organisation TUKES Finnish Safety and Chemicals Agency (Finland)

TÜV TÜVRheinland (Germany)

UJV Ústav Jaderného Výzkumu (Czech Republic)

VNDS Vienna Declaration on Nuclear Safety WANO World Association of Nuclear Operators WENRA Western European Nuclear Regulators Association WGRISK Working Group on Risk Assessment (NEA)

WGWD Working Group on Waste and Decontamination (NEA) xx

1 INTRODUCTION The purpose of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Fire Incidents Records Exchange (FIRE) Database Project is to provide a platform for multiple countries to collaborate and exchange fire data and thereby to enhance the understanding of how fire phenomena affects the safe operation of nuclear power plants (NPP) and, in turn, to improve the quality of quantitative fire risk assessment requiring fire-related data.

In that context, the Project aims to:

• Collect fire event experience (by international exchange) in an appropriate format in a quality-assured and consistent database (the OECD FIRE Database).

• Collect and analyze fire events over the long-term so as to better understand such events and their causes and to encourage their prevention.

• Generate qualitative insights into the root causes of fire events to derive approaches or mechanisms for their prevention and to mitigate their consequences.

• Establish a mechanism for efficient operation feedback on fire event experience including the development of policies of prevention such as indicators for risk-informed and performance-based inspections.

• Record characteristics of fire events to facilitate fire risk analysis including quantification of fire occurrence frequencies.

The Topical Reports are developed by members of the OECD/NEA participating in the FIRE Database Project (14 countries). The selections of the topic and of the participant members who are going to undertake the task are agreed upon during the FIRE Database Project Review Group (PRG) meetings.

The following is a list of topics where work has either commenced, or which are identified for future analysis:

• Fire event apparent causes and root cause analyses: This ongoing activity has a specific interest in the ability to derive approaches or mechanisms for future fire prevention and mitigation of potential consequences.

• Risk significant contributions in probabilistic safety assessment (PSA) for fires: This work involves an analysis of the available Fire PSAs to identify the significant elements in the PSA, either compartment wise or component wise. These would be analysed and investigated for consistency within the database;

• Multiple unit/fire area impacts: This effort is related to an earlier Topical Report on Event Combinations of Fire and Other Events [NEA-16] with a specific focus on potential effects for fire affecting multiple units or fire areas and the unique attributes for plant safety.

• Fire brigade effectiveness and implementation strategies: This topic includes an investigation of country-specific approaches to fire brigade and fire suppression techniques.

This Topical Report supports the database project by compiling the regulations of the member countries related to fire safety in NPPs. This compilation does not attempt to evaluate the approaches and merits of regulatory strategies in one country versus the other members but rather to provide a platform for understanding fire regulations across the member countries. This Topical Report evaluates regulatory trends as well as attempts to discuss the processes and 1

methods each country undertakes at a high level to meet specific regulatory goals and requirements related to fire safety.

1.1 Previous Topical Reports Two Topical Reports have already been published: The first one issued in 2013 [NEA-13] has provided results of the analysis of fires resulting from high energy arcing faults (HEAF). This Topical Report has resulted in an additional experimental research program by OECD/NEA, which included full-scale HEAF testing and analysis. The first test program results were published in 2017 [NEA-17]; a second phase, the international HEAF Project (HEAF, Phase 2),

which covers further test series, has been started recently. Another Topical Report of the FIRE Database Project has presented results of analyses with respect to event combinations of fires and other anticipated events [NEA-16]. This report provided valuable insights on the operating experience feedback concerning the potential impact and consequences of fire event combinations. The insights have supported recent activities of the WGRISK regarding risk aggregation for hazards (including fires) for Site-Level PSA.

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2 SCOPE AND OBJECTIVES This report addresses the regulatory framework and structures of the member countries involved in the Fire Incidents Records Exchange (FIRE) Database Project. One of the primary aspects addressed in this report is the methods in which member countries address the risks associated with fire protection. Generally, member countries historically choose between two different approaches for assessing fire safety in nuclear power plants (NPPs): deterministic fire safety measures versus probabilistic fire risk assessment with the latter applied more for regulatory decision making. Although a deterministic assessment is performed to ensure that specific requirements for prescribed elements of defense-in-depth are met, the elements of the defense-in-depth concept can also provide a framework for the analysis within a Fire probabilistic safety assessment (PSA). Some countries have regulations containing a mix of these two approaches.

One benefit of the international FIRE Database Project is to provide the member countries a venue to discuss NPP fire-related experience in their countries. This Topical Report is an attempt to assemble and consolidate the fire protection and post-fire safe shutdown regulations of each FIRE Database Project member country. This will enable experts from member countries to better understand how their fire-specific regulations compare with those from other member countries.

The regulatory information reflected in this report is currently up to January 1, 2018, unless otherwise noted in the member countries regulatory section. The scope of this document covers commercial NPPs unless otherwise noted in the member countries regulatory section.

Due to differences between countries, the information was collected in a manner to facilitate a general understanding and to draw parallels between regulatory processes. The information is presented in the following format.

2.1 Description of the Work 2.1.1 Format/Structure

a. Existing Reactors - General overview of the fire-safety-specific regulations currently in place for existing reactors either in commercial operation or in post-commercial safe shutdown or under decommissioning.

b. New Reactors - Some member countries are currently in the process of building new nuclear facilities. An attempt was made to collect information as to the regulatory process for new build reactors to reveal if the regulations differed or have been updated to account for new information or designs.

c. Supplementary Information - Member countries were encouraged to provide supplementary information concerning the regulatory framework and structure within their regulatory bodies.

2.1.2 Member Country Survey Questionnaire The purpose of the following survey was to generalize member countries fire-safety-related regulatory structures to provide feedback for a general comparison.

Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

Q.2 Are Fire Probabilistic Risk Assessments (PRAs) used to support license applications?

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Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

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3 FIRE PROTECTION AND POST-FIRE SAFE SHUTDOWN REGULATIONS IN MEMBER COUNTRIES The following section presents the regulatory framework with respect to fire protection and post-fire safe shutdown in each of the fourteen Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Fire Incidents Records Exchange (FIRE) member countries and the corresponding nuclear and applicable non-nuclear fire protection regulations. This covers existing operational reactors as well as reactors under construction. In principle, all countries consider the general requirements and high-level guidance documents related to fire protection by the International Atomic Energy Agency (IAEA), specifically NS-G2.1

[IAEA-00b] and NS-G-1.7 [IAEA-04]. Soon a revised guide [IAEA-19] will replace NS-G-1.7 merging it with NS-G-1.11 [IAEA-04a]. However, a lot of more detailed guidance is currently available on a national basis for most of the FIRE member countries.

3.1 Belgium The information contained herein was accurate as of December 2018 and applies to major nuclear installations (defined as Class I in the Belgian regulation) in Belgium, including commercial nuclear power generating reactors.

3.1.1 Existing Reactors 3.1.1.1 Overview of the Regulatory Framework The current form of the Belgian nuclear regulatory framework has been established by the Royal Decree of 20 July 2001 that enforces many articles of the Law of 15 April 1994 and made the Federal Agency for Nuclear Control (FANC), created by that Law, operational. This agency, which is endowed with wide competences, constitutes the Safety Authority.

In 2007, the FANC created Bel V as a subsidiary with the statute of a so-called foundation as defined in Belgian law. Bel V is given a mandate to perform regulatory missions that can be delegated by the FANC, in line with the provision of the above law, including on-site routine inspections and technical support. The association of the FANC and Bel V is referred to as the Regulatory Body.

The main Belgian regulations regarding the nuclear safety are:

• The Law of 15 April 1994 on the protection of the population and the environment against the hazards of ionizing radiation and on the Federal Agency for Nuclear Control, amended for the last time in 2017 [BEL-01];

• Royal Decree of 20 July 2001, enforcing the above Law and laying down the General Regulation for the protection of the public, workers and the environment against the hazards of ionizing radiation (RGPRI/ARBIS), amended for the last time in 2018 [BEL-02];

• Royal Decree of 01 March 2018 establishing the nuclear and radiological emergency plan for the Belgian territory [BEL-03];

• Royal Decree of 30 November 2011 on the safety requirements for nuclear installations, transposing the WENRA (Western European Nuclear Regulators Association) Reference Levels of 2008 [WENRA-08] in the Belgian legislation, amended for the last time in 2015 [BEL-04].

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3.1.1.2 Fire Protection Regulation For historical reasons, the Belgian nuclear power sector followed the American rules for the design and construction of its nuclear power plants, i.e. the requirements of the Code of Federal Regulations (10 CFR 50) [NRC-04, NRC-01, NRC-02] and the documents issued by the U.S.

NRC such as the Regulatory Guides, the Standard Review Plans, the NUREGs, as well as the application of the ASME code, or the ANS/IEEE standards, etc. As a consequence, National Fire Protection Association (NFPA) standards were largely adopted for the original design of the fire protection systems in the nuclear power plants currently in operation.

Nevertheless, the Belgian rules and regulations, including the transposition in the Belgian law of the relevant European directives have now to be followed. Bel V inspects and assesses the application of national regulations, and consequently national or European standards, wherever applicable.

Belgian nuclear facilities have to adhere to both conventional and nuclear specific fire protection regulations.

The main conventional fire protection regulations are:

• Royal Decree of 7 July 1994 [BEL-01] and subsequent amendments, defining the Base Standards for fire protection in buildings, other than individual housings. More specifically the Appendix 6 addresses the topic of fire protection in industrial buildings.

Minimal requirements for compartmentalization and fire rating of structures, as well as automatic detection and suppression are covered in this document.

• Article 52 of the Belgian regulation for the protection of the workers (RGPT/ARAB) also known as the code for the welfare at work. This article covered multiple organizational aspects of the fire protection, among which the prevention measures, the evacuation planning, the fire-fighting measures, the periodical tests and control or the training of the workers. It was recognized for decades as the reference for planning and organization of fire protection in Belgium but was largely superseded by the Royal Decree of 28 March 2014 [BEL-07]. Some parts remain applicable.

• Articles 104 to 110 of the Belgian regulation for electrical installations (RGIE/AREI).

These cover the fire resistance rating of electrical equipment and most notably electrical cables, along with the anti-deflagration properties in relation with the European ATEX directives.

• Royal Decree of 28 March 2014 f, now integrated in the Code of well-being at work (Code du bien-tre au travail) [BEL-07], or fire protection at workplaces is the most recent development in the field of fire protection in Belgium. It updates and complements the requirements of the article 52 of the Belgian regulation for the protection of the workers. It also introduces the necessity to perform a risk analysis to demonstrate that fire prevention and protection measures are sufficient to ensure the safety of the workers.

For nuclear facilities, additional requirements are enforced through the Royal Decree of 30 November 2011 [BEL-04]. As presented above, this Royal Decree transposes the WENRA Safety Reference Levels of 2008 in the Belgian legislation.

Article 17 of this so-called WENRA Royal Decree covers the protection against internal fires for all Class I nuclear facilities and is a slightly amended transposition of the Issue S of the WENRA 2008 RLs. This includes the production of a Fire Hazard Analysis, using a deterministic approach.

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17.1 Strategy for protection against internal fires; 17.2 Basic design principles; 17.3 Fire hazard analysis; 17.4 Fire protection systems; 17.5 Administrative controls and maintenance; 17.6 Firefighting organization.

The deterministic fire risk analysis should address the defense-in-depth principles applied to fire protection by systematically evaluating fire scenarios, using conservative assumptions, in all rooms containing, or related to rooms containing, equipment identified as important to safety. All plant operating modes must be covered and/or enveloped by the scenarios taken into consideration and the combinations of fire events with postulated independent initiating events such as long-term management of a large break loss of coolant accident (LOCA), loss of offsite power (LOOP) or safe shutdown state after an earthquake are studied.

The objectives of this analysis are met if the following safety functions (as well as all support systems to these functions) can be guaranteed at all time under the above conditions:

1. Removal of residual heat

2. Safe shutdown capabilities

3. Confinement

4. Long-term internal accident mitigation Because Article 17 was established to be applicable to all Class I nuclear facilities, specific provisions for the protection against internal fires in nuclear power reactors were added in Article 32, including the production of a Fire Probabilistic Safety Analysis, or Fire PSA:

32.1 Basic design principles The ability to perform the reactor shutdown, decay heat removal, radioactive products confinement and plant state monitoring must be maintained during and after the fire event; 32.2 Fire Risk Analysis The fire hazard analysis shall be complemented by probabilistic fire analysis. In PSA level 1, the fires shall be assessed in order to evaluate the fire protection arrangements and to identify risks caused by fires; 32.3 Fire Protection Systems The distribution loop for fire hydrants outside building and the internal standpipes shall provide adequate coverage of areas of the plant relevant to safety. The coverage shall be justified by the fire hazard analysis.

A new regulatory project was started by the FANC in order to translate the updated WENRA Safety Reference Levels of September 2014 [WENRA-14a] into the Belgian regulations. The regulatory project will amend the existing royal decree on the safety of nuclear installations (30 November 2011 [BEL-04) and is in final revision stage for publication in 2019. This regulatory evolution will especially result in stronger requirements for the analysis of events combinations.

In addition to the regulatory evolution, the BEST (Belgian Stress Test) project has led to all licensees re-evaluating the emergency preparedness and response plans as aftermath to the Fukushima disaster. Among the outcomes of these tests, additional intervention and emergency back-up means, including specific fire-fighting equipment for the on-site and off-site fire brigades, have been provisioned.

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3.1.2 New Reactors In 2015 the FANC published a new guidance on the Safety demonstration of new class I nuclear installations -Approach to Defense-in-Depth, radiological safety objectives and application of a graded approach to external hazards, updated in April 2017 [BEL-05]. This guideline is written in relation to the council directive 2014/87/Euratom of 8 July 2014 [EUR-14] amending directive 2009/71/Euratom [EUR-09]. In addition, to support the requirements already formulated in the Belgian regulation, this guideline is mainly inspired from recent WENRA publications on the safe design of new NPPs, from safety reference levels, as well as from recent IAEA publications.

The chapter specific to nuclear power reactors of the Belgian regulation [BEL-04] specifies that the selected design basis events will be grouped into a limited number of categories according to their probability of occurrence and requires the definition of acceptance criteria for each category such that there are no or minor radiological consequences for frequent events and that events with potential severe consequences must have a very low probability of occurrence.

These general terms can be generalized to new class I nuclear installations. Design basis categories are addressed in this guideline, in relation to Defense-in-Depth and radiological safety objectives to be associated to these categories are discussed.

In addition to the above regulation, all new Class I nuclear installation (including nuclear reactors) will have to adhere to these guidelines for the safety demonstration, and in particular for internal hazard fire.

3.1.3 Supplementary Information

[BEL-01] Law of 15 April 1994 on the protection of the population and the environment against the hazards of ionizing radiation and on the Federal Agency for Nuclear Control, amended for the last time in 2017, http://www.jurion.fanc.fgov.be/jurdb-consult/consultatieLink?wettekstId=2182&appLang=fr&wettekstLang=fr.

[BEL-02] Royal Decree of 20 July 2001, enforcing the above Law and laying down the General Regulation for the protection of the public, workers and the environment against the hazards of ionizing radiation (RGPRI/ARBIS), amended for the last time in 2018, http://www.jurion.fanc.fgov.be/jurdb-consult/consultatieLink?wettekstId=7460&appLang=fr&wettekstLang=fr.

[BEL-03] Royal Decree of 01 March 2018 establishing the nuclear and radiological emergency plan for the Belgian territory, http://www.jurion.fanc.fgov.be/jurdb-consult/consultatieLink?wettekstId=26393&appLang=fr&wettekstLang=fr.

[BEL-04] Royal Decree of 30 November 2011 on the safety requirements for nuclear installations, transposing the WENRA (Western European Nuclear Regulators Association) Reference Levels of 2008 in the Belgian legislation, amended for the last time in 2015, http://www.jurion.fanc.fgov.be/jurdb-consult/consultatieLink?wettekstId=15152&appLang=fr&wettekstLang=fr.

[BEL-05] Guideline on the Safety demonstration of new class I nuclear installations -

Approach to Defence-in-Depth, radiological safety objectives and application of a graded approach to external hazards, https://afcn.fgov.be/fr/system/files/guideline-safety-demonstration-new-classi-installations-rev1-final.pdf.

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[BEL-06] Guidelines for new Class I nuclear installation (including external hazards aircraft crash, earthquake and flooding),

https://afcn.fgov.be/fr/dossiers-dinformation/autres-etablissements-nucleaires/directives-pour-une-nouvelle-installation.

[BEL-07] Code du Bien-tre au Travail, Book III, Title 3, Fire prevention in the workplace, http://www.emploi.belgique.be/moduleDefault.aspx?id=1958.

[EUR-09] Council Directive 2009/71/Euratom of 25 June 2009 establishing a Community framework for the nuclear safety of nuclear installations, https://eur-lex.europa.eu/eli/dir/2009/71/oj.

[EUR-14] Council Directive 2014/87/Euratom of 8 July 2014 amending Directive 2009/71/

Euratom establishing a Community framework for the nuclear safety of nuclear installations, http://data.europa.eu/eli/dir/2014/87/oj.

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3.2 Canada The Canadian Nuclear Safety Commission (CNSC) is the federal organization responsible for regulating the use of nuclear energy and materials in Canada. It regulates to protect health, safety, security and the environment, and to implement Canada's international commitments on the peaceful use of nuclear energy. The CNSC also disseminates objective scientific, technical, and regulatory information to the public.

The CNSC regulates the conduct of activities related to the use, production and distribution of nuclear energy and substances as defined by section 26 of the Nuclear Safety and Control Act (NSCA) [CAN-01]. This includes activities related to:

- Uranium mines and mills;

- Uranium fuel fabrication and processing;

- Nuclear power plants;

- Nuclear substance processing;

- Industrial and medical applications:

- Nuclear research and educational activities;

- Transportation of nuclear substances;

- Nuclear security and safeguards;

- Import and export activities;

- Waste management facilities; The CNSCs regulatory framework (see Figure 1) consists of the NSCA and laws passed by Parliament that govern the regulation of Canada's nuclear industry, as well as regulations, licenses and documents that the CNSC uses to regulate the industry.

Figure 1 Key Elements of the CNSC's Regulatory Framework 10

The regulatory framework also includes guidance, which is used to inform the applicant or licensees on how to meet requirements, elaborate further on requirements, or provide best practices. While the CNSC sets requirements and provides guidance on how to meet requirements, an applicant or licensee may put forward a case to demonstrate that the intent of a requirement is addressed by other means. Such a case must be demonstrated with supportable evidence. CNSC staff consider guidance when evaluating the adequacy of any case submitted. This does not mean that the requirement is waived; rather, it is an indication that the regulatory framework provides flexibility for licensees to propose alternative means of achieving the intent of the requirement. The Commission is always the final authority as to whether the requirement has been met.

CNSC requirements and guidance take into account international regulatory best practices and modern codes and standards and align with the International Atomic Energy Agencys Safety Fundamentals and Safety Requirements. The CNSC cooperates with other organizations and jurisdictions to foster the development and application of a consistent, effective regulatory framework in Canada and for international nuclear regulators. The CNSC welcomes stakeholder feedback on its regulatory framework at any time.

Further information on the CNSCs regulatory framework can be found on the CNSCs regulatory framework overview Web page.

The regulatory requirements for fire risk mitigation are determined based upon:

1. Class of the facility pursuant to the regulations;

2. Risk to persons and the environment (graded requirements e.g. license conditions);

and

3. Achieving the regulatory fire protection goals.

To meet the requirements of the NSCA and associated regulations, the regulatory fire projection goals are:

1. Health and safety of persons;

2. Protection of the environment;

3. Nuclear substances safety;

4. Nuclear criticality safety;

5. Reactor safety.

The CNSCs regulatory model in fire protection is based upon the implementation of the defense-in-depth concept (REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants

[CNSC-14]) to ensure the protection of the health and safety of persons and the environment.

From a fire protection perspective, defense in depth is achieved through a combination of design (e.g., physical barriers, spatial separation, fire protection detection and suppression systems), management of fire protection (e.g., operational procedures), quality assurance and emergency arrangements. The defense-in-depth applies to fire protection at all levels of the facility and its associated activities, from establishing high-level facility objectives to defining the detailed procedures and equipment required to meet those objectives.

To achieve a high level of confidence that the fire protection goals will be met, an appropriate level of defense-in-depth should be maintained throughout the lifetime of the facility, through the fulfilment of the five elements of the defense-in-depth principles (Figure 2).

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Level 5: Mitigate radiological consequences of significant releases of radioactive substances pth De in Level 4: Control severe nuclear facility conditions and mitigate the consequences of severe accidents fe nc e De Level 3 : Minimize the consequences of fires Level 2: Detect rapidly, control and promptly extinguish fires that do occur Level 1: Prevent Fires from starting Figure 2 The Five Levels of Defense in Depth with Respect to Fire Protection This is achieved by the implementation of the following fire protection standards:

• CSA N293, Fire protection for CANDU nuclear power plants [CSA-12] required for nuclear power plants;

• CSA N393, Fire protection for facilities that process, handle or store nuclear substances

[CSA-13] required for others nuclear facilities.

These standards require the implementation of the National Building Code of Canada (NBCC)

[NRCC-15], and the National Fire Code of Canada (NFCC) [NRCC-15a] (similar to NFPA 5000

[NFPA 18], 101 [NFPA 18a] & 1 [NFPA 18b]). The CSA standards are prescriptive in nature but allow for the use of alternative solutions to meet prescriptive requirements. The NBCC and the NFCC are objective based Codes and contain prescriptive requirements but state the objectives and functional statements for each prescriptive requirement and allow alternative solutions to meet these requirements.

The CSA N293 and CSA N393 standards require that:

• Licensees implement and maintain a comprehensive fire protection program (FPP) in order to reduce the occurrence of fires and limit their consequences and severity (the required elements of the program are prescribed in the CSA standards). The FPP is defined as a set of planned, coordinated and controlled activities which is documented and integrated into the operation of the facility; and

• Complete fire safety assessments include Fire hazard assessment (FHA), Fire safe shutdown analysis (FSSA), and Code Compliance review. The FHA and FSSA are deterministic analyses.

A Level 2 (Fire) PSA is a requirement of NPP licenses (through the license requirement for the preparation of a PSA in accordance with REGDOC 2.4.2, Probabilistic Safety Assessment (PSA) for Nuclear Power Plants [CNSC-14a] (but is not currently required to be incorporated into to the Fire Protection Program. REGDOC 2.4.2 requires licensee to see CNSC acceptance of the methodology and computer codes to be used for the PSA before using them. The methodology typically used for the completion of Fire PSAs is NUREG/CR-6850 [NRC-08].

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3.2.1 Existing Reactors Existing NPPs are required to comply with CSA N293 [CSA-12]. As CSA N293 was written after the construction of all the Canadian NPPs, the design and construction requirements of this standard are not retroactively applied to existing structures, systems and components.

However, these requirements apply for modification to the plant. The operational requirements of the standards do apply.

3.2.1.1 Refurbishment and Life Extension As part of its licensing process, the CNSC requires Periodic Safety Reviews (PSRs), a technical assessment to be completed before authorizing a NPP refurbishment or life extension project through REGDOC-2.3.3 [CNSC-15]. A PSR involves an assessment of the current state of the plant and its performance to determine the extent to which it conforms to applicable modern codes, standards and practices, and to identify any factors that would limit safe long-term operation.

The PSR includes a systematic review of the all safety systems of the reactor to identify possible safety improvements to enhance safety and minimize environmental impacts. The PSR identifies all practicable safety improvements that could be made during the refurbishment.

The PSR provides for a rigorous review of the reactor's systems, structures and components against modern codes and standards, experience, best practices and research findings. The PSR considers a wide range of safety related topics (including plant design, environmental qualification, and probabilistic safety analysis).

For existing facilities preparing for refurbishment, a code compliance review (CCR) is completed against modern codes and standards. That is, the facility and operations are reviewed against the current edition of CSA N293, the NBCC and NFCC and the codes and standards referenced therein. The gaps identified in the CCR are either rectified during the refurbishment process, after the refurbishment (based on an approved corrective action plan) or are dispositioned (i.e.

justified as requiring no action) using a documented performance-based approach.

3.2.2 New Reactors The CNSC REGDOC 2.5.2 [CNSC-14] requires that suitable incorporation of operational procedures, redundant SSCs, physical barriers, spatial separation, fire protection systems, and design for fail-safe operation achieves the following general objectives:

• prevent the initiation of fires;

• limit the propagation and effects of fires that do occur by quickly detecting and suppressing fires to limit damage and confining the spread of fires and fire by-products that have not been extinguished;

• prevent loss of redundancy in safety and safety support systems;

• provide assurance of safe shutdown;

• ensure that monitoring of safety-critical parameters remains available, and

• prevent exposure, uncontrolled release, or unacceptable dispersion of hazardous substances, nuclear material, or radioactive material, due to fires;

• prevent the detrimental effects of event mitigation efforts, both inside and outside of containment; and

• ensure structural sufficiency and stability in the event of fire 13

To achieve the noted objectives, the REGDOC 2.5.2 requires new NPPs to comply with CSA N293, NBCC and NFCC. In addition, the REGDOC recommends the following as guidance documents: U.S. NRC, NUREG-1852 Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire, 2007 [NRC-35], and Nuclear Energy Institute, NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, Washington, D.C., 2005 [NEI-05].

3.2.3 Supplementary Information

[CAN-01] Nuclear Safety and Control Act (NSCA), paragraph 3(a), subparagraph 9(a)(i),

paragraph 24(4)(b).

https://laws-lois.justice.gc.ca/PDF/N-28.3.pdf.

[CSA-12] Canadian Standard Association (CSA): Fire protection for nuclear power plants, CSA N293-12 (R 2017), Toronto, ONT, Canada, 2012.

[CSA-13] Canadian Standard Association (CSA): Fire protection for facilities that process, handle or store nuclear substances, CSA N393-13 (R 2018), Toronto, ONT, Canada, 2013.

[CNSC-14] Canadian Nuclear Safety Commission (CNSC): REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants, Ottawa, ONT, Canada, May 2014, http://www.nuclearsafety.gc.ca/pubs_catalogue/uploads/REGDOC-2-5-2-Design-of-Reactor-Facilities-Nuclear-Power-Plants-eng.pdf.

[CNSC-14a] Canadian Nuclear Safety Commission (CNSC): Probabilistic Safety Assessment (PSA) for Nuclear Power Plants, REGDOC-2.4.2, Ottawa, ONT, Canada, May 2014, http://www.nuclearsafety.gc.ca/pubs_catalogue/uploads/REGDOC-2-4 Probabilistic-Safety-Assessment-NPP-eng.pdf.

[CNSC-15] Canadian Nuclear Safety Commission (CNSC): Operating Performance Periodic Safety Review, REGDOC-2.3.3, Ottawa, ONT, Canada, April 2015, http://www.nuclearsafety.gc.ca/pubs_catalogue/uploads/REGDOC-2-3 Periodic-Safety-Reviews-eng.pdf.

[NRCC-15] National Research Council Canada: National Building Code of Canada (NBCC) 2015, Toronto, ONT, Canada, 2015, https://nrc.canada.ca/en/certifications-evaluations-standards/codes-canada/codes-canada-publications.

[NRCC-15a] National Research Council Canada: National Fire Code of Canada (NFCC) 2015, Toronto, ONT, Canada, 2015, https://nrc.canada.ca/en/certifications-evaluations-standards/codes-canada/codes-canada-publications.

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3.3 Czech Republic 3.3.1 Existing Reactors Fire Specific Regulations in Czech NPPs is described in document Protection against internal fires - Safety instructions JB-3.1 [SONS-10], which was issued by the State Office for Nuclear Safety (Czech Regulatory Authority) in 2010.

Safety instructions JB-3.1 is part of a series of safety guidelines that elaborate requirements defined by the WENRA Reactor Safety Reference Levels, Issue S [WENRA-08], the Waste and Spent Fuel Safety Reference Levels Report [WENRA-14] and also apply the recommendations of the International Atomic Energy Agency (from [IAEA-94], [IAEA-95], [IAEA-96], [IAEA-97],

[IAEA-98], [IAEA-98a], [IAEA-00], [IAEA-00a], [IAEA-00b], [IAEA-02], [IAEA-03], [IAEA-04],

[IAEA-04a], [IAEA-04b], [IAEA-06].

Instruction JB-3.1 focuses primarily on nuclear facilities which include nuclear reactors with power above 50 MWth, thus covering civilian nuclear power under the Convention on Nuclear Safety [IAEA-94]. Its principles and practices can be applied also to other kind of nuclear facilities defined by the Convention on the Safety Spent Fuel Management and the Safety of Radioactive Waste Management [IAEA-97] or research reactors.

It also addresses the influence of firefighting on nuclear safety. Personnel safety and property protection shall be ensured in accordance with valid legislation for fire protection.

The instruction JB-3.1 consists of eight main chapters covering the following topics:

• Chapter 3 - Background, objectives and relevance

- (3.2) Requirements for fire protection of nuclear facilities are defined by the implementing regulations of the Atomic Law - Decree no. 195/1999 [CZE-01]. The decree states in paragraph 9 the default requirements for fire protection of nuclear facilities and clarifies the requirements of legislation in the field of fire protection of nuclear facilities.

- (3.3) The basic law in the field of fire protection is Act No. 133/1985 [CZE-02]

about Fire Protection as amended, and the implementing regulations issued on its basis (Especially Decree No. 246/2001 [CZE-03] about Fire prevention and Decree No. 23/2008 [CZE-04] about Technical conditions of buildings fire protection). Fire protection requirements are further refined by technical standards, which contain requirements for fire protection and fire safety.

- (3.4) Instruction JB-3.1 is designed especially for the licensee to operate nuclear facilities and offers possible processes for activities in the field of fire protection in accordance with the requirements of the Act No. 18/1997 [CZE-05] (so called Atomic law, from year 2016 replaced by new Act No. 263/2016 [CZE-06]), its implementing regulations and with relevant WENRA reference levels [WENRA-08],

[WENRA-14a].

- (3.8) Instruction JB-3.1 addresses the issue of internal risks associated with the possibility of fire and explosion caused by a fire during the operation of nuclear installation and also establishes requirements for fire detection and firefighting systems.

• Chapter 4 - Overall concepts; Combination of events External fires Explosion protection

• Chapter 5 - Building design; Fire Hazard Analysis 15

• Chapter 6 - Fire prevention

• Chapter 7 - Detection and firefighting; smoke and heat exhaust

• Chapter 8 - Limitation of secondary impact of fires

• Annex 2 to instructions JB-3.1 - Fire Hazard Analysis - General requirements for FHA (10.2) FHA reflects the depth of information that corresponds to the preparation stage of construction, respectively the operation of nuclear installations. The scope of input data changes and the timeliness of the methodology used for assessment will be assessed regularly (once a year). The update of the document or its part (new revision) shall be carried out flexibly with regard to the rate of input data changes, or changes of the assessment methodology. A new revision will be issued immediately after the completion of the electronic version. The printed document will be updated with regard to the scope of the changes. The edition of the printed document may be required by SONS (Czech regulatory authority).

(10.4) The documentation of FHA, its creation, revision and updating, has to be checked and approved in accordance with the QA principles of the licensee [CZE-07]

(QA principles are currently based on Decree No. 408/2016 [CZE-08] and Decree No.

358/2016 [CZE-09]). The current conditions shall be described for each fire zone throughout the nuclear facility. FHA shall be maintained as a "living document" throughout the operation of nuclear installations.

(10.5) In processing of FHA is recommended to respect the concept and the breakdown corresponding to IAEA documentation.

(10.6) Deterministic part of FHA should be following up on Level 1 PSA.

3.3.2 New Reactors The acts and standards related to the nuclear regulation and the new fire protection guidelines and guides mentioned above are applicable for the existing reactors as well as for new reactors.

The paragraphs with relationship to new installations are primarily: 5.9, 5.21, 5.22, 6.1, 6.11, 6.30, 10.2, 10.4, 10.5 (see above) and naturally many others.

3.3.3 Supplementary Information Annex 1 to instructions JB-3.1 - The comparison with WENRA Reactor Safety Reference Levels, Area S WENRA Reactor Safety Reference Levels, Issue S Referenced in chapter(s) of Guide JB-3.1

1. Fire safety objectives 4-8 1.1 The licensee shall implement the defence-in-depth principle to fire protection, providing measures to prevent fires from starting, to detect and extinguish quickly any fires that do start and to prevent the spread of fires and their effects in or to any area that may affect safety.

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2. Basic design principles 5.8 - 5.20 2.1 SSCs important to safety shall be designed and located so as to minimize the frequency and the effects of fire and to maintain capability for shutdown, residual heat removal, confinement of radioactive material and monitoring of plant state during and after a fire event.

2.2 Buildings that contain equipment that is important to safety shall be designed as fire resistant, subdivided into compartments that segregate such items from fire loads and segregate redundant safety systems from each other. When a fire compartment approach is not practicable, fire cells shall be used, providing a balance between passive and active means, as justified by fire hazard analysis.

2.3 Buildings that contain radioactive materials that could cause 8.13 - 8.15 radioactive releases in case of fire shall be designed to minimize such releases.

2.4 Access and escape routes for fire-fighting and operating personnel 8.4 shall be available.

3. Fire hazard analysis 5.21 - 5.28, 10, Annex No. 2 3.1 A fire hazard analysis shall be carried out and kept updated to demonstrate that the fire safety objectives are met, that the fire design principles are satisfied, that the fire protection measures are appropriately designed and that any necessary administrative provisions are properly identified.

3.2 The fire hazard analysis shall be developed on a deterministic basis, covering at least:

- For all normal operating and shutdown states, a single fire and consequential spread, anywhere that there is fixed or transient combustible material;

- Consideration of credible combination of fire and other postulated initiating events (PIEs) likely to occur independently of a fire.

3.3 The fire hazard analysis shall demonstrate how the possible consequential effects of fire and extinguishing systems operation have been taken into account.

3.4 The fire hazard analysis shall be complemented by probabilistic fire analysis. In PSA level 1, the fires shall be assessed in order to evaluate the fire protection arrangements and to identify risks caused by fires.

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4. Fire protection systems 7.1 - 7.41 4.1 Each fire compartment or fire cell shall be equipped with fire 7.1 - 7.17 detection and alarm features, with detailed annunciation for the control room staff of the location of a fire. These features shall be provided with non-interruptible emergency power supplies and appropriate fire resistant supply cables.

4.2 Fixed or mobile, automated or manual extinguishing systems shall 7.18 - 7.28 be installed. They shall be designed and located so that their rupture, spurious or inadvertent operation does not significantly impair the capability of SSCs important to safety to carry out their safety functions.

4.3 The distribution loop for fire hydrants outside building and the 7.32 - 7.51 internal standpipes shall provide adequate coverage of areas of the plant relevant to safety. The coverage shall be justified by the fire hazard analysis.

4.4 Ventilation systems shall be arranged such that each fire 7.53 - 7.56, 8.1 compartment fully fulfils its segregation purpose in case of fire. - 8.12 4.5 Parts of ventilation systems (such as connecting ducts, fan rooms and filters) that are located outside fire compartments shall have the same fire resistance as the compartment or be capable of isolation from it by appropriately rated fire dampers.

5. Administrative controls and maintenance 4.10 - 4.14 5.1 In order to prevent fires, procedures shall be established to control and minimize the amount of combustible materials and minimize the potential ignition sources that may affect items important to safety. In order to ensure the operability of the fire protection measures, procedures shall be established and implemented. They shall include inspection, maintenance and testing of fire barriers, fire detection and extinguishing systems.

6. Firefighting organization 3.1 - 4.9 6.1 The licensee shall implement adequate arrangements for controlling and ensuring fire safety, as identified by the fire hazard analysis 6.2 Written emergency procedures that clearly define the responsibility and actions of staff in responding to any fire in the plant shall be established and kept up to date. A fire-fighting strategy shall be developed, kept up-to date, and trained for, to cover each area 6.3 When reliance for manual fire-fighting capability is placed on an offsite resource, there shall be proper coordination between the plant personnel and the off-site response group, in order to ensure that the latter is familiar with the hazards of the plant.

6.4 If plant personnel are required to be involved in fire-fighting, their organization, minimum staffing level, equipment, fitness requirements, and training shall be documented, and their adequacy shall be confirmed by a competent person.

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[CZE-01] Decree No. 195/1999, Requirements for nuclear installations to assure nuclear safety, radiation protection and emergency preparedness, SÚJB, 1999, https://www.sujb.cz/fileadmin/sujb/docs/legislativa/vyhlasky/R195_99.pdf.

[CZE-02] Act No. 133/1985, Fire protection, 1985, http://www.zakony.cz/zakon-SB1985133.

[CZE-03] Decree of the Ministry of Interior No. 246/2001, "The determination of fire safety conditions and state fire supervision (Decree on fire prevention), 2001, http://www.zakony.cz/zakon-SB2001246.

[CZE-04] Decree of the Ministry of Interior No. 23/2008, Technical conditions of fire protection of buildings, 2008, http://www.zakony.cz/zakon-SB2008023.

[CZE-05] Law No. 18/1997, Peaceful utilization of nuclear energy and ionizing radiation, SÚJB, 1997, http://www.zakony.cz/zakon-SB1997018.

[CZE-06] Law No. 263/2016, Atomic Law, SÚJB, 2016, http://www.zakony.cz/zakon-SB2016263.

[CZE-07] Decree no. 132/2008, Quality Assurance System for conducting activities related to the use of nuclear energy and radiation and quality assurance of classified equipment with respect to their safety classification, SÚJB 2008, http://www.zakony.cz/zakon-SB2008132.

[CZE-08] Decree No. 408/2016, The management system requirements, SÚJB, 2016, http://www.zakony.cz/zakon-SB2016408.

[CZE-09] Decree No. 358/2016 Requirements for quality assurance and technical security and compliance auditing and assessment of classified equipment, SÚJB 2016, http://www.zakony.cz/zakon-SB2016358.

[SONS-10] State Office for Nuclear Safety (SONS): Protection against Internal Fires - Safety Instructions JB-3.1, Prague, Czech Republic, 2010, https://www.sujb.cz/fileadmin/sujb/docs/dokumenty/publikace/Ochrana_proti_vnitr nim_pozarum_BN_JB_3.1.pdf.

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3.4 Finland The regulatory information provided by Finland is up to August 1, 2019.

The task of the Radiation and Nuclear Safety Authority (STUK) as the national authority responsible for oversight of the safety of the use of nuclear energy is based on the Nuclear Energy Act (990/1987) [FIN-01] and the Nuclear Energy Decree (161/1988) [FIN-02]. According to Section 7 r of the Nuclear Energy Act, STUK shall specify detailed safety requirements for the implementation of the safety level in accordance with the Nuclear Energy Act. STUK's general oversight procedures in regulating nuclear facilities are given in Guide YVL A.1. STUKs oversight includes the oversight of the fire protection arrangements of nuclear facilities in so far as they affect the nuclear and radiation safety of the facilities. STUKs fire protection requirements of nuclear facilities are given in Guide YVL B.8 [FIN-19].

The Regulation STUK Y/1/2018 [FIN-03] presents requirements for the safety design of nuclear power plants:

• Section 9 requires implementation of the defense-in-depth principles to prevent accidents and to mitigate their consequences;

• Section 15 requires that the internal hazards to be considered include at least fire, flood, explosion, electromagnetic radiation, pipe breaks, container ruptures, drop of heavy objects, missiles due to explosions or component failures, and other possible internal hazards;

• Section 16 presents requirements for the nuclear power plants control room arrangements;

• Sections 18-23 present requirements for the nuclear power plants construction, commissioning, operation, processing of operational experiences, safety research and the Operational Limits and Conditions; and

• Section 25 presents requirements for the organisation and personnel of a nuclear power plant.

The Regulation STUK Y/4/2018 [FIN-04] presents requirements for the safety design of the final disposal of nuclear waste: To prevent operational occurrences and accidents, Section 18 requires, among other things, that in a nuclear waste facility, the placement and protection of systems alongside operative methods shall ensure that fire, explosion or other internal hazards do not pose a threat to safety; Sections 22-24 present requirements for the construction, commissioning and operation of a nuclear facility.

The Ministry of the Environment issues technical regulations and guidelines on construction and structural fire protection [FIN-07]. The building inspection authority in each municipality sees to it that the regulations and guidelines issued by the Ministry are complied with in all construction activities.

Leadership and control of fire and rescue services, as well as the availability and quality of its services, rests with the Ministry of the Interior; the Ministry is also responsible for the preparation and arrangement of fire and rescue services at national level; and for co-ordination of the performance of different ministries involved in the fire and rescue services under the Rescue Act (379/2011) [FIN-08] and the Government Decree (407/2011) on fire and rescue services [FIN-09]. Regional State Administrative Agencies are responsible for the duties of rescue services in their sphere of activity. Municipalities are responsible in co-operation for fire and rescue services in a region determined by the Government (regional fire and rescue services). As regards the requirements, design, installation, maintenance, inspection 20

and demonstration of conformity of the equipment of the rescue services, the Rescue Equipment Act (10/2007) [FIN-10] shall be observed.

The Government Decree (917/1996) [FIN-11] and the Ministry of Trade and Industry Decision (918/1996) [FIN-12] present the requirements for equipment and protective systems intended for potentially explosive atmospheres. The Government Decree (576/2003) [FIN-13] presents the requirements for prevention of personnel hazards caused by potentially explosive atmospheres. The Finnish Safety and Chemicals Agency (TUKES) and the Ministry of Social Affairs and Health provide guidelines on the application of the ATEX legislation in Finland

[FIN-14].

STUKs activities do not affect any oversight activities required in the Land Use and Building Act (132/1999) [FIN-05], the Land Use and Building Decree (895/1999) [FIN-06], the Rescue Act (379/2011) [FIN-08] and the Government Decree (407/2011) on Rescue Services [FIN-09], unless otherwise agreed between the authorities. The regulatory framework in Finland is shown in Figure 3.

Nuclear Energy Act Nuclear Energy Decree STUK Regulations STUK YVL Guides Codes and Standards Figure 3 Overview of the Regulatory Framework in Finland 3.4.1 Existing Reactors STUK Guide YVL B.8, Fire Protection at a Nuclear Facility [FIN-19] was published in November 2013. The publication of a new or revised YVL Guide shall not, as such, alter any previous decisions made by STUK. After having heard the parties concerned STUK will issue a separate decision as to how a new or revised YVL Guide is to be applied to operating nuclear facilities or those under construction, and to licensees operational activities.

When considering how the new safety requirements presented in the YVL Guides shall be applied to the operating nuclear facilities, or to those under construction, STUK will take due account of the principles laid down in Section 7a of the Nuclear Energy Act (990/1987): The safety of nuclear energy use shall be maintained at as high a level as practically possible.

For the further development of safety, measures shall be implemented that can be considered 21

justified considering operating experience, safety research and advances in science and technology.

Under Section 7 r(3) of the Nuclear Energy Act, the safety requirements of the Radiation and Nuclear Safety Authority (STUK) are binding on the licensee, while preserving the licensee's right to propose an alternative procedure or solution to that provided for in the regulations. If the licensee can convincingly demonstrate that the proposed procedure or solution will implement safety standards in accordance with this Act, the Radiation and Nuclear Safety Authority (STUK) may approve a procedure or solution by which the safety level set forth is achieved.

3.4.2 New Reactors STUK Guide YVL B.8, Fire Protection at a Nuclear Facility [FIN-19] was published in November 2013. The Guide shall apply as it stands to new nuclear facilities. The content of STUK Guide YVL B.8 is handled below.

3.4.2.1 Scope of Application When this Guide sets requirements for nuclear facilities, reference is made, under the Nuclear Energy Act (990/1987), to facilities necessary for producing nuclear energy (nuclear power plants), including research reactors, facilities performing extensive final disposal of nuclear wastes, and facilities used for extensive fabrication, production, use, handling, storage of nuclear materials or nuclear wastes. Requirements for nuclear facilities always apply to nuclear power plants unless a requirement separately says they only apply to other nuclear facilities.

This Guide applies to the planning and implementation of fire protection during the design, construction and operation of the nuclear facility. The Guide shall be applied to the decommissioning of nuclear facilities. This guide shall be complied with at the entire plant area and in all its buildings.

As regards fire protection at a nuclear facility construction site, this guide shall apply whenever fire protection is significant for the safety of nearby nuclear facilities and to ensure fulfilment of the design criteria of the nuclear facility under construction.

In addition to the fire protection requirements of this Guide, the following Guides also contain fire protection related requirements to be followed:

- Guide YVL A.1, Regulatory oversight of safety in the use of nuclear energy, sets forth requirements for nuclear facility design and oversight.

- Guide YVL A.3, Management system for a nuclear facility, sets forth detailed requirements related to the management system and quality management.

- Guide YVL A.5, Construction and commissioning of a nuclear facility, sets forth requirements for the management and oversight of the construction project at different stages of a nuclear facility's construction.

- Guide YVL A.6, Conduct of operations at a nuclear power plant, sets forth requirements for the operation of a nuclear power plant, such as for outages.

- Guide YVL A.7, Probabilistic risk assessment and risk management of a nuclear power plant [FIN-20], sets forth requirements for probabilistic fire risk assessments.

- Guide YVL A.11, Security of a nuclear facility, sets forth requirements for physical protection at a nuclear facility and its planning.

- Guide YVL B.1, Safety design of a nuclear power plant [FIN-17], sets forth requirements for the nuclear power plants safety design and the design of systems 22

important to safety.

- Guide YVL B.7, Provisions for internal and external hazards at a nuclear facility [FIN-18], sets forth requirements for nuclear facility layout design and the design to protect against internal and external threats.

- Guide YVL E.6, Buildings and structures of a nuclear facility, sets forth requirements for the design of civil structures.

- Guide YVL E.7, Electrical and I&C equipment of a nuclear facility, sets forth electrical equipment-specific requirements for protection against fire load-induced explosions.

3.4.2.2 General Design Requirements Under Section 15 of the Regulation STUK Y/1/2018, structures, systems and components important to safety of a nuclear power plant shall be designed and located as well as protected in a way to make the likelihood of internal events (such as fires) small and their effect on facility safety insignificant.

A basis for the quality management of the nuclear power plants construction and operation is provided in Section 25 of the Regulation STUK Y/1/2018 on the safety of nuclear power plants.

It stipulates that organizations participating in the design, construction, operation and decommissioning of a nuclear power plant shall employ a management system for ensuring the management of nuclear safety, radiation safety and quality.

The fire protection for the nuclear facility shall be so planned that during and after a potential fire situation the nuclear facility can be brought to a safe state and the release of radioactive substances into the environment can be prevented.

The licensee can propose that also foreign regulations and guides be applied in designing the nuclear facilitys fire protection arrangements. It shall then be demonstrated, however, that they form a feasible entity. The application of foreign regulations and guides is subject to STUKs approval.

An organization carrying out the fire protection design of buildings shall have an SFS-EN ISO 9001 compliant management system that has been documented and implemented for this purpose.

For the inclusion of all aspects of fire protection, an expert responsible for fire protection design shall be nominated for the duration of the nuclear facilitys design and construction. The expert shall have sufficient qualifications and experience in nuclear, radiation and fire safety.

Management of the entirety of the nuclear facilitys fire protection arrangements places specific requirements on the combination of several design areas, such as facility layout, structural, heating/ventilation/air-conditioning, as well as electrical and I&C design.

In addition to the design requirements of this Guide, to complied with in the design of nuclear facilities are:

- the fire and building legislation in force in Finland.

- for applicable parts, the practices of risk-informed fire protection planning for nuclear power plants described in the IAEA Guides [IAEA-00b], [IAEA-14], [IAEA-15], [IAEA-16],

[IAEA-16a], [IAEA-16b], [IAEA-17], as well as in a Technical Report [IAEA-98a].

- the practices of the WENRA reference requirement area S, protection against internal fires [WENRA-08].

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3.4.2.3 Detailed Design Requirements Detailed requirements are given on the following items:

- Defense-in-depth Approach to Fire Protection;

- Fire Hazard Analysis;

- Structural Fire Protection;

- Active Fire Protection;

- Emergency Lighting;

- Provision for Outages / Annual Maintenance;

- Fire Safety During Facility Operation.

3.4.3 Supplementary Information

[FIN-01] Nuclear Energy Act (990/1987), Finland, 1987, http://www.finlex.fi/en/laki/kaannokset/1987/en19870990.pdf.

[FIN-02] Nuclear Energy Decree (161/1988), Finland, 1988, http://www.finlex.fi/en/laki/kaannokset/1988/en19880161.pdf.

[FIN-03] Radiation and Nuclear Safety Authority STUK: Regulation on the Safety of a Nuclear Power Plant, STUK-Y/1/2018, Helsinki, Finland, 2018 http://www.finlex.fi/data/normit/42574-STUK-Y-1-2018.en.pdf.

[FIN-04] Radiation and Nuclear Safety Authority STUK: Regulation on the Safety of Disposal of Nuclear Waste, STUK-Y/1/2018, Helsinki, Finland, 2018, http://www.finlex.fi/data/normit/42578-STUK-Y-4-2018.en.pdf.

[FIN-05] Land Use and Building Act (132/1999), Finland, 1999, http://www.finlex.fi/en/laki/kaannokset/1999/en19990132.pdf.

[FIN-06] Land Use and Building Decree (895/1999), Finland, 1999 ,

http://www.finlex.fi/en/laki/kaannokset/1999/en19990895.pdf.

[FIN-07] Ministry of the Environment, Decree on the fire safety of buildings (848/2017),

Finland, 2017, http://www.ym.fi/download/noname/%7B1A60A60B-75F6-4834-A746-767838898A8C%7D/139918.

[FIN-08] Rescue Act (379/2011), Finland, 2011, http://www.finlex.fi/en/laki/kaannokset/2011/en20110379.pdf.

[FIN-09] Government Decree on Rescue Services (407/2011), Finland, 2011, in Finnish only.

[FIN-10] Rescue Equipment Act (10/2007), Finland, 2007, in Finnish only.

[FIN-11] Government Decree on Equipment and Protection Systems Intended for Use in Potentially Explosive Atmospheres (917/1996), Finland, 1996, in Finnish only.

[FIN-12] Decision of the Ministry of Trade and Industry on Equipment and Protective Systems Intended for Use in Potentially Explosive Atmospheres (918/1996),

Finland, 1996, in Finnish only.

[FIN-13] Government Decree on the Prevention of Danger for Workers Caused by Explosive Atmospheres (576/2003), Finland, 2003, in Finnish only.

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[FIN-14] Ministry of Social Affairs and Health, Department for Occupational Safety and Health, Finnish Safety and Chemicals Agency (Tukes): ATEX - Safety of Explosive Spaces, 2003, in Finnish only.

[FIN-15] Ministry of the Interior, Directive for Rescue Diving (48/2007), SM050:00/2006, in Finnish only.

[FIN-16] Ministry of the Interior: Decree on Automatic Fire Extinguishing Equipment (SM-1999-967/Tu-33), Finland, 1999, in Finnish only.

[FIN-17] Radiation and Nuclear Safety Authority STUK: Safety design of a nuclear power plant, STUK Guide YVL B.1, Helsinki, Finland, June 2019, http://www.finlex.fi/data/normit/41774-YVL_B.1e.pdf.

[FIN-18] Radiation and Nuclear Safety Authority STUK: Provisions for Internal and External Hazards at a Nuclear Facility, STUK Guide YVL B.7, Helsinki, Finland, November 2013, http://www.finlex.fi/data/normit/41791-YVL_B.7e.pdf.

[FIN-19] Radiation and Nuclear Safety Authority STUK: Fire Protection at a Nuclear Facility, STUK Guide YVL B.8, Helsinki, Finland, November 2013, http://www.finlex.fi/data/normit/41792-YVL_B.8e.pdf.

[FIN-20] Radiation and Nuclear Safety Authority STUK: Probabilistic Risk Assessment and Risk Management of a Nuclear Power Plant, STUK Guide YVL A.7, Helsinki, Finland, February 2019, http://www.finlex.fi/data/normit/41813-YVL_A.7e.pdf.

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3.5 France 3.5.1 Existing Reactors The new French regulation was initiated by the law of 13 June 2006 on transparency and security in the nuclear field, called "law TSN [FRA-01]. It renovates in depth the legislative framework for nuclear activities and their control. It creates a Nuclear Safety Authority (ASN), an independent administrative authority responsible for supervising Nuclear Safety and Radiation Protection and public information in these areas.

The order laying down general rules for Basic Nuclear Installations of 7 February 2012, said order INB, was published in the Official Journal on 8 February 2012 [FRA-02]. It is a major element of this approach. This includes in particular the French law rules corresponding to international best practices. The provisions of the order INB are mainly dealing in the organization and responsibilities of the Basic Nuclear Installation operators, demonstration nuclear safety, nuisance control and their impact on health and the environment, waste management and the preparation and management of emergency situations. The scheme in Figure 4 below shows the new regulatory architecture.

Figure 4 French Regulatory Architecture The ASN resolution No 2014-DC-0417 of 28th January 2014 concerning the rules applicable to BNI with regard to management of fire risks [FRA-03] is applicable since 1st July 2014, except six articles dealing with new technical requirements which are applicable since 1st January 2017.

This resolution is divided into four titles. The first concerns the general provisions. It defines some terms and describes the objectives of the risk management process related to the fire that the operator must implement. It states that it is the operator who identifies, on the basis of the 26

safety demonstration, the important elements for protection to be protected from the effects of fire and arrangements to ensure this protection. The three other titles detail the arrangements to be taken according to the different levels defined by the concept of defense-in-depth. Title II is devoted to the prevention of fire starts. Title III focuses on the detection and intervention against fire, and title IV deals with provisions to prevent the spread of fire and limit its consequences.

One of the principles of this resolution is the obligation of results and not of means. It sets targets without specifying the means to achieve it.

French ASN resolution 2014-DC-0417 ASN resolution 2014-DC-0417 of 28th January 2014 concerning the rules applicable to basic nuclear installations (BNI) with regard to the management of fire risks ASN (Autorité de Sûreté Nucléaire - French Nuclear Safety Authority),

Having regard to the Environment Code, particularly title IX of book V; Having regard to the Labour Code; Having regard to decree 2007-1557 of 2nd November 2007, amended, relative to BNIs and to the regulation of the transport of radioactive substances in terms of nuclear safety, and its articles 3, 20, 37 and 43 in particular; Having regard to the order of 21st July 1994, amended, constituting the classification and certification of the conformity of the fire performance of electrical conductors and cables and approval of the test laboratories; Having regard to the order of 7th February 2012, amended, setting out the general rules relative to basic nuclear installations, more specifically its articles 3.5 and 3.6; Having regard to the results of the public consultations carried out on the ASN website from 27th December 2012 to 28th February 2013 and from 9th to 30th September 2013; Having regard to the opinion of the French High Council for technological risk prevention, dated 17th January 2013; Whereas a fire in a BNI can have significant safety consequences; Whereas the order of 31st December 1999, amended, setting the general technical regulations intended to prevent and mitigate off-site detrimental effects and risks resulting from the operation of BNIs, more specifically its title VI-B, comprised detailed regulatory provisions concerning fire risks; Whereas the above-mentioned order of 7th February 2012, which replaced the previous regulatory orders concerning BNIs, more specifically the above-mentioned order of 31st December 1999, requires that fire risks be taken into account, while leaving it up to ASN regulatory resolutions to clarify the corresponding procedures; Whereas the specific technical aspects of the disposal of radioactive waste in deep geological formations could be the subject of special provisions with regard to the management of fire risks; Whereas the WENRA association of the heads of European safety regulators adopted reference levels in January 2008 for protection against fire risks, which should be integrated into the French regulations; 27

Whereas, in a regulatory resolution, ASN shall specify the contents of the BNI safety analysis reports and that this resolution will create the framework for demonstrating management of the fire risks, Article 1 The appendix to this resolution specifies the rules applicable to BNIs with regard to the management of fire risks. In this respect, it supplements the implementation procedures in Title III of the above- mentioned order of 7th February 2012.

Article 2 This resolution shall apply as of the issue of their creation authorizations to BNIs which, on the date of approval of this present resolution, do not yet have such an authorization and are not operating with benefit of acquired rights.

For the other BNIs, this resolution shall apply as of the first day of the first civil six-month period following approval of this resolution, except for Articles 1.3.2, 4.1.2, 4.1.3, 4.1.5, 4.3.2 and 4.4.1 of its appendix, which shall apply as of 1st January 2017. However, if the installation is the subject of a commissioning authorization application under examination on the date of approval of this resolution, or which is submitted no later than one year after this date, this resolution shall apply to it six months after issue of the commissioning authorization.

Article 3 As an interim measure, the elements concerning fire risk management contained in the safety analysis report that exists on the date of approval of this resolution shall constitute the fire risks management case as defined in Article 1.1.1 of the appendix to this resolution. These elements are updated in the conditions specified for implementation of the provisions concerning the nuclear safety case in Article 9.4 of the above-mentioned order of 7th February 2012. These conditions could be supplemented by an ASN resolution on the safety analysis report.

Article 4 In the event of particular difficulties with implementing this resolution, the licensee may send ASN a duly justified waiver request. It shall enclose with its request proposed compensatory measures accompanied with an implementation time frame. The licensee shall justify that given current knowledge and the best available techniques, the practices and the vulnerability of the installation, these measures enable a level of protection against fire risks to be achieved that is as high as possible, in economically acceptable conditions.

ASN may grant a waiver to which compensatory prescriptions are attached, by means of a resolution issued in accordance with the procedures defined in Article 18 of the above-mentioned decree of 2nd November 2007.

Article 5 The ASN Director General is tasked with implementation of this resolution, which shall be published in the ASN Official Bulletin after its approval by the Minister in charge of nuclear safety.

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Provisions concerning the management of fire risks TITLE 1 GENERAL PROVISIONS Chapter 1.1 Definitions Article 1.1.1 For the purposes of this resolution, the expressions: accident, activity important for protection (AIP), internal failure, nuclear safety case, element important for protection (EIP),

establishment, defined requirement, licensee, normal operation, incident, emergency situation, dangerous substance, nuclear safety are as defined in the above-mentioned order of 7th February 2012.

For implementation of this resolution, the following definitions are adopted:

- general fire alarm: audible signal with the purpose of warning the occupants of the need to evacuate the premises;

- limited fire alarm: audible and visual signal other than the general alarm signal, with the purpose of warning either the establishments fire safety unit, or the management or security guard, or the specially designated personnel, of the existence and location of an incident;

- protected route: a route needed by the personnel and the emergency services for access in the event of a fire to the locations necessary for attaining and maintaining a safe state in the BNI.

- fire risks management case: part of the nuclear safety case concerning the prevention of fire risks and protection against their effects;

- provisions concerning the management of fire risks: all technical and organizational steps taken to demonstrate the management of fire risks to prevent the fire risks and mitigate the effects;

- containment sector: a volume the characteristics of which, in a fire situation, ensure that dispersion outside this volume of radioactive or dangerous substances liable to compromise the interests mentioned in Article L. 593-1 of the Environment Code, is limited;

- fire sector: a volume bounded by walls such that a fire occurring inside it cannot extend outside or so that a fire occurring outside cannot propagate inside it for a time long enough to enable it to be extinguished;

- fire zone: a volume bounded by barriers (geographical separation or wall) such that a fire occurring inside it cannot extend outside or so that a fire occurring outside cannot propagate inside it for a time long enough to enable it to be extinguished; Chapter 1.2 Objectives Article 1.2.1 Pursuant to Article 3.1 of the above-mentioned order of 7th February 2012, the licensee applies the principle of defence in depth to the management of fire risks.

The licensee thus implements successive and sufficiently independent levels of defence designed to protect or perform the functions defined in Article 3.4 of the above-mentioned order of 7th February 2012.

These levels in particular apply to:

- preventing the outbreak of fire; 29

- detecting and rapidly extinguishing any outbreaks, on the one hand to prevent them leading to a fire and, on the other, to restore a normal operating situation or, failing which, attain then maintain a safe BNI state;

- mitigating the aggravation and propagation of a fire which has not been stopped, in order to minimize its impact on nuclear safety and enable a safe BNI state to be attained or maintained;

- the management of accident situations resulting from a fire which could not be stopped, in order to mitigate the consequences for individuals and the environment.

Article 1.2.2 With regard to the management of fire risks and for implementation of the provisions concerning the nuclear safety case defined in Title III of the above-mentioned order of 7th February 2012, a fire risks management case is presented by the licensee in its safety analysis report. This case justifies that the design, construction and operating provisions regarding fire risks are appropriate and defined in accordance with the principles set out in Article 1.2.1. It includes the assessments of the consequences specified in Article 3.7 of the above-mentioned order of 7th February 2012. It is established using an approach that is proportionate to the issues and implications, pursuant to the provisions of Article 1.1 of the above-mentioned order of 7th February 2012.

Article 1.2.3 Within the framework of Articles 1.2.1 and 1.2.2, the licensee implements fire risk management provisions taking account of all technical aspects and pertinent organizational and human factors.

In particular, in the event of a fire, these provisions help ensure the protection of the persons necessary for the operations involved in attaining and maintaining a safe BNI state and in providing the firefighting response.

Article 1.2.4 Prior to taking up their posts, all the licensee's personnel receive general training in what to do in the event of a fire and the fire risks specific to their workstation or their activity.

For outside contractor personnel, the licensee ensures that they have received appropriate training in particular BNI risks, according to the duties assigned to them.

Chapter 1.3 Identification of provisions and EIP concerning the management of fire risks Article 1.3.1 Among the EIP identified pursuant to Article 2.5.1 of the above-mentioned order of 7th February 2012, the licensee determines those which need to be protected from the effects of a fire, as well as the related defined requirements.

Article 1.3.2 On the basis of the fire risks management case, the licensee:

• identifies the EIP to be protected from the effects of a fire and the related defined requirements;

• determines the provisions for the prevention of fire risks and protection against their effects.

Among these and in accordance with Articles 2.5.1 and 2.5.2 of the above- mentioned order of 7th February 2012, the licensee identifies the EIP and any AIP as well as the related defined requirements. These EIP are designed and installed in the BNI such as to reduce the probability of a fire occurring, ensuring detection and mitigating the consequences.

Chapter 1.4 Periodic checks and tests Article 1.4.1 The fire risk management provisions are the subject of periodic checks, maintenance and tests, in accordance with the applicable regulations and standards and with the requirements arising from the fire risks management case.

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The licensee defines and justifies the appropriate provisions to ensure management of the fire risks, as well as the nature and frequency of the planned checks.

TITLE 2 PROVISIONS TO PREVENT THE OUTBREAK OF FIRE Chapter 2.1 Construction materials and layout Article 2.1.1 The licensee chooses and utilizes construction materials, interior layouts and equipment such as to minimize the outbreak of fire, the development and the propagation of a fire and its effects.

Chapter 2.2 Management of combustible materials Article 2.2.1 The licensee defines the procedures for managing, monitoring and tracking combustible materials and the organization put into place to minimize their quantities, in each volume, room or group of rooms, considered in the fire risk management demonstration. The nature, maximum quantity and location of the combustible materials considered in the fire risks management case are defined in documents belonging to the licensees integrated management system.

The areas in which the combustible materials considered in the fire risks management case are either prohibited or authorized, are shown by continuous, visible and permanent marking in the rooms or groups of rooms, or outside the buildings.

Article 2.2.2 The licensee limits the quantities of combustible materials in the places they are used to only what is strictly necessary for the normal operation of the BNI and, in any case, to values not to exceed those considered in the fire risks management case.

Given the rapid spread of a fire involving flammable liquids or gases, fire risk management provisions are adopted to prevent such liquids or gases, which are present in the BNIs, from being able to cause a fire or fuel its spread. When not in use, they are placed in areas, rooms or equipment appropriate to their nature and quantity.

Article 2.2.3 The licensee takes the necessary steps to ensure that the flammability ranges for the gases or vapors present or generated in its BNI cannot be reached, except in specific situations justified in the fire risks management study.

Chapter 2.3 Prevention plan and fire permit Article 2.3.1 Hot spot work can only be performed after issue of a fire permit for which a specific nuclear safety risk assessment was carried out and which has been duly signed by the licensee, after checking the possible interactions between simultaneous worksites.

Article 2.3.2 The licensee ensures the compatibility between the fire risks management case and the steps included in the prevention plan specified in Articles R. 4512-6 to R. 4512-12 of the Labour Code, or the fire permit for the works envisaged.

Article 2.3.3 The fire permit specifies the special measures to be taken to prepare for and carry out the work with regard to the fire risk. This document officially lays out all prevention and consequences mitigation measures to be taken to manage the fire risk presented by this work. It identifies any scheduled unavailability of the fire risk management provisions and defines the compensatory measures.

The fire risk management provisions disabled for the duration of this work, shall be returned to service as soon as their unavailability is no longer required.

Chapter 2.4 Prevention of risks of electrical or electrostatic origin 31

Article 2.4.1 The licensee takes steps to prevent any risk of an outbreak of a fire of electrical origin. It in particular ensures that the electrical equipment and its components and the ventilation systems removing the heat generated by the electrical equipment are suitably maintained and that the electrical protections are appropriately adjusted.

Article 2.4.2 The electrical conductors and cables present in the buildings housing radioactive or dangerous substances liable, in the event of a fire, to compromise the interests mentioned in Article L. 593-1 of the Environment Code, or the EIP to be protected from the effects of a fire, are in conformity with class C1, defined by the above-mentioned order of 21st July 1994 with respect to their reaction to fire.

However, if it is technically impossible to employ electrical conductors and cables conforming to this class, the licensee shall justify the use of another class in the fire risks management case.

Article 2.4.3 To protect the installations from the effects of stray currents, the licensee takes precautions to limit the build-up of electrostatic charges which could create a fire risk situation, in particular in the premises containing flammable substances, and ensures their evacuation in conditions that do not affect the safety of the BNI.

Article 2.4.4 The following construction and operation measures are in particular applied:

- electrical continuity and earthing of permanent or temporary conductors;

- limitation of the use of insulating materials liable to accumulate an electrostatic charge;

- limitation of flow rates of low-conductivity flammable fluids and flammable dusts.

Failing which, in the fire risks management case, the licensee justifies that the steps taken are compliant with the requirements of Article 2.4.3.

TITLE 3 FIRE DETECTION AND INTERVENTION PROVISIONS Chapter 3.1 Fire detection and associated safety systems Article 3.1.1 The BNI comprises one or more fire detection systems, designed to:

- monitor the premises and outdoor areas identified in the fire risks management case;

- ensure the operation of the associated safety systems, whether or not automatically actuated.

These systems and devices meet the requirements assigned to them in the fire risks management case.

The design and operation of these systems enable a fire outbreak to be rapidly, easily and accurately located, the general fire alarm concerned to be tripped plus, as necessary, the automatically actuated safety devices. These systems and devices are designed and built so as to be effective and function permanently; they are maintained so as to keep any period of unavailability to a minimum. They have a back-up electrical power supply of sufficient autonomy to ensure that a safe BNI state is maintained in the event of failure of the main power supply source.

Article 3.1.2 The limited fire alarm is transmitted to a location when monitoring personnel are permanently present. It allows easy interpretation of the information by the response teams. It is clearly distinct from any other alarm which could appear in the BNI.

Article 3.1.3 Failure of the fire detection systems or devices and automatically actuated safety devices is indicated by an alarm transmitted to a location where monitoring personnel are permanently present.

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Chapter 3.2 Firefighting resources Section 3.2.1 Infrastructures Article 3.2.1-1 The BNIs are permanently equipped with fire response and firefighting resources specified in the fire risks management case. These resources are defined in particular to take account of the foreseeable types of fire, the risks specific to the BNI and any difficulties with access to the premises. The risk of criticality is in particular examined.

Article 3.2.1-2 The material response and firefighting resources put into place, as well as the system for recovery of the extinguishing agents used, are such that their implementation cannot lead to the loss of one of the functions mentioned in Article 3.4 of the above-mentioned order of 7th February 2012 or a loss of containment of dangerous substances liable to compromise the interests mentioned in Article L. 593-1 of the Environment Code in the event of a fire.

Article 3.2.1-3 The BNIs internal material response and firefighting resources are positioned at clearly marked locations that are rapidly accessible in all circumstances and kept in good working order.

Article 3.2.1-4 A network protected from frost, preferably a meshed network, supplies water to the firefighting systems such as the fire plugs and fire hydrants located outside the buildings and, as applicable, the dry or wet risers and the fire hose reels (RIA) inside the buildings. It is designed and installed consistently with the fire risks management case.

Section 3.2.2 Operational organization Article 3.2.2-1 The fire response and firefighting resources available to the licensee internally are designed pursuant to III of Article 2.1.1 of the above-mentioned order of 7th February 2012.

They are implemented according to an organization predetermined by the licensee. This organization enables action to be taken, the rapidity and effectiveness of which are compatible with the interventions included in the fire risks management case, more specifically for management of plausible trigger event combination situations, both in the BNI considered and in all the BNIs on a site. This leads to the definition of the equipment and personnel necessary for fire response and firefighting, consistently with the fire risks management case. All firefighting actions, in response to either a call-out or an alarm, shall be carried out at least in pairs, to ensure that it is done effectively.

Article 3.2.2-2 If the licensee does not itself have all the response and firefighting resources described in the fire risks management case, it proves that at all times it has sufficient material and human resources to carry out the steps necessary, pending the arrival of emergency resources from outside the BNI, taking account of any access difficulties.

The licensee justifies resorting to these outside services by considering the material, human and organizational provisions and their foreseeable deployment times in order to perform the actions identified in the fire risks management case. The steps taken to facilitate their intervention are specified. The licensee more specifically takes account of the risks associated with radioactive or dangerous substances and plausible combinations of the trigger events considered in the nuclear safety case, in accordance with the provisions of Article 3.2 of the above-mentioned order of 7th February 2012.

Article 3.2.2-3 In order to ensure the effectiveness of the organization of the response teams and their operational capability, the licensee uses regular exercises to test:

- the response methods, instructions, plans and organization notices designed to restore normal operation of the BNI or, failing which, to attain and maintain a safe BNI state, in the event of a fire; 33

- the use of response and personnel evacuation resources

- calling in and admitting outside emergency resources.

The response procedures take account of the risk of the dissemination of radioactive or dangerous substances liable to compromise the interests mentioned in Article L. 593-1 of the Environment Code, in the event of a fire.

Article 3.2.2-4 A sufficient number of available persons are designated as members of the response and firefighting teams. They are regularly trained and drilled, according to an annual programme appropriate to their duties.

Chapter 3.3 Access and circulation routes Article 3.3.1 The access and circulation routes necessary for the fire risks management case, are clearly marked out and kept clear of anything liable to obstruct circulation. The circulation and maneuvering areas necessary for access by the fire and emergency services and for deployment of response and firefighting resources are designed and laid out so that the vehicles of these services can move around without difficulty, including with regard to extending ladders.

They are thus designed and distributed so that they can be used in complete safety, taking account of the dimensions and layout of the buildings and their access bays.

Steps are taken to ensure that parked vehicles never obstruct the vehicles of the fire and emergency services.

Article 3.3.2 Inside the buildings, circulation ways and protected routes are laid out, marked and kept clear at all times, to facilitate the circulation and intervention of the emergency teams in the event of a fire.

TITLE 4 PROVISIONS DESIGNED TO PREVENT THE PROPAGATION OF A FIRE AND MITIGATE ITS CONSEQUENCES Chapter 4.1 Sectorization Article 4.1.1 The fire risks management case ensures that the BNIs fire sectors and zones can be identified and justified.

The use of fire sectors shall be considered a priority.

Special steps are taken in particular to minimize the propagation of smoke and the propagation of a fire by hot gases or by flaming flows or projections, in particular in the case of fire zones.

The possible temporary presence of combustible materials is taken into account when defining the steps taken.

Article 4.1.2 Steps are taken to ensure that a fire cannot simultaneously affect EIP that are to be protected from the effects of a fire and that provide functional redundancy. In this respect, they are not placed in the same fire sector or zone or, failing which, have sufficient protection to prevent failure caused by the same fire.

Article 4.1.3 The fire risks management case ensures that the protected routes can be identified and justified.

Article 4.1.4 The fire risks management case ensures that the containment sectors can be identified and justified.

The effectiveness of these sectors is maintained even in the case of deployment of the fire response and firefighting resources specified in the fire risks management case.

Article 4.1.5 In order to make the personnel more accountable for the implementation of preventive measures and to facilitate the response and firefighting, all accesses to the various 34

sectors and zones and the entire length of all protected routes are clearly signposted inside the BNI.

Chapter 4.2 Fire resistance of structures Article 4.2.1 The fire resistance of the structures of the buildings identified in the fire risks management case is sufficient to enable a safe BNI state to be attained and maintained in the event of a fire. The fire stability of the supporting elements of the structure of the buildings identified in the fire risks management case is at least two hours. This stability is calculated for a fire occurring inside or outside the buildings, considering the possible interactions with a fire developing in a nearby structure. The fire stability of the support elements of the structure of the buildings must not compromise the fire resistance of the fire sectors or zones inside them.

Article 4.2.2 In the case of existing buildings for which such a fire stability requirement could not be met in acceptable technical-economic conditions, the licensee identifies and justifies the specific measures for ensuring that a safe BNI state is attained and maintained in the event of a fire.

Chapter 4.3 Ventilation - smoke extraction Article 4.3.1 The ventilation systems are designed and operated so that, in the event of a fire, they do not contribute to its propagation, while limiting:

- the dissemination of radioactive substances inside the BNI

- release into the environment of radioactive or dangerous substances liable to compromise the interests mentioned in Article L. 593-1 of the Environment Code.

Moreover, these systems facilitate intervention and the mitigation of the consequences in the areas involved in the fire, by managing the risks linked to pyrolysis gases and other unburned products.

If the objectives prove to be incompatible, the fire risks management case justifies the solution adopted.

Pursuant to the fire risks management case, operation of the ventilation in the event of a fire is the subject of a specific analysis and of particular procedures in the BNI. The organization put into place by the licensee enables these procedures to be implemented.

Article 4.3.2 The elements of the ventilation systems necessary for attaining and maintaining a safe BNI state are capable of performing their function despite the effects of a fire which could affect them for a given period, consistently with the fire risks management case or, as applicable, are protected from the effects of a fire. More specifically, when they participate in the boundaries between the fire sectors or zones they serve or through which they pass, these elements have a fire resistance equivalent to that of the fire sectors they serve, or are isolated from them by appropriate fire dampers.

In the case of premises presenting a risk of the release of radioactive or dangerous substances liable to compromise the interests mentioned in Article L. 593-1 of the Environment Code, in the event of a fire, the licensee justifies the situations for which static containment is preferable to dynamic containment or smoke extraction.

Article 4.3.3 The smoke-extraction systems in the buildings, identified by the fire risks management case, are also designed and utilized in order to:

- limit the propagation of the fire;

- facilitate the intervention by the response teams; 35

while limiting the dispersion into the environment of radioactive or dangerous substances liable to compromise the interests mentioned in Article L. 593-1 of the Environment Code, in the event of a fire.

Chapter 4.4 Operating devices Article 4.4.1 The operating devices necessary for management of the fire risk, such as the fire damper controls, are designed and installed so that they can be operated and are operational in the event of a fire. More specifically, they are accessible from protected routes whenever necessary. The licensee has at its disposal the necessary trained personnel for utilization of these devices, along with the appropriate documentation.

3.5.2 New Reactors There is no specific regulation for new reactors. New reactors are regulated using the same guidelines as existing reactors described in paragraph 3.5.1.

3.5.3 Supplementary Information

[FRA-01] Act No. 2006-686 of 13 June, 2006 on Transparency and Security in the Nuclear Field, http://www.french-nuclear-safety.fr/References/Regulations/Act-No.-2006-686-of-13-June-2006.

[FRA-02] Order of 7 February, 2012 setting the general rules relative to basic nuclear installations, http://www.french-nuclear-safety.fr/References/Regulations/Order-of-7-February-2012.

[FRA-03] ASN resolution 2014-DC-0417 of 28 January 2014 concerning the rules applicable to basic nuclear installations (BNI) with regard to the management of fire risks.

36

3.6 Germany 3.6.1 Existing Reactors In principle, the regulatory framework for nuclear power plants (NPP) in Germany is based on deterministic requirements supplemented by probabilistic ones for safety assessment. The regulation comprises high level comprehensive claims such as the most recent Safety Requirements for Nuclear Power Plants [BMU-15] as well as lower level detailed technical nuclear safety standards and rules incorporated in a corresponding pyramid type legal structure as shown in Figure 5.

Basic Law generally binding Federal legislator Atomic Energy Act Ordinances Federal Government, binding for Federal Council General administrative authorities provisions BMU publications Federal Government, - Safety Requirements for NPP Lnder authorities

- Guidelines and recommendations binding by specification in the licence or by Advisory bodies RSK Guidelines, RSK and SSK recommendations supervisory measures in the individual case KTA KTA Safety Standards Technical specifications for components and systems Industry Organisation and operating manuals Figure 5 Nuclear Regulatory Framework in Germany For nuclear facilities in Germany - covering commercially operated reactors as well as non-commercially operated ones (e.g., research reactors) and other installations such as nuclear waste storage facilities, fire safety is addressed in non-nuclear as well as nuclear specific regulations. These are valid for already existing facilities as well as for new built installations.

High level nuclear specific requirements on fire safety at nuclear power plants are most recently provided in the Safety Requirements for Nuclear Power Plants and its specific Annex 3 on Internal and External Hazards as promulgated in the Federal Gazette in January 2013 and updated in March 2015 [BMU-15]. These require in principle that all items required for safe shutdown of the nuclear reactor, for maintaining it in a shutdown state, for residual heat removal or for prevention of the release of radioactive materials shall be designed such and constantly kept in such a condition that they can fulfil their safety related tasks even in case of any internal hazard or site specifically identified external hazard, e.g., internal or external fire. The safety system as well as the emergency systems shall be designed such that they remain effective in the event of fire.

The general safety concept is based on a multi-level confinement of the radioactive inventory (Barrier Concept).

It is principally required that all equipment needed for the safe shutdown of the nuclear reactor, for maintaining it in a shutdown state, for residual heat removal or the prevention of a release of 37

radioactive materials shall be designed as and constantly kept in a condition that they can fulfil their safety related functions even in the event of fire. Fires that might inadmissibly impair the required functions of equipment of the safety system shall either be reliably prevented or limited in their consequences. In this context, passive protection means shall be preferred. If inadmissible consequences cannot be reliably prevented by passive means, reliable active means shall be in place. Redundant parts of items important to safety shall be either installed in physically separated plant areas or protected such that in the event of any plant internal fire, a failure of more than one redundant train is reliably prevented.

Moreover, the technical safety concept requests that the design of systems, structures and components (SSC) against hazards including fires is based on:

- those natural hazards with the most severe consequences or other external hazards to be postulated at the site under consideration;

- the special characteristics of external hazards of long duration;

- combinations of several external hazards or combinations of these hazards with plant internal event; these combinations shall be postulated, if the combined events or hazards either show a causal relationship or if their simultaneous occurrence has to be postulated according to its probability and the expected extent of damage.

These general requirements shall be met by an in-depth protection concept starting with fire prevention. Precautionary measures shall ensure that fires impairing the required function of items important to safety shall be either prevented or sufficiently limited in their effects. The requirements concerning effectiveness and reliability of preventive measures depend on the occurrence frequency of those hazards, against which the protection is provided, and on their potential effects.

Protection means for the protection against plant internal fires and their consequences shall be in place both inside and outside of buildings. Inadmissible impacts of fires and their consequences shall be prevented by active and passive fire protection means.

Fire protection means shall be planned and implemented such that defense-in-depth is realized:

- Suitable protection means shall be in place to prevent the occurrence of incipient fires.

- Fires which have nevertheless occurred shall be quickly detected and extinguished.

- The propagation of any fire neither extinguished nor self-extinguished shall be limited.

A fire protection concept shall be developed and documented. The documentation shall be kept up to date. In case of any plant modification, its effects on the existing fire protection concept shall be assessed and as far as necessary enhanced.

A fire hazard analysis shall be performed and documented. The documentation shall be kept up to date.

The entire fire protection means shall ensure that even in the event of a random failure of a single fire protection means the required safety functions are not inadmissibly impaired.

An ignition of combustibles should be postulated. Deviations from this requirement are admitted, if the combustible is encapsulated and it has been demonstrated that the encapsulation maintains its operability during specified normal operation and in the event of any accident.

38

Fire loads and potential ignition sources shall be limited to the extent necessary for safe operation.

For prevention of an ignition by potential ignition sources, fire loads needed for plant operation shall be sufficiently physically separated from these ignition sources at any location, where permitted by design and requirements for the operation of items important to safety. Plant areas containing considerable fire loads should be separated by sufficiently rated fire barriers.

Redundant trains of the safety system should be separated by sufficiently rated fire barriers to prevent a loss of more than one redundant train in case of fire. If the protection required in the event of fire cannot be ensured by structural protection means due to systems engineering or operational reasons, an equivalent level of protection shall be ensured by other (compensatory) fire protection means or by a combination of different fire protection means.

For transient combustibles in connection with maintenance work special protection means shall ensure that the plant safety is not inadmissibly impaired.

Passive structural fire protections means shall ensure the fire safety of buildings and structures.

Only non-combustible constructions and structural elements should be used. The use of combustible materials is only permissible, if the use of such materials cannot be avoided, e.g.

insulation materials for cooling pipes, de-contaminable coatings. Only non-combustible operating supplies should be used. Exceptions are control and lubricating fluids as well as other combustible materials that cannot be avoided for operational reasons.

Instrument and Control (I&C) wires and cables should be routed separately from heated pipes or pipes carrying combustible media. Power cables shall be sufficiently separated from signal and control cables. In the exceptional case of unavoidable crossings of I&C wires and cables with high-temperature pipes or pipes carrying combustible media or with power cables, particular protection means shall be in place. Adequate protection means shall ensure that even in the event of fire cables for power supply or I&C cables are not inadmissibly impaired.

The restrictions for the controlled area shall be considered in the selection and installation of active and passive fire protection means.

In the event of fire, particularly in plant areas with equipment of the safety system and in controlled areas, adequate protection means shall ensure a reliable and fast fire detection and alarm.

Adequate protection means for fire detection, alarm and suppression shall ensure that fires in the containment can be quickly and reliably detected and extinguished efficiently, even without smoke removal.

Adequate protection means for a timely detection and alarm of any hazard and appropriate precautions for rapid escape and rescue activities via escape and rescue routes shall ensure that in case of danger persons can reach the outside quickly and can be rescued from the outside.

Access and escape routes shall be provided within the buildings. These shall be protected against fire effects for an appropriate time period to allow for self-rescue, rescue of persons, fire extinguishing as well as for personnel actions required for safety reasons.

Stationary fire extinguishing systems should be actuated automatically. Remote controlled or local manually actuated extinguishing systems are permissible, if the fire effects are controlled until these extinguishing systems come into effect.

Automatically actuated stationary extinguishing systems shall be designed and secured in such a way that neither disturbances occurring at them or at parts of them nor faulty actions /

39

maloperations do neither impair the required function of equipment of the safety system nor of structural elements for separation of fire compartments.

The entire fire protection means shall regularly be subject to in-service inspections with respect to their required function. Test intervals shall be specified according to the safety significance of the equipment to be protected.

For fire suppression, an efficient professional on-site fire brigade shall be established, equipped and maintained according to the existing non-nuclear regulations. In addition, the local off-site fire brigade shall be familiarized with the plant and the different plant areas as well as with the specific boundary conditions at a nuclear power plant. The corresponding instructions shall be repeated at regular intervals. Fire drills shall be conducted at appropriate time intervals.

It shall be ensured that all means required for ensuring safe operation and control of events on levels of defense 3 and 4a can also be taken in the event of fire extinguishing.

3.6.2 New Reactors Since Germany is in the process of nuclear phase out, regulations for new power reactors do not exist.

3.6.3 Supplementary Information Detailed technical requirements are provided by the German nuclear standards of KTA (German abbreviation for Kerntechnischer Ausschuss, Nuclear Safety Standards Commission) available also in English, in particular:

[KTA-15] Nuclear Safety Standards Commission (KTA, German for Kerntechnischer Ausschuss): Fire Protection in Nuclear Power Plants, Part 1: Basic Requirements, KTA 2101.1, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2015-11, December 2015, http://www.kta-gs.de/e/standards/2100/2101_1_engl_2015_11.pdf.

[KTA-15a] Nuclear Safety Standards Commission (KTA, German for Kerntechnischer Ausschuss): Fire Protection in Nuclear Power Plants, Part 2: Fire Protection of Civil Structures, KTA 2101.2, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2015-11, December 2015, http://www.kta-gs.de/e/standards/2100/2101_2_engl_2015_11.pdf.

[KTA-15b] Nuclear Safety Standards Commission (KTA, German for Kerntechnischer Ausschuss): Fire Protection in Nuclear Power Plants, Part 3: Fire Protection of Mechanical and Electrical Plant Components, KTA 2101.3, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2015-11, December 2015, http://www.kta-gs.de/e/standards/2100/2101_3_engl_2015_11.pdf.

The KTA fire safety standards provide nuclear specific technical requirements with specific focus on all deviations from non-nuclear ordinances, standards and norms, resulting from the nuclear requirements as given by the Atomic Energy Act and the next level of requirements represented by the above mentioned Safety Requirements for Nuclear Power Plants.

The German and/or European non-nuclear requirements with regard to fire safety of industrial buildings and in industrial facilities cover:

Atomic Energy (01/2016) Act on the Peaceful Utilisation of Atomic Energy and the Act Energy Act Protection against its Hazards (Atomic Energy Act) of 23 December 1959, as amended and promulgated on 15 July 1985, last Amendment of 20 November 2015, 40

http://www.bfs.de/SharedDocs/Downloads/BfS/EN/hns/a1-english/A1-01-16-AtG.html ArbStttV (12/2008) Ordinance on work places (Work Place Ordinance -

ArbStttV) of August 12, 2004 (BGBl. I, p. 2179) most recently changed by Article 9 of the ordinance of December 18, 2008 (BGBl. I, p. 2768)

AtSMV (10/2010) Ordinance on the Nuclear Safety Officer and the Reporting of Incidents and other Events (Nuclear Safety Officer and Reporting Ordinance) of 14 October 1992, last Amendment of 8 June 2010, http://www.bfs.de/SharedDocs/Downloads/BfS/EN/hns/a1-english/A1-10-10.pdf?__blob=publicationFile&v=3 BGV C 16, VBG (01/1987) Accident prevention regulation - nuclear power plants 30 BetrSichV (02/2015) Verordnung zur Neuregelung der Anforderungen an den Arbeitsschutz bei der Verwendung von Arbeitsmitteln und Gefahrstoffen (Artikel 1 Verordnung über Sicherheit und Gesundheitsschutz bei der Verwendung von Arbeitsmitteln (Betriebssicherheitsverordnung -

BetrSichV; English: Operational Safety Ordinance)

SiAnf (03/2015) Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety, Safety Requirements for Nuclear Power Plants, as amended and published on November 22, 2012 and revised version of March 3, 2015, [BMU-15]

MBO (12/2009) Reference building code (Special Commission on Construction Surveillance of the ARGEBAU)

Construction (01/1984) Construction supervision guideline on fire protection Supervision requirements regarding ventilation systems Guideline Ventilation Systems German non-nuclear norms DIN DIN 3221 (01/1986) Fire hydrants, under ground DIN 3222 (01/1986) Fire hydrants, above ground DIN 4102-1 (05/1998)

DIN 4102-2 (09/1977) Fire behaviour of building materials and building components, Part 2: Building components; definitions, requirements and tests DIN 14 090 (06/1977) Areas for the fire brigade on premises DIN 14 095 (08/1998) Ground plans of buildings for fire brigade use DIN 14 210 (11/1982) Water pool for fire fighting 41

DIN 14 220 (04/1991) Fire wells DIN 14 230 (04/1991) Underground water-tanks for fire fighting DIN 14 461-1 (02/1998) Delivery valve installation - Part 1: Hose reel with semi-rigid hose DIN 14 461-6 (06/1998) Delivery valve installation - Part 6: Dimensions of cabinets and installation of hose reels with lay-flat hoses according to DIN EN 671-2 DIN 18230-1 (2010-09) Structural fire protection in industrial buildings - Part 1:

Analytically required fire resistance time DIN EN 54-1 (10/1996) Fire detection and fire alarm systems - Part 1:

Introduction; German version EN 54-1:1996 DIN EN 671-1 (02/1996) Fixed firefighting systems - Hose systems - Part 1: Hose reels with semi-rigid hose; German version EN 671-1:2001 DIN EN 671-2 (02/1996) Fixed firefighting systems - Hose systems - Part 2: Hose systems with lay-flat hose; German version EN 671-2:2001 DIN VDE 0833-1 (2014-10) Alarm systems for fire, intrusion and hold-up - Part 1:

General requirements ASR 13/1, 2 (06/1997) Fire suppression equipment (BArbBl. 1997, Nr. 7/8, p. 70-73)

ZH 1/201 (1996) Standards on equipping work places with fire extinguishers In addition, the following nuclear regulatory documents, such as ordinances and standards have to be considered:

BMI Guideline (10/1980) Guideline relating to the assurance of the necessary Necessary knowledge of the persons otherwise engaged in the Knowledge operation of nuclear power plants of October 30, 1980 (GMBl. 1980, p. 652)

RSK Guidelines (10/1981) RSK Guidelines for Pressurized Water Reactors, 3rd edition for PWR of October 14, 1981 (BAnz. No. 69 of April 14, 1982, Supplement No. 19/82)

Incident (10/1983) Guidelines on the evaluation of the design of nuclear power Guidelines plants with pressurized water reactors against incidents in terms of Sec. 28 para. 3 Radiological Protection Ordinance (Incident Guidelines) of October 18, 1983 (BAnz. No. 245 of December 31, 1983)

Radiation (02/2012) Ordinance on the protection from damage by ionizing Protection radiation (Radiological Protection Ordinance - StrlSchV) of Ordinance July 20, 2001, (BGBl. I S. 1714; 2002 I S. 1459), most recently changed by article 5, par. 7 of the law of February 24, 2012 (BGBl. I, p. 212 42

Recommendation (10/1977) Recommendations on the planning of accident management Accident measures by the operator of nuclear power plants of Management December 27, 1976 (GMBl. 1977, p. 48), most recently Measures changed by ordinance of October 18, 1977 (GMBl. 1977, S.

664)

PSA Guide (11/2005) Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit (BMU): Sicherheitsüberprüfung für Kernkraftwerke gem §19a des Atomgesetzes - Leitfaden Probabilistische Sicherheitsanalyse, 31. Januar 2005, Bekanntmachung vom 30. August 2005, Bundesanzeiger, Jahrgang 57, Nummer 207a, ISSN 0720-6100, 3. November 2005 [BMU 05]

Supplement on (10/2005) Facharbeitskreis (FAK) Probabilistische Sicherheitsanalyse PSA Methods to für Kernkraftwerke: Methoden zur probabilistischen PSA Guide Sicherheitsanalyse für Kernkraftwerke (Probabilistic Safety Analysis for Nuclear Power Plants: Methods for Probabilistic Safety Analysis), Stand: August 2005, BfS-SCHR-37/05, Salzgitter, Germany, October 2005 (in German only)

[FAK 05]

Supplement on (10/2005) Facharbeitskreis (FAK) Probabilistische Sicherheitsanalyse PSA Data to PSA für Kernkraftwerke: Daten zur Quantifizierung von Guide Ereignisablaufdiagrammen und Fehlerbumen (Probabilistic Safety Analysis for Nuclear Power Plants: Data for Quantification of Event Sequence Diagrams and Fault Trees), Stand: August 2005, BfS-SCHR-38/05, Salzgitter, Germany, October 2005 (in German only) [FAK 05a]

Additional (09/2016) Facharbeitskreis (FAK) Probabilistische Sicherheitsanalyse Supplement on für Kernkraftwerke: Methoden und Daten zur PSA Methods and probabilistischen Sicherheitsanalyse für Kernkraftwerke, Data Stand: Mai 2015, BfS-SCHR-61-16, Salzgitter, Germany, September 2016 (in German only) [FAK 16]

Further German Nuclear KTA standards:

KTA-1201 (2015- Requirements for the operating manual,

11) http://www.kta-gs.de/e/standards/1200/1201_engl_2015_11.pdf KTA-1202 (2017- Requirements for the testing manual,

11) http://www.kta-gs.de/e/standards/1200/1202_engl_2017_11.pdf KTA-1301.1 (2017- Radiation Protection Considerations for Plant Personnel in the

11) Design and Operation of Nuclear Power Plants; Part 1: Design, http://www.kta-gs.de/e/standards/1300/1301_1_engl_2017_11.pdf KTA-1301.2 (2014- Radiation Protection Considerations for Plant Personnel in the

01) Design and Operation of Nuclear Power Plants; Part 2: Operation, http://www.kta-gs.de/e/standards/1300/1301_2_engl_2014_11.pdf KTA-1401 (2017- General Requirements Regarding Quality Assurance,

11) http://www.kta-gs.de/e/standards/1400/1401_engl_2017_11.pdf 43

KTA-1402 (2017- Integrated Management Systems for the Safe Operation of

11) Nuclear Power Plants, http://www.kta-gs.de/e/standards/1400/1401_engl_2017_11.pdf KTA-1403 (2017- Ageing Management in Nuclear Power Plants,

11) http://www.kta-gs.de/e/standards/1400/1403_engl_2017_11.pdf KTA-1404 (2013- Documentation During the Construction and Operation of Nuclear

11) Power Plants, http://www.kta-gs.de/e/standards/1400/1404_engl_2013_11.pdf KTA-2103 (2015- Explosion Protection in Nuclear Power Plants with Light Water

11) Reactors (General and Case-Specific Requirements),

http://www.kta-gs.de/e/standards/2100/2103_engl_2015_11.pdf KTA-2201.1 (2011- Design of Nuclear Power Plants Against Seismic Events, Part 1:

11) Principles, http://www.kta-gs.de/e/standards/2200/2201_1_engl_2011_11.pdf KTA-2201.2 (2012- Design of Nuclear Power Plants Against Seismic Events, Part 2:

11) Subsoil, http://www.kta-gs.de/e/standards/2200/2201_2_engl_2012_11.pdf KTA-2201.3 (2013- Design of Nuclear Power Plants Against Seismic Events, Part 3:

11) Structural Components, http://www.kta-gs.de/e/standards/2200/2201_3_engl_2013_11.pdf KTA-2201.4 (2012- Design of Nuclear Power Plants Against Seismic Events, Part 4:

11) Components, http://www.kta-gs.de/e/standards/2200/2201_4_engl_2012_11.pdf KTA-2201.5 (2015- Design of Nuclear Power Plants Against Seismic Events, Part 5:

11) Seismic Instrumentation Components, http://www.kta-gs.de/e/standards/2200/2201_5_engl_2015_11.pdf KTA-2201.6 (2015- Design of Nuclear Power Plants Against Seismic Events,

11) Part 6: Post-Seismic Measures, http://www.kta-gs.de/e/standards/2200/2201_6_engl_2015_11.pdf KTA-2206 (2009- Design of Nuclear Power Plants Against Damaging Effects from

11) Lightning, http://www.kta-gs.de/e/standards/2200/2206_engl_2009_11.pdf KTA-2207 (2004- Flood Protection for Nuclear Power Plants,

11) http://www.kta-gs.de/e/standards/2200/2207_engl_2004_11.pdf KTA-2501 (2010- Structural Waterproofing of Nuclear Power Plants,

11) http://www.kta-gs.de/e/standards/2500/2501_engl_2015_11.pdf KTA-3301 (2015- Residual Heat Removal Systems of Light Water Reactors,

11) http://www.kta-gs.de/e/standards/3300/3301_engl_2015_11.pdf KTA-3402 (2014- Airlocks on the Reactor Containment of Nuclear Power Plants -

11) Personnel Airlocks, http://www.kta-gs.de/e/standards/3400/3402_engl_2014_11.pdf 44

KTA-3403 (2010- Cable Penetrations through the Reactor Containment Vessel,

11) http://www.kta-gs.de/e/standards/3400/3403_engl_2015_11.pdf KTA-3501 (2015- Reactor Protection System and Monitoring Equipment of the

11) Safety System, http://www.kta-gs.de/e/standards/3500/3501_engl_2015_11.pdf KTA-3601 (2017- Ventilation Systems in Nuclear Power Plants,

11) http://www.kta-gs.de/e/standards/3600/3601_engl_2017_11.pdf KTA-3602 (2003- Storage and Handling of Fuel Assemblies and Associated Items in

11) Nuclear Power Plants with Light Water Reactors, http://www.kta-gs.de/e/standards/3600/3602_engl_2003_11.pdf KTA-3604 (2005- Lagerung, Handhabung und innerbetrieblicher Transport

11) radioaktiver Stoffe (mit Ausnahme von Brennelementen) in Kernkraftwerken, http://www.kta-gs.de/e/standards/3600/3604_engl_2005_11.pdf KTA-3605 (2017- Treatment of Radioactively Contaminated Gases in Nuclear

11) Power Plants with Light Water Reactors, http://www.kta-gs.de/e/standards/3600/3605_engl_2017_11.pdf KTA-3701 (2014- General requirements for the electrical power supply in nuclear

11) power plants, http://www.kta-gs.de/e/standards/3700/3701_engl_2014_11.pdf KTA-3702 (2014- Emergency Power Generating Facilities with Diesel-Generator

11) Units in Nuclear Power Plants, http://www.kta-gs.de/e/standards/3700/3702_engl_2014_11.pdf KTA-3705 (2013- Switchgear Facilities, Transformers and Distribution Networks for

11) the Electrical Power Supply of the Safety System in Nuclear Power Plants, http://www.kta-gs.de/e/standards/3700/3705_engl_2013_11.pdf KTA-3901 (2017- Communication Means for Nuclear Power Plants,

11) http://www.kta-gs.de/e/standards/3900/3901_engl_2017_11.pdf KTA-3904 (2017- Control Room, Remote Shutdown Station and Local Control

11) Stations in Nuclear Power Plants, http://www.kta-gs.de/e/standards/3900/3904_engl_2017_11.pdf 45

3.7 Japan 3.7.1 Existing Reactors The legal framework on fire protection in Japan consists of three acts:

a. Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors [JPN-01],

b. Fire Service Act [JPN-02],

c. Building Standard Act [JPN-03].

Under the act (a), new fire protection requirements are defined in the following ordinances (i), (ii), regulatory guides (iii), (iv), and review standard (v), which came into force on July 8th, 2013:

i. The NRA Ordinance on Standards for the Location, Structures and Equipment of Commercial Power Reactors (Article 6 and 8) [JPN-04],

ii. The NRA Ordinance on Technical Standards for Commercial Power Reactors Facilities (Article 7 and 11) [JPN-05],

iii. The Regulatory Guide of the NRA Ordinance on Standards for the Location, Structure, and Equipment of Commercial Power Reactors [JPN-06],

iv. The Regulatory Guide of the NRA Ordinance on Technical Standards for Commercial Power Reactors Facilities [JPN-07],

v. The Fire Protection Review Standard for Commercial Power Reactor Facilities [JPN-08].

3.7.1.1 General Technical Requirements for Nuclear Reactor Facilities In the ordinance (i) [JPN-04], there are two requirements on the basic design of the fire protection:

• Article 6: Prevention of damage due to external hazards Structures, systems and components (SSCs) with safety functions shall be designed so that the safety of the nuclear reactor facilities will not be impaired by other postulated external hazards than earthquake, tsunami and concomitant events. SSCs with safety functions of especially high importance shall be of the design that reflects appropriate safety considerations against the severest conditions of postulated external hazards or appropriate combinations of natural forces and design basis accidents induced loads.

"Postulated external hazards" refer to on-site natural phenomena possible to occur including flood, wind (typhoon), tornado, freezing, rainfall, snowing, lightning, landslide, volcanic effects, biological effects, external fires (forest fires, nearby industrial facilities fires and aircraft crash fires), etc.

• Article 8: Prevention of damage due to fire The nuclear reactor facilities shall be designed such that their safety will not be impaired by fire considering individual protective measures for preventing, detecting and extinguishing fire, and mitigating its effects. These protective measures shall also be designed such as not to impair the required functions of SSCs with safety functions as a result of their failure or malfunction.

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The Regulatory Guide of the NRA Ordinance that specifies the standard of the location, structure and equipment of the power reactor facilities (iii) [JPN-06]

provides supplemental explanations on the standard (i) [JPN-04].

3.7.1.2 Detailed Technical Requirements for Nuclear Reactor Facilities In the ordinance that specifies the technical standard of the power reactor facilities (ii) [JPN-05],

two articles (Article 7 and Article 11) are required on the detailed design of the fire protection similar to the two articles in (i) [JPN-04].

The Regulatory Guide of the NRA Ordinance that specifies the technical standard of the power reactor facilities (iv) [JPN-07] provides supplemental explanations on the technical standard (ii)

[JPN-05].

3.7.2 New Reactors In Japan, there are currently no regulatory requirements that stipulate application to new reactors.

3.7.3 Supplementary Information Fire Protection Review Standard for Power Reactor Facilities The Fire Protection Review Standard [JPN-08] specifies the matters to be considered regarding details of fire protection measures for light water nuclear power reactor facilities (hereinafter referred to as nuclear reactor facilities) based on the fire protection design policy defined in Article 8 of the standard (i).

1.1 General Requirements (1) Appropriate fire protection measures shall be taken to protect the SSCs equipped with safety functions that are installed in the fire areas and fire zones of the nuclear reactor facilities through fire prevention, fire detection and suppression, and mitigation of the effects of fires, based on whether the fire areas and fire zones that have the SSCs to be protected fall into fire area/zone category [1] or [2] defined below:

[1] Fire areas and fire zones in which SSCs equipped with safety functions for hot and/or cold shutdowns of the nuclear reactor and for maintaining the reactor shut down are installed,

[2] Fire areas in which SSCs equipped with functions to store or contain radioactive materials are installed.

(2) A Fire Protection Plan shall be formulated that includes detailed description of the fire protection measures to be taken and of the procedures, equipment and staffing required to implement the fire protection measures.

1.2 Fire prevention (1) Protection of leakage, ventilation and explosion resistance for combustible or inflammable substances (2) Protection for hydrogen produced by radiolysis (3) Use of incombustible or fire-retardant substances (4) Use of fire-retardant electric cables (should be tested by IEEE383/IEEE1202, UL1581, etc.)

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1.3 Fire detection and suppression (1) Design to be able to conduct early fire detection and suppression

• Fire area for installation of fire detectors

• Combination of different types of detectors or devices with equivalent capability

• Ensuring electricity in case of loss of electrical source

• Design to be able to monitor fires in the main control room

• Installation of fixed automatic or manual suppression systems in a fire areas/zones with SSCs to ensure the safety function

• Redundancy or diversity of water supply and suppression pumps

• Design of fire suppression systems with independency for the separation of redundant trains

• Design of water-based fire suppression system to ensure the largest expected flow rate for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (2) Design to maintain the fire detection and suppression function, even if natural phenomena such as earthquake, freezing, flood, high wind (typhoon), etc. have occurred (3) Design not to lose the safety function by the malfunction or mishandling (4) Influence to the safety function by flooding should be evaluated by the guide of evaluation for internal flooding 1.4 Fire mitigation (1) Design for implementing the fire mitigating measures according to the safety significance for SSCs

• Separation of fire area having SSCs with function for hot and cold shutdown by fire barriers of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> resistance rating

• Design for preventing the intra- and inter-fire propagation of the fire zones for the separation of redundant SSCs and related non-safety cables (2) Design to be able to maintain hot and cold shutdown without losing simultaneously the function of each multiple redundant train, in case that the safety protection system and the reactor shutdown system are required to act in case of fire (3) It should be evaluated by the fire hazard analysis that hot and old shutdown can be maintained.

1.5 Confirmation of separation of redundant trains (1) Separation of redundant SSCs and cables for fire protection by a fire barrier with a 3-hours rating should be achieved.

(2) Separation of redundant SSCs and cables for fire protection by a horizontal distance of more than 6.1 m (20 ft) with no fire hazard should be achieved. In addition, fire detectors and an automatic fire suppression system should be installed in the fire area.

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(3) Enclosure of redundant SSCs and cables for fire protection by means of fire barriers with a 1-hour rating should be achieved. In addition, fire detectors and an automatic fire suppression system should be installed in the fire area.

Guide for Evaluating the Effects of Internal Fires at Nuclear Power Stations For fire protection of safety related SSCs installed in fire areas and fire zones in power generation nuclear reactor facilities, the fire protection requirements of Article 8 of the standard (i) [JPN-04] and Article 11 of the standard (ii) [JPN-05] should be properly implemented in the design.

The Evaluation Guide [JPN-09] presents examples of procedures for evaluations of the effects of internal fires conducted to confirm that the fire protection measures taken based on these requirements to ensure that the safety functions relating to hot shutdowns and cold shutdowns of nuclear reactors (hereinafter collectively referred to as safe shutdowns) will function correctly in the event of a fire in a nuclear reactor facility. In addition, this Evaluation Guide will be used by examiners as a source of reference when they judge the adequacy of evaluations of the effects of internal fires. With regard to methods for evaluating the effects of fires, it is considered necessary, in view of their present technical standard, to continuously review them in the future taking into consideration the experience of their application.

Guide for Evaluating the Effects of External Fires at Nuclear Power Stations The fire protection requirements of Article 6 of the standard (i) [JPN-04] and Article 7 of the standard (ii) [JPN-05] should be properly implemented in the design. For equipment of nuclear power stations (power stations) that is important for safety, measures have been taken to prevent it from being damaged, such as adopting sufficient margin in the design, adopting configurations with redundancy and diversity and performing appropriate maintenance.

The Evaluation Guide [JPN-09] is a guide for evaluations to verify that, even if a forest fire approaches a nuclear power station, the nuclear reactor facilities will be unaffected, as part of efforts to increase the safety against fires that occur off the premises of power stations.

[JPN-01] Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors, http://www.japaneselawtranslation.go.jp/law/detail/?printID=&ft=2&re=02&dn=1&y o=Act+on+the+Regulation+of+Nuclear+Source+Material%2C+Nuclear+Fuel+Mat erial+and+Reactors&x=8&y=20&ky=&page=1&vm=02.

[JPN-02] Fire Service Act, http://www.japaneselawtranslation.go.jp/law/detail/?printID=&ft=2&re=02&dn=1&y o=fire+service+act&x=36&y=21&ky=&page=1&vm=02.

[JPN-03] Building Standards Act, in Japanese only, http://elaws.e-gov.go.jp/search/elawsSearch/elaws_search/lsg0500/detail?lawId=325AC000000 0201_20180925_430AC0000000067&openerCode=1.

[JPN-04] NRA Ordinance on Standards for the Location, Structures and Equipment of Commercial Power Reactors, http://nsr-portal/kyoyu0101/SARIS_Attachment/L03The%20NRA%20Ordinance%20on%20 Standards%20for%20the_.pdf.

[JPN-05] NRA Ordinance on Technical Standards for Commercial Power Reactors Facilities, 49

http://nsr-portal/kyoyu0101/SARIS_Attachment/L02The%20NRA%20Ordinance%20on%20 Technical%20Standards.pdf.

[JPN-06] Regulatory Guide of the NRA Ordinance on Standards for the Location, Structure, and Equipment of Commercial Power Reactors, http://nsr-portal/kyoyu0101/SARIS_Attachment/L04The%20Regulatory%20Guide%20of%2 0the%20NRA%20Ordinance_.pdf.

[JPN-07] Regulatory Guide of the NRA Ordinance on Technical Standards for Commercial Power Reactors Facilities, http://nsr-portal/kyoyu0101/SARIS_Attachment/L05(For%20SARIS%20module%20Module 9-5%20QID63)Article%2035%20of%20the_.pdf.

[JPN-08] Fire Protection Review Standard for Power Reactor Facilities, in Japanese only, http://warp.da.ndl.go.jp/info:ndljp/pid/8729504/www.nsr.go.jp/nra/kettei/data/2013 0628_jitsuyounaiki03.pdf.

[JPN-09] Guide for Evaluating the Effects of Internal Fires at Nuclear Power Stations, in Japanese only, http://warp.da.ndl.go.jp/info:ndljp/pid/8729504/www.nsr.go.jp/nra/kettei/data/2013 0628_jitsuyounaikasai.pdf.

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3.8 Korea Koreas regulatory framework (Figure 6) that the nuclear regulatory body (RB 2) has established 1F for fire protection in nuclear power plant consists of two enforcements, two supporting notices, and regulatory guidelines: Article 20, Details of Periodic Safety Review to Enforcement, Enforcement Regulation of the Nuclear Safety Act [KOR-01]; Article 14, protection against fire protection, etc. [KOR-02]and Article 59, Fire Protection Program to Enforcement, Regulations on Technical Standards for Nuclear Reactor Facilities, Etc. [KOR-03]; Notice 2015-11, Technical Standards for Fire Hazard Analysis [KOR-04]; Notice 2015-12, Establishment and Implementation of Fire Protection Program [KOR-05]. Except regulatory guidelines, these NPP fire protection regulations promulgated by the RB are categorized as regulations. The use of the RBs regulatory guides for nuclear reactors and industrial codes/standards, which are referred to in those regulatory guides, is acceptable in complying those regulations and achieving license.

The compliance with those regulations is required to not only existing reactors but also new reactors on the basis of the plants applicability of specific regulations.

Figure 6 Overview of the Regulatory Framework for Fire Protection in Korea 3.8.1 Existing Reactors All existing nuclear power plants were adopting plant-specific licensing base for fire protection which was provided from exporting countries such as U.S. NRC SRP CMEB 9.5-1 [NRC-34]

and CAN/CSA N293 [CSA-12] in accordance with Article 14,protection against fire protection, etc. to Enforcement, Regulations on Technical Standards for Nuclear Reactor Facilities, Etc.

[KOR-02] the RB considers acceptable for use in implementing those plant-specific licensing 2 Koreas Regulatory Body: Koreas RB for nuclear safety consists of two organization, Nuclear Safety and Security Commission (NSSC) which is governmental agency, Korea Institute of Nuclear Safety (KINS) which is the governmental regulatory organization.

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bases for fire protection as compliance with Article 14,protection against fire protection, etc. to Enforcement, Regulations on Technical Standards for Nuclear Reactor Facilities, Etc.[KOR-02].

The RB requires that licensee should provide results of fire hazard analysis every ten years after their submittal of fire hazard analysis for permit of operating license including post-fire safe shutdown analysis in accordance with Notice which is effective at the date of evaluation. Most licensees are requested to re-establish their fire protection plan in accordance with recently issued Notices like Notice 2015-12 [KOR-05] regardless of plant-specific licensing base.

Periodic Safety Reviews should be required every ten years for licensee to perform reanalysis of the aging effect of fire protection system, to reanalyze fire and explosion hazard in NPP, and to reflect the operational experience and recommendations from recent research related to fire protection in accordance with Article 20, Details of Periodic Safety Review to Enforcement, Enforcement Regulation of the Nuclear Safety Act [KOR-01]. In the process of analyzing fire and explosion hazard, licensee should re-evaluate fire hazard analysis based on latest Notice such as Notice 2015-11. Regardless of plant-specific licensing base for fire protection, licensee should be required to comply with the paragraph for safe shutdown analysis specified in currently effective and latest Notice such as Notice 2015-11 [KOR-04]. Fire Probabilistic Safety Assessment (PSA) is also required in accordance with Article 20, Details of Periodic Safety Review to Enforcement, Enforcement Regulation of the Nuclear Safety Act [KOR-01]. The methodology of Fire PSA should be adopted NUREG/CR-6850 (ERPI TR-105928) [NRC-08]

based on the effective date of evaluation.

KINS Regulatory Guide 10.6 [KOR-06] or the applicable U.S. NRC Regulatory Guide [NRC-09]

and CSA standard [CSA-12] apply to all facilities where these guidance documents are referenced as a license condition by the regulatory body.

For facilities obtained the construction permit (CP) prior to the current edition of guidance, the deviations from previous edition of guidance are enhanced through Licensees Periodic Safety Review (PSR) which is submitted every 10 years after obtaining the operating license (OL). The RB does not require retrofits to facilities constructed prior to the effective edition of guidance but enhancements are recommended by the regulatory body based on operational experiences and recommendations from research related with fire protection. However, regardless of referenced edition of guidance, all deviations apart from fire safety shutdown criteria shall be retroactively enhanced based on currently effective edition of guidance. In case that compliance with the deterministic requirements is not possible due to existing condition, the regulatory body can accept the implementation of the equivalent level of spatial or physical separation, fire detection and suppression as a fire safety shutdown criterion.

3.8.1.1 Overview

1. Article 20,Details of Periodic Safety Review, Enforcement, Enforcement Regulation of the Nuclear Safety Act [KOR-01]

Fire PSA, internal/external fire risk assessment, degradation due to aging, and operational experience/research findings should be re-evaluated through every 10 years Periodic Safety Review process, which is required for all licensees to conduct and submit for regulatory bodys approval. For licensees being not required to submit Fire PSA in the construction permit/operating license phase, Fire PSA should be conducted on the basis of KINS/GE-07, Safety Review Guides for PSR [KOR-07] and submitted on the upcoming PSR after November 2014.

2. Article 14 Protection against Fire Protection, etc., Enforcement, Regulations on Technical Standards for Nuclear Reactor Facilities, Etc. [KOR-02]

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This regulation is based on the concept of defense-in-depth, (1) prevent fires from starting, (2) detect rapidly, suppress and extinguish fire promptly, and (3) protection SSCs important to safety, so that a fire which is not extinguished will not prevent the safe shutdown abilities.

Three categories of fire protection apply to all existing NPP as regulation. This regulation also require that licensee should provide fire hazard analysis and conduct reanalysis in accordance with recently issued FHA Notice such as Notice 2015-11 [KOR-04].

3. Article 59 Fire Protection Program, Enforcement, Regulations on Technical Standards for Nuclear Reactor Facilities, Etc. [KOR-03]

Nuclear Safety Enforcement Regulation on Technical Standards for Nuclear Reactor Facilities, Etc. Article 59(Fire Protection Program) requires that all licensee implement and maintain comprehensive fire protection program (FPP) which consists of fire prevention activities, firefighting and response strategy including initial fire brigade, and maintenance of fire protection features and fire safety shutdown condition.

4. NSSC Notice 2015-11 Technical Standards for Fire Hazard Analysis [KOR-04]

Fire Hazard Analysis (FHA) is also required to access fire hazard, fire protection features, and safety shutdown ability based on Nuclear Safety and Security Commission (NSSC)

Notice 2015-11 Technical Standards for Fire Hazard Analysis KOR-04]. Every ten years, every in-situ and transient fire hazard should be identified including updates of plant changes for each fire area (like fire compartment) and fire protection features should be assessed based on applicable current codes and standards. Fire safety shutdown ability shall be obtained and maintained based on current safety shutdown criteria which is mentioned in NSSC Notice 2015-11 KOR-04]. Any deviations from current applicable codes and standards related to fire protection features could be enhanced by the RBs recommendation but not mandatory. The RB does not strictly allow any deviations from fire safety shutdown criteria but can accept the implementation of the equivalent level of spatial or physical separation, fire detection and suppression as fire safety shutdown criteria. MSO (multiple spurious operation) should be included and submitted on the upcoming PSR after December 2015.

5. NSSC Notice 2015-12 Establishment and Implementation of Fire Protection Program

[KOR-05]

NSSC Notice 2015-12 requires that licensees of all facilities establish organization and entitlement, fire prevention activities, fire response strategy, fire safety shutdown procedure, and fire brigade drill and education program.

6. KINS Regulatory Guide 10.6 Fire Protection for Nuclear Power Plants [KOR-06]

KINS Regulatory Guide 10.6 [KOR-06] presents the detailed guidelines for nuclear power plants related to fire protection program, fire prevention, fire detection and suppression, building design and passive features, safe shutdown capability, fire protection for areas important to safety, protection of special fire hazards exposing areas important to safety, etc.

3.8.2 New Reactors The acts and standards related to the nuclear regulation and the fire protection guides mentioned above are applicable for the existing reactors as well as for new reactors. The chapter 8 fire protection for new reactor in the KINS regulatory guide 10.6 [KOR-06] presents enhanced standard and guide such as NFPA 804 [NFPA-15].

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3.8.3 Supplementary Information Regulatory Requirements Regulation Level

[KOR-01] Nuclear Laws of the Republic of Korea: Enforcement Regulation of the Nuclear Safety Act, Article 20: Details of Periodic Safety Review, Korea Institute of Nuclear Safety (KINS), Enacted by Regulation of the Prime Minister No. 2, Nov.

11, 2011 (Entered into force, Nov. 11, 2011), Amended by Presidential Decree No. 24689, August 16, 2013, http://www.nssc.go.kr/nssc/en/nci/elif/Enforcement_Regulation_of_the_Nuclear_S afety_Act.pdf.

[KOR-02] Nuclear Laws of the Republic of Korea: Regulation on Technical Standards for Nuclear Reactor Facilities, Etc., Article 14: Protection against Fire Protection, Etc.,

Korea Institute of Nuclear Safety (KINS), Enacted by Ordinance of the Ministry of Education, Science and Technology No. 16, Apr. 18, 2000, Amended by Ordinance of the Ministry of Education, Science and Technology No. 31, Jul. 28, 2001, Ordinance of the Ministry of Education, Science and Technology No. 92, Jul. 19, 2006, Ordinance of the Ministry of Education, Science and Technology No. 1, Mar. 4, 2008Regulation of the Nuclear Safety and Security Commission No. 3, Nov. 11, 2011, http://www.nssc.go.kr/nssc/en/nci/elif/Regulations_on_Technical_Standards_for_

Nuclear_Reactor_Facilities,ETC.pdf.

[KOR-03] Nuclear Laws of the Republic of Korea: Regulation on Technical Standards for Nuclear Reactor Facilities, Etc., Article 59: Fire Protection Program, Korea Institute of Nuclear Safety (KINS), 2008, Regulation of the Nuclear Safety and Security Commission No. 3, Nov. 11, 2011.

http://www.nssc.go.kr/nssc/en/nci/elif/Regulations_on_Technical_Standards_for_

Nuclear_Reactor_Facilities,ETC.pdf.

[KOR-04] Nuclear Safety and Security Commission (NSSC): Notice 2015-11: Technical Standards for Fire Hazard Analysis, in Korean only, http://www.law.go.kr/admRulLsInfoP.do?chrClsCd=010202&admRulSeq=210000 0033523.

[KOR-05] Nuclear Safety and Security Commission (NSSC) Notice 2015-12: Regulation on Establishment and Implementation of Fire Protection Program, 2015, in Korean only, http://www.law.go.kr/admRulLsInfoP.do?chrClsCd=010202&admRulSeq=210000 0033524.

Guidance Level

[KOR-06] Korea Institute of Nuclear Safety (KINS): Fire Protection for Nuclear Power Plants, Regulatory Guide 10.6, in Korean only, http://www.kins.re.kr/nsic.do?menu_item=technologyStatus.

[KOR-07] Korea Institute of Nuclear Safety (KINS): KINS/GE-07: Safety Review Guides for PSR.

[NRC-09] United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research: Fire Protection for Nuclear power Plants, Regulatory Guide 54

1.189, Rev. 2, Washington, DC, USA, October 2009, https://www.nrc.gov/docs/ML0925/ML092580550.pdf.

[NRC-34] United States Nuclear Regulatory Commission (NRC): Fire Protection for NPPs, NUREG-0800 BTP CMEB 9.5-1 Standard Review Plan, Rev. 3, Washington, DC, USA, July 1981, https://www.nrc.gov/docs/ML0706/ML070660454.pdf.

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3.9 The Netherlands 3.9.1 Existing Reactors All nuclear facilities in the Netherlands, including the Borssele NPP, operate under license, authorized after a safety assessment has been carried out. The license is granted by the regulatory body under the Nuclear Energy Act [NED-01], called Kew.

An important step in 2015 was an update of the Kew with legally establishing the new Authority for Nuclear Safety and Radiation Protection, ANVS, as an independent administrative authority (see https://english.autoriteitnvs.nl/).

The Kew is a framework law, which sets out the basic rules on the application of nuclear technology and materials, makes provision for radiation protection, designates the competent authorities and outlines their responsibilities. More detailed legislation is provided by associated Decrees and Ordinances. In the Netherlands the (modified) International Atomic Energy Agency (IAEA) requirements and guides are the basis of the regulation of the existing NPP, including the Western European Nuclear Regulators Association (WENRA) Reference Levels.

International Atomic Energy Agency (IAEA):

• Fire Safety in the Operation of Nuclear Power Plants Safety Guide, IAEA Safety Standards Series NS-G-2.1, 2000 [IAEA-00b]:

In the Netherlands this IAEA SG has been adopted to: NVR NS-G-2.1 Brandveiligheid in de bedrijfsvoering van kernenergiecentrales[NED-02];

• Protection against Internal Fires and Explosions in the Design of Nuclear Power Plants Safety Guide, IAEA Safety Standards Series, NS-G-1.7, 2004 [IAEA-04]:

In the Netherlands this IAEA SG is adopted to: NVR NS-G-1.7 Bescherming tegen interne branden en explosies in het ontwerp van kernenergiecentrales [NED-03];

• IAEA Safety Guide Safety Standard Series No. NS-G-1.11, Protection Against Internal Hazards other than Fires and Explosions in the Design of NPPs [IAEA-04a]:

In the Netherlands this IAEA SG is adopted to: NPPs NVR NS-G-1.11 Bescherming tegen interne gevaren anders dan branden en explosies in het ontwerp van kernenergiecentrales [NED-04];

• Performance of a Fire Hazard Analysis, IAEA Safety Standards Series No. 50-P-9, Vienna, 1995 [IAEA-95];

• Preparation of a Fire Hazard Analysis, IAEA Safety Reports Series No. 8, Vienna, 1998 [IAEA-98];

• Use of operational experience in fire safety assessment of nuclear power plants, IAEA-TECDOC-1134, Vienna, January 2010 [IAEA. [IAEA-00c].

Under Article 41 of the Kew, the local authorities also have the responsibility for making regional/local contingency plans for emergencies. Firefighting service, police and health services will be involved. These include: NVR-NS-R1 [NED-05] and NVR-SSG-2 [NED-06]

stating that a full range of events must be postulated in order to ensure that all credible events with potential for serious consequences and significant probability have been anticipated and can be accommodated by the design base of the plant. For the safety analysis of the Borssele NPP, the postulated initiating events have been defined in the following categories according to their entrance probability.

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The Safety Regions Act (Wet Veiligheidsregio's, WVR) [NED-07] is a Dutch law that came into force on October 1, 2010. This act replaces the Fire Service Act (Brandweerwet, 1985) and provides the legal frame for firefighting.

The Decree Safety Regions (Besluit veiligheidsregios) [NED-08], of June 24, 2010 lays down rules on the organization and functions of the Safety Regions (Veiligheidsregios) and municipal fire departments.

The Building Decree (Bouwbesluit, 2012 ) [NED-09] came into force on April 1, 2012. This is the legal frame for fire and smoke resistance of walls, floors and ceilings, protected escape routes and fire compartments of the buildings.

3.9.2 New Reactors In 2015 the new Dutch Safety Guidelines have been completed for (new) water cooled reactors, with Dutch acronym VOBK. The VOBK consists of an (extensive) introductory part and a technical part, the Dutch Safety Requirements, the DSR [NED-10]. The DSR is based on the IAEA Safety Fundamentals, several Safety Requirements guides and some Safety Guides, safety objectives for new NPPs published by WENRA. An annex to the DSR is dedicated to Research Reactors and describes application of the DSR with a graded approach. The DSR takes into account the latest (post-Fukushima) insights and is in line with the European Directive on Nuclear Safety 22 and the objectives of the Vienna Declaration on Nuclear Safety (VNDS)

[NED-11].

3.9.3 Supplementary Information Further guidance can be revealed from harmonized European non-nuclear fire protection standards (e.g., NEN-EN for technical requirements regarding fire detection and extinguishing equipment). Consideration can also be given to existing state-of-the-art international nuclear fire safety standards for operating NPPs taking also into account the difficulties of older plants meeting formally more recent requirements, such as the German nuclear fire protection standard KTA 2101, Part 1-3 [KTA-00], [KTA-00a], [KTA-00b], representing rules written for recently designed plants as well as for plants designed to former standards.

[NED-01] Nuclear Energy Act (Kernenergiewet or Kew), 1963, https://wetten.overheid.nl/BWBR0002402/2018-10-16.

[NED-02] NVR NS-G-2.1: Brandveiligheid in de bedrijfsvoering van kernenergiecentrales.

[NED-03] NVR NS-G-1.7: Bescherming tegen interne branden en explosies in het ontwerp van kernenergiecentrales.

[NED-04] NVR NS-G-1.11: Bescherming tegen interne gevaren anders dan branden en explosies in het ontwerp van kernenergiecentrales.

[NED-05] NVR-NS-R1 (Safety Requirements for Nuclear Power Plant Design),

https://www.iaea.org/publications/6002/safety-of-nuclear-power-plants-design.

[NED-06] NVR-SSG-2 (Deterministic Safety Analysis),

https://www-pub.iaea.org/MTCD/publications/PDF/Pub1428_web.pdf.

[NED-07] The Safety Regions Act (Wet Veiligheidsregio's, WVR), October 1, 2010, http://wetten.overheid.nl/BWBR0027466/geldigheidsdatum_11-10-2014.

[NED-08] The Decree Safety Regions (Besluit veiligheidsregios, une 24, 2010, http://wetten.overheid.nl/BWBR0027844/geldigheidsdatum_11-10-2014.

[NED-09] The Building Decree (Bouwbesluit, 2012), April 1, 2012, http://wetten.overheid.nl/BWBR0030461/geldigheidsdatum_11-10-2014.

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[NED-10] Dutch Safety Requirements for Nuclear Reactors: Fundamental Safety Requirements, 19.03.2015, https://www.oecd-nea.org/nsd/docs/2015/csni-r2015-15.pdf.

[NED-11] Vienna Declaration on Nuclear Safety (VNDS),

https://english.autoriteitnvs.nl/topics/guidelines-on-the-safe-design-and-operation-of-nuclear-reactors.

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3.10 Spain 3.10.1 Existing Reactors The Nuclear Safety Councils Instruction IS-30 [CSN-13] on the requirements of the fire protection program at nuclear power plants is the national regulation that requires nuclear power station license holders to setup and maintain a fire protection program at their facilities. This fire protection program includes all the features regarding fire protection (prevention, detection, extinction, firefighting) and analysis performed to ensure that the safe shutdown is achieved and maintained in case of any postulated fire event at any fire area of the facility, including the main control room. This capacity includes the adequate confinement of radioactive materials so that the likelihood of offsite releases of radioactive materials is minimized.

For some time, the Nuclear Safety Council (CSN) has required the nuclear power plant licensees to implement a fire protection program in keeping with the requirements demanded of U.S. plants and with the licensing conditions for fire protection applied to each plant in particular.

Pursuant to the provisions of article 8.3 of the Regulation on Nuclear and Radioactive Facilities, (Royal Decree 1836/1999, of December 3rd [SPN-03], modified by Royal Decree 35/2008, of January 18th [SPN-04]), and further to the need to incorporate these requirements into the Spanish legal framework, Nuclear Safety Council Instruction IS-30 dealing with the requirements of the fire protection program at nuclear power plants (Official State Gazette BOE No. 40 of February 16th, 2011) was approved on January 19th, 2011.

In drawing up this Council Instruction, consideration was given to the work performed by the Western European Nuclear Regulators Association (WENRA) in order to harmonize the regulations of the different countries. As a result of this effort, a set of common requirements known as <<reference levels>> was established, these to be reflected in the national standards.

Specifically, in its chapter S (Protection against internal fires) [WENRA-08] the WENRA reference levels document sets out the basic applicable requirements which, in the terminology traditionally used within the Spanish documentary and legal framework, are known as Fire Protection at nuclear power plants.

In order to give consistency to the standards development process undertaken by the CSN as a result of this harmonization effort, it was considered necessary to draw up a Council Instruction contemplating the aforementioned requirements, this giving rise to approval of the said Instruction IS-30, of January 19th, 2011 Subsequently, in view of the experience gleaned from application, the need to regulate the different specific characteristics of both the design and the original licensing basis of the system for fire protection of each of the different Spanish nuclear power plants and the evolution of the fire protection regulations, revision 1 of Instruction IS-30 of February 21st, 2013 was approved. Finally, the current version of the IS-30 was issued in order to clarify and facilitate the practical application of the term <<exemption>>, splitting the term coined in revision 1 into two new terms, exemption and equivalent measures, which fit perfectly into the regulatory framework governing nuclear safety and radiological protection. Revision 2 of the Instruction, approved by the Council on November 16th, 2016, came into force the day after its publication in the Official State Gazette (Wednesday November 30th, 2016) [SPN-05].

3.10.1.1 Purpose and scope of application The purpose of this Council Instruction is requiring nuclear power station license holders to implement a fire protection program and defining the criteria that must be fulfilled by such program. This Instruction shall apply to the licensees of all Spanish nuclear power plants with an operating license.

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3.10.1.2 The Nuclear Safety Councils criteria for fire protection at nuclear power plants.

The fire protection objectives must be fulfilled by any license holder under the scope of the Instruction under the principle of defense-in-depth in fire protection, namely implementing measures to prevent a fire before it starts, to detect, control and extinguish it as soon as possible in case it occurs, and to prevent the spread thereof to other areas that might affect safety.

On the other hand, by means of confinement in fire areas, it must be ensured that a fire that cannot be extinguished will not damage at least one of the redundant safe shutdown trains such that the power plant may achieve and maintain such safe shutdown and the likelihood of offsite radioactive releases is minimized.

3.10.1.3 Safe shutdown capacity A fire risk analysis that proves that fire safety objectives are fulfilled, design bases are complied with, active and passive fire protection systems have been properly designed and administrative controls have been properly implemented must be conducted and kept up to date. This analysis must prove that the possible consequences and effects of both the intentional and spurious actuation of fire extinction systems has been taken into consideration.

On the other hand, a safe shutdown analysis must demonstrate, from the identification of the redundant safe shutdown trains considered in the facility that, under a postulated fire in any fire area of the plant, damages to systems are limited so that one train of the systems needed to achieve and maintain safe shutdown conditions from the control room or from the panel for remote shutdown in case of a fire is undamaged by the fire; and the systems needed to achieve and maintain cold shutdown from the control room or from the panel for remote shutdown in case of a fire can be repaired within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the start of the fire.

For fire areas where all redundant trains of any system necessary to achieve safe shutdown in case of fire maybe affected by a fire an alternative or dedicated shutdown capacity independent from the cables, systems and components located in this area, or even control room abandonment is required, if the deterministic post-fire shutdown analysis concludes so for the specific configuration of these areas. The use of operator manual actions in case of fire is an acceptable means of compliance requiring a case-by-case assessment by the CSN.

There must be an alternative or dedicated shutdown capability, independent from the cables, systems and components located in the main control room. The analysis of deterministic safe shutdown capacity in case of fire in this area shall be carried out in accordance with the methodology provided in NEI 00-01 [NEI-05].

A valid alternative to meet these requirements is to follow a risk-informed, performance-based methodology previously accepted by the CSN.

3.10.1.4 Additional requirements The Instruction also establishes additional requirements to fire protection systems in areas important to safety at the facility as well as to the quality assurance program applicable to their design, acquisition, assembly, testing and the administrative controls.

Procedures must also be established to control and minimize the amount of combustible material and ignition sources that might affect equipment important to safety.

Effective firefighting capability is also under the scope of the Instruction IS-30, which establishes specific requirements onto fire brigade organization and co-ordination, composition, duties, physical conditions, training and available resources.

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3.10.1.5 Regulatory supervision of the fire protection program at nuclear power plants in Spain The former set of analysis, procedures and documents constitute the Fire Protection Program of the facility and any change in them that could impair the capacity in fulfilling the objectives of fire protection must be approved by CSN. With this aim, a Complementary Technical Instruction was issued by the CSN last June 2018 to all holders of an operation license in Spain to set the conditions for the regulatory control of this Fire Protection Program and its changes caused by either physical or document modifications.

3.10.2 New Reactors The Instruction IS-30 of CSN [SPN-05] states, in its Sole Additional Provision, that:

In the case of new nuclear power plants, it shall be considered from the very early stages of design that among the fire protection requirements for being capable of achieving and maintaining safe shutdown and minimizing the likelihood of off-site radioactive releases, the requirements of section 1) of Article 3.2.5 will not be taken into account such that, outside the containment building, the redundant safe shutdown trains, including their associated circuits, must be located in different fire areas. In addition, their design shall minimize or eliminate the use of alternative or dedicated shutdown systems, except for the case of the main control room.

Likewise, the execution of operator manual actions shall be avoided in case of a fire, and the use of fire-resistant coating in electrical raceways shall be minimized.

That means, for new design facilities, requirements for the separation of redundant trains of systems necessary to achieve and maintain the safe shutdown of the facility in case of fire must be achieved via fire area compartmentation and therefore no alternative configurations will be considered as equivalent of such separation required in the terms that were accepted for the existing facilities, also to ensure the alternative shutdown capacity.

3.10.3 Supplementary Information The Law 15/1980 [SPN-02], creating the Nuclear Safety Council, explicitly states the faculty of the CSN at proposing and reviewing regulations, as well as elaborating and approving different kinds of mandatory rules related to nuclear and radioactive installations and to the activities associated with nuclear safety and radiological protection. In that sense, Law 15/1980 assigns to the CSN the function of proposing the necessary regulations on nuclear safety and radiation protection, as well as their reviews, for final approval. Furthermore, this article states that the CSN has the legal capacity to issue Instructions of the CSN (legally binding), Circulars, and Safety Guides on technical issues concerning nuclear installations and radioactive facilities and the activities related to the nuclear safety and radiation protection (under these terms, transport, emergency and security are also included). Therefore, the CSN elaborates rules (Instructions of the CSN and Safety Guides) which will become part of the Regulatory Framework on the subject under the scope of the rule.

The Board of the CSN set up a Regulation Commission in charge of promoting, supervising and coordinating the activity related to regulation development. The members of this commission consist of two Commissioners (acting as President and Vice-president of the Commission) and other members from both Technical Directorates and Secretariat. Besides CSN membership, the competent Ministry is represented in the Regulation Commission as well so that this Commission serves as a channel for coordinating and monitoring the activities regarding regulations and guides.

The CSN provides, within this legal framework, processes for establishing or adopting, promoting and amending regulations and guides. These processes involve compulsory 61

consultation with interested parties and general public in the development of regulations and guides, taking into account internationally agreed standards and the feedback of relevant experience. Moreover, technological advances, research and development work, relevant operational lessons learned, and institutional knowledge are valuable tools in reviewing regulations and guides. The consideration of different kinds of foreign regulations (IAEA, OECD/NEA, Europe, etc.) is considered a strength, and it is important to mention that the CSN Creation Law includes provisions focused on foreign relationships that contains international agreements on the CSN's responsibilities.

In Spain, the regulatory framework establishes the principles, requirements and associated criteria for safety upon which the regulatory judgements, decisions and actions are based. This framework has a hierarchical structure, as it is shown on next figure, starting with International Treaties (Conventions), and following top to bottom with Laws, mandatory regulations and Instructions and ending with Guides that contain acceptable technical approaches to comply with the regulations. Moreover, the competent Ministry and the CSN establish Limits and Conditions applicable to the license granted, and Complementary Technical Instructions for each installation, as a way to establish technical requirements about specific matters not included in other regulations (Figure 7).

Figure 7 Hierarchical Pyramid of Nuclear Regulation in Spain There are two main laws serving as the framework for the regulatory requirements and conditions:

• Law 25/1964, of 29th April, on Nuclear Energy [SPN-01],

• Law 15/1980, of 22nd April, creating the Nuclear Safety Council. [SPN-02].

The following Royal Decrees are also a relevant part of the Spanish nuclear regulatory framework:

• Royal Decree 1836/1999, of December 3rd, approving the Regulation on Nuclear and Radioactive Facilities (under review, BSS EU Directive) [SPN-03];

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• Royal Decree 783/2001, of July 6th, which approves the Regulation on Sanitary Protection against Ionising Radiations (under review, BSS EU Directive);

• Royal Decree 146/2004, of June 25th, approving the basic Nuclear Emergency Plan.

• Royal Decree 1440/2010, of November 5th, approving the Statute of the Nuclear Safety Council.

• Royal Decree 1400/2018, of November 23rd, approving the Regulation on nuclear safety in nuclear facilities [SPN-04].

So far, there are 42 Instructions of the CSN; these legally binding documents develop Laws and Royal Decrees on nuclear safety, radiation protection, waste management, security and transport. In addition, there are more than 70 Safety Guides on different topics such as power reactors and nuclear power plants, fuel cycle installations, radiation protection, environmental radiological control, radioactive installation and devices, radioactive waste management, transport, security, natural radiation, etc. All Instructions of the CSN are available in English at the CSN website (https://www.csn.es/en/normativa-del-csn).

The most relevant regulations are provided in the following:

[SPN-01] The Nuclear Energy Act, Law 25/1964, of April 29th, 1964, https://www.boe.es/buscar/pdf/1964/BOE-A-1964-7544-consolidado.pdf, https://www.csn.es/documents/10182/1369702/Law+251964%2C+of+29th+of+Ap ril%2C+on+Nuclear+Energy (unofficial English version).

[SPN-02] The Law15/1980, of April 22nd, 1980, creating the Nuclear Safety Council, https://www.boe.es/buscar/pdf/1964/BOE-A-1964-7544-consolidado.pdf, https://www.csn.es/documents/10182/1369702/Law+Creating+the+Nuclear+Safet y+Council (in official translation into English).

[SPN-03] The Royal Decree 1836/1999, of December 3rd, 1999, approving Regulation on Nuclear and Radioactive Facilities, https://www.boe.es/buscar/pdf/1999/BOE-A-1999-24924-consolidado.pdf, https://www.csn.es/documents/10182/1369702/Royal+Decree+1836-1999%2C+of+December+3rd%2C+approving+the+Regulation+on+Nuclear+and+

Radioactive+Facilities (unofficial translation into English).

[SPN-04] Royal Decree 1400/2018, of November 23rd, 2018, Regulation on nuclear safety in nuclear facilities, https://www.boe.es/buscar/pdf/2018/BOE-A-2018-16041-consolidado.pdf.

[SPN-05] Instrucción de Seguridad IS-30 del CSN, sobre requisitos del programa de protección contra incendios en centrales nucleares, https://boe.es/boe/dias/2016/11/30/pdfs/BOE-A-2016-11342.pdf, https://www.csn.es/documents/10182/1348817/Instruction%20IS-30,%20Revision%202,%20of%20November%2016th%202016,%20on%20the%2 0requirements%20of%20the%20fire%20protection%20programme%20at%20nucl ear%20power%20plants (unofficial English version).

[SPN-06] The Guia de Seguridad GS 1.19, Programa de Protección contra Incendios en Centrales Nucleares, https://www.csn.es/documents/10182/896572/GS+01-19+Requisitos+del+programa+de+protecci%C3%B3n+contra+incendios+en+cent rales+nucleares.

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3.11 Sweden The regulatory information reflected in this section is current up to 2015.

3.11.1 Existing Reactors Nuclear facilities in Sweden have to adhere both to conventional and specific nuclear fire protection regulations.

3.11.1.1 Nuclear Specific Regulations The Act on Nuclear Activities (lagen (1984:3) om krnteknisk verksamhet, also called krntekniklagen), the Radiation Protection Act (strlskyddslagen (1988:220)) and the Ordinance with instructions for the Swedish Radiation Safety Authority (frordning (2008:452) med instruktion fr Strlskerhetsmyndigheten) have been translated into English. The English versions of the acts and ordinances do not include changes that have been made after 2008.

When the Swedish nuclear reactors were designed there were no nuclear specific national fire protection requirement established. Instead the General Design Criteria 3 of Appendix A to 10 CFR 50 [NRC-01] were used as guidance along with general national building requirements.

Today the nuclear specific regulations including fire protection are primarily:

• SSMFS 2008:1 [SSM-09], and

• SSMFS 2008:17 [SSM-09a].

SSMFS 2008:1 specifies amongst others that fire protection provisions in nuclear facilities must ensure that:

• The capacity of a facilitys barriers and defense-in-depth system to prevent radiological accidents and mitigate the consequences in the event of an accident are analysed using deterministic methods before the facility is constructed or modified and taken into operation.

• The facilities are analysed using probabilistic methods in order to obtain as comprehensive a view as possible of safety.

• The Civil Protection Act (2003:778) [SWD-01] and the Civil Protection Ordinance (2003:789) [SWD-02] are used to specify the emergency preparedness.

• An emergency response plan that should cover all types of accidents for which the facility is designed as well as measures to mitigate the consequences of possible accident sequences which can occur in addition to this (combinations of events should be taken into account, such as fire or sabotage in combination with a radiological accident).

• Instructions of how to minimize combustible materials and ignition sources should be available. Instructions should also be available for inspection, maintenance and testing of fire protection measures (this is stated as a general advice in SSMFS 2011:3 [SWD-03] (not included in the English version of SSMFS 2008:1)).

SSMFS 2008:17 specifies amongst others that fire protection provisions in nuclear facilities have to ensure the following:

• The design of a facility is able to withstand a fire event and its consequences.

• Initiating fire events included in the deterministic safety analysis are divided into a limited number of event classes (based on an analyzed probability with which the event is expected to occur) with specified analysis assumptions and acceptance criteria.

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• The facility in the event of fire (up to and including the event class improbable events) and a simultaneously single fault shall be able to reach a safe state with acceptance criteria according to the event classification.

• Reasonable technical and administrative measures are taken in order to counteract common cause failures in connection with design, manufacturing, installation, start-up, operation and maintenance of safety systems.

• The fire event are analyzed at all operating modes of the reactor. When analyzing fire as an initiating event, an additional fire need not be assumed in the facility.

• A fire that causes all equipment in a fire compartment to fail are assumed to occur, unless a fire hazard analysis can demonstrate that protection measures are sufficient to prevent the failure of redundant items important to safety.

• A fire event are postulated wherever a fire can effect equipment included in safety functions or other equipment used to take the facility to a safe state.

• When analyzing initiating events other than fire, which in turn can result in a fire, a fire should be assumed to occur as a possible consequential failure from the initiating event.

• When analyzing events other than fire, which in turn cannot result in a fire, a fire should nonetheless be assumed to occur no earlier than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the initiating event. This sequence need not be combined with a single failure. This applies to initiating events up to and including the event class unanticipated events, apart from pipe breaks.

• Fire events that can threaten continued activity in the main control room are identified and an established action plan shall be available for dealing with such threats while maintaining reactor safety.

3.11.1.2 Conventional fire protection regulations The following conventional fire protection regulations have to be considered when designing the fire protection:

1. Swedish Environmental Code (1998:808)

2. Civil Protection Act (2003:778)

3. Civil Protection Ordinance (2003:789)

4. Flammable and Explosive Goods Act (2010:1011)

5. Work Environment Act (1977:1160)

6. Work Environment Ordinance (1977:1166)

7. Planning and Building Act (2010:900)

8. Planning and Building Ordinance (2011:338)

The Swedish Environmental Code (1998:808) [SWD-04] for example includes requirement regarding handling with substances that can influence the environment negatively and can be needed to consider when designing confinement of leakage and/or firefighting water.

The Civil Protection Act (2003:778) [SWD-01] includes requirement regarding preparedness for accidents and the manual firefighting resources at a nuclear power plant shall be designed in accordance with this. A general advice for The Civil Protection Act (2003:778), chapter 2, paragraph 2 are given regarding documentation, inspection, maintenance and control of the fire protection measures.

The Flammable and Explosive Goods Act (2010:1011) [SWD-05] includes requirements regarding handling and storage of flammable liquids with a flashpoint not exceeding 100C and flammable gases that can be ignited 20ºC (also including handling of flammable gases in liquid 65

phase such as LPG). Example of fire protection measure that can be needed to fulfil the requirements regarding flammable liquids and gases are fire compartmentation, containment of liquids and gas detection.

The Work Environment Act (1977:1160) [SWD-06] and Ordinance (1977:1166) [SWD-07]

include requirements needed to be considered for fire fighter training, emergency lighting in work places and design of equipment according to ATEX.

The requirement resulting from the Planning and Building Act (2010:900) [SWD-08] and the Planning and Building Ordinance (2011:338) [SWD-09] regarding fire safety are specified in the building regulation, chapter 5 [SWD-11]. This regulation has altered since the Swedish reactors were built. The building regulation of today is therefore not mandatory for Swedish reactors in general. However, it is mandatory when making changes to the building structure, design or when changing the use of the building. The main purpose for the building regulation are evacuation safety.

According to the building requirements the fire protection shall be designed, developed and verified through simplified or analytical design. Simplified design means that the given general advices to the requirements shall be followed. Analytical design shall be used when the solutions according to the general advice are not suitable or if the building is too complex for simplified design to be used. General advice for analytical design is given in BBRAD (BFS 2011:27) [SWD-10].

Some design requirements according to building regulation relevant for nuclear application are summarized under Supplementary material.

3.11.1.3 Combining nuclear specific and conventional requirements on fire safety The purpose of Figure 8 is to explain how the conventional and nuclear specific fire related requirements can be combined to result in a facility adapted fire protection.

Figure 8 How Conventional and Nuclear Specific Fire Related Requirements can be Combined to Result in a Facility Adapted Fire Protection 66

The abbreviations in the figure are described below:

FHA:

A fire hazard analysis according to IAEA NS-G 1.7 [IAEA-04], section 3.24 can include detailed analysis of fire growth and consequence; a quality manual for fire engineering analysis has been produced by NBSG.

SSA/DSA:

Deterministic safety analysis (also called safe shutdown analysis) according to SSMFS 2008:1

[SSM-9], chapter 4, paragraph 1, and SSMFS 2008:17 [SSM-09a], paragraph 14 regarding fire corresponds to IAEA SSG-2 [IAEA-09] regarding fire.

PSA:

Probabilistic fire risk analysis according to SSMFS 2008:1, chapter 4, paragraph 1

- corresponds to PSA according to IAEA NS-G-1.7 [IAEA-04], paragraph 3.27,

- is used to prioritize and evaluate the effect of different fire protection features.

Maintaining fire safety:

Systematic fire protection work according to SRVFS 2004:3 [SWD-13]

- corresponds to the purpose of IAEA NS-G-2.1 [IAEA-00b] 1.4,

- fulfils SSMFS 2008:1 (including 2014:3), chapter 5, paragraph 2 regarding control of combustibles and ignition sources and inspection, maintenance and testing of fire protection measures.

To fulfil the demands of separation of redundant equipment and single failure as well as the demands of safe evacuation the resulting fire protection design can look like the following Figure 9 and Figure 10.

Figure 9 Need for Fire Compartments and Evacuation Routes when all Equipment in a Fire Compartment are Assumed to Fail during Fire - Sweden 67

Figure 10 Need for Fire Compartments and Evacuation Routes when Analytical Fire Hazard Analysis can Demonstrate that Protection Measures are Sufficient to Prevent the Failure of Redundant Items Important to Safety - Sweden 3.11.2 New Reactors There are no nuclear specific requirements established for new reactors in Sweden. The conventional fire related requirements are the same as stated for existing reactors.

3.11.3 Supplementary Information A summary of the building requirements [SWD-11] regarding, fire compartmentation, safe evacuation, automatic fire alarm, suppression system is presented below.

3.11.3.1 Fire Compartmentation Buildings should be divided into fire compartments to the extent that it creates sufficient time for evacuation and restricts the consequences of a fire (with simplified design not more than two floors and areas with specific fire load higher than 1600 MJ/m2 shall be separated). Fire compartment classification may fully or partially be replaced by fire resistant installations.

The design of the fire compartment shall limit the spread of fire and smoke to the adjacent fire compartment over a specified time.

Separating structures in buildings in class Br1 should be designed for at least the fire resistance class given in the table below.

Table 1 Fire Resistance Class Requirement Related to Fire Load - Sweden 68

3.11.3.2 Safe Evacuation Buildings shall be designed to ensure that there is an adequate time for evacuation during a fire.

Adequate time for evacuation means that people who evacuate are not exposed to falling structural elements, high temperatures, high levels of heat radiation, toxic gases or reduced visibility that might impede evacuation to a safe location with sufficient certainty (BFS 2011:26)

[SWD-11].

Spaces where people are present other than occasionally shall be designed with access to at least two independent escape routes. An escape route shall be an exit to a secure location (refers to a space in the open where fire and smoke cannot affect evacuated people) or a space in a building which leads from a fire compartment to such an exit (BFS 2011:26). The walking distance to the nearest escape route or to another fire compartment should not exceed a distance of 45 m in industrial buildings. In a space that is protected by an automatic extinguishing system, the walking distance may be increased by one third.

3.11.3.3 Automatic Fire Alarm Automatic fire alarms shall be installed where this is necessary for the fire protection's design.

The system shall be designed with the necessary properties that have the ability to detect fire reliably and give signals to the functions that depend on the alarm. The system shall be designed with sufficient coverage and shall activate quickly enough to ensure proper function.

The reliability and the ability of automatic fire alarms can be verified in accordance with Section 3 of the Swedish Fire Protection Association's publication, Regler fr automatisk brandlarmsanlggning (automatic fire alarm installations), SBF 110:6 [SWD-12]. The components of a Rule for automatic fire alarm can be verified in accordance with the standard series SS-EN 54 [SIS-06] with properties tailored to suit their intended use.

3.11.3.4 Suppression Systems If an automatic fire suppression system is essential for fire protection, the design shall be such that it has the capability to extinguish or control a fire over the appropriate time with high reliability.

The system shall activate quickly enough and shall be designed with sufficient coverage to ensure proper functionality.

The reliability and capability of automatic water sprinkler systems can be verified in accordance with SS-EN 12845 [SIS-04] and the standard series SS-EN 12259 [SIS-01]. For spaces in occupancy class 5C, the water source should consist of improved, doubled or combined water inlets as specified in 9.6.2 - 9.6.4 in SS-EN 12845.

The reliability and capacity of the water spray and deluge systems can be verified in accordance with SIS-CEN TS 14816 [SIS-09]. Other systems can be verified in accordance with SBF 120 (cf. BFS 2011:26 [SWD-11]).

[SIS-01] Swedish Standards Institute (SIS): Fixed firefighting systems - Components for automatic sprinkler and water spray systems, Series of Standards SS-EN 12259, 2001,.

[SIS-04] Swedish Standards Institute (SIS): Fixed firefighting systems - Automatic sprinkler systems - Design, installation and maintenance, Standard SS-EN 12845:2004, October 15, 2004.

[SIS-06] Swedish Standards Institute (SIS): Fire detection and fire alarm systems, Series of Standards SS-EN 54, 2006, https://www.notifier.se/filer/EN%2054%20STANDARDER.pdf.

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[SIS-09] Swedish Standards Institute (SIS): Fixed firefighting systems - Water spray systems - Design, installation and maintenance, Technical Specification SIS-CEN TS 14816, February 26, 2009.

[SSM-09] Swedish Radiation Safety Authority (SSM): The Swedish Radiation Safety Authoritys Regulations and General Advice concerning Safety in Nuclear Facilities, SSMFS 2008:1, ISSN 2000-0987, January 30, 2009, (latest version in Swedish only),

https://www.stralsakerhetsmyndigheten.se/contentassets/e8ba282f8ef0461ca392 137c9495f466/ssmfs-20081-the-swedish-radiation-safety-authoritys-regulations-concerning-safety-in-nuclear-facilities.

[SSM-09a] Swedish Radiation Safety Authority (SSM): The Swedish Radiation Safety Authoritys Regulations concerning the Design and Construction of Nuclear Power Plants, SSMFS 2008:17, ISSN 2000-0987, January 30, 2009, https://www.stralsakerhetsmyndigheten.se/contentassets/ace5827a9cfc43fc9576f ca14443da32/ssmfs-200817-the-swedish-radiation-safety-authoritys-regulations-concerning-the-design-and-construction-of-nuclear-power-reactors.

[SWD-01] Lag (2003:778) om skydd mot olyckor (Civil Protection Act), 20 November 2003, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/lag-2003778-om-skydd-mot-olyckor_sfs-2003-778.

[SWD-02] Frordning (2003:789) om skydd mot olyckor (Civil Protection Ordinance), 20 November 2003, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/forordning-2003789-om-skydd-mot-olyckor_sfs-2003-789.

[SWD-03] Freskrifter om ndring i Strlskerhetsmyndighetens freskrifter och allmnna rd (SSMFS 2008:1) om skerhet i krntekniska anlggningar, SSMFS 2011:3, 1 November 2011, https://www.stralsakerhetsmyndigheten.se/contentassets/6b6ce39b86b845c998c dc0062c07353e/ssmfs-20113-foreskrifter-om-andring-i-stralsakerhetsmyndighetens-foreskrifter-ssmfs-20081-om-sakerhet-i-karntekniska-anlaggningar.pdf.

[SWD-04] Miljbalk (1998:808) (Swedish Environmental Code), 11 June 1998, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/miljobalk-1998808_sfs-1998-808.

[SWD-05] Lag (2010:1011) om brandfarliga och explosiva varor (Flammable and Explosive Goods Act), 1 July 2010, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/lag-20101011-om-brandfarliga-och-explosiva_sfs-2010-1011.

[SWD-06] Arbetsmiljlag (1977:1160) (Work Environment Act), 19 December 1977, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/arbetsmiljolag-19771160_sfs-1977-1160.

[SWD-07] Arbetsmiljfrordning (1977:1166) (Work Environment Ordinance), 19 December 1977, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/arbetsmiljoforordning-19771166_sfs-1977-1166.

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[SWD-08] Plan- och bygglag (2010:900) (Planning and Building Act), 1 July 2010, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/plan--och-bygglag-2010900_sfs-2010-900.

[SWD-09] Plan- och byggfrordning (2011:338) (Planning and Building Ordinance), 31 March 2011, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/plan--och-byggforordning-2011338_sfs-2011-338.

[SWD-10] The Swedish National Board of Housing, Building and Plannings general recommendations (BFS 2011:27) on the analytical design of a buildings fire protection BBRAD, 18 June 2013, https://www.boverket.se/globalassets/publikationer/dokument/2013/bbrad-bfs-2011-27-tom-2013-12-english.pdf.

[SWD-11] The Swedish National Board of Housing, Building and Plannings building regulations - mandatory provisions and general recommendations, BBR, BFS 2011:6 with amendments up to BFS 2011:26, 10 October 2011.

[SWD-12] Swedish Fire Protection Association: Regler fr automatisk brandlarmsanlggning (automatic fire alarm installations), SBF 110:6, 1 January 2001.

[SWD-13] Statens rddningsverk, Statens rddningsverks allmnna rd och kommentarer om systematiskt brandskyddsarbete, SRVFS 2004:3, ISSN 0283-6165, February 6, 2004, https://www.msb.se/contentassets/244bc5f1e438414eb6b58339cc58939e/statens

-raddningsverks-allmanna-rad-och-kommentarer-om-systematiskt-brandskyddsarbete.pdf.

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3.12 Switzerland 3.12.1 Existing Reactors Figure 11 shows the structure of the legal framework for oversight of nuclear power plants in Switzerland and the bodies compete. According to Art. 90 of the Federal Constitution of the Swiss Confederation [SWT-01], the Confederation is responsible for legislation in the field of nuclear energy. Utilizing this competence, the Nuclear Energy Act [SWT-02] was created to regulate the peaceful use of nuclear energy, especially the safety of nuclear goods, nuclear installations and radioactive waste.

Figure 11 Regulatory Framework for Fire Protection in Switzerland It establishes fundamental principles for safety and security, licensing processes for nuclear related facilities and activities, the regulatory apparatus and financial and governmental aspects.

The accompanying Nuclear Energy Ordinance [SWT-03] establishes more detailed principles for nuclear safety and security for nuclear facilities, e.g., detailed requirements for licensing documents, requirements for the organization of a nuclear facility, requirements for assessment of safety and security analysis. The Nuclear Energy Ordinance explicitly names fire as a hazard to be considered in devising preventive and protective measures against accidents in nuclear installations. The hazard assumptions and associated evaluation criteria are defined in the Ordinance on the Hazard Assumptions and the Assessment of the Protection against Accidents in Nuclear Installations of the Federal Department of the Environment, Transport, Energy and Communications (DETEC) [SWT-04].

Requirements specifically related to radiological protection after incidents and accidents in nuclear installation derive from the Radiological Protection Act [SWT-05] and its implementing provisions in the Radiological Protection Ordinance [SWT-06].

The guideline HSK-R-50 Requirements Important to Safety for Fire Protection in Nuclear Installations [HSK-03] substantiates the implementation of the legal requirements. HSK-R-50 72

describes the fire protection goals and oversight processes and introduces high-level requirements on the performance of fire protection measures, such as the physical effects of fire to be considered. It also enumerates the documentation to be generated by the licensee, notably the fire protection conception, the fire protection plans and inventories of fire loads, the training schedule as well as the various procedures for firefighting and related operator actions.

This guideline is currently being revised. Oversight on fire protection measures is provided by the designated cantonal authorities, especially the cantonal fire insurance institutions, which usually refer to regulations also applicable to non-nuclear facilities (Vereinigung Kantonaler Feuerversicherungen) [VKF-15].

The DETEC Ordinance on the Hazard Assumptions and the Assessment of the Protection against Accidents in Nuclear Installations specifies fires to be among the incidents and accidents to be considered in the safe shutdown analysis by the licensee. It subdivides design-base accidents into three categories according to their frequency with internal events less frequent than 10-6 a-1 being beyond-design-base accidents. The categories within the design base are connected to requirements regarding the nuclear defense-in-depth and barrier concepts as well as limits of the dose to the public provided in the Radiological Protection Ordinance. The Guideline ENSI-A01 Requirements for Deterministic Accident Analysis for Nuclear Installations: Scope, Methodology and Boundary Conditions of the Technical Accident Analysis [ENSI-18] provides the framework for the safe shutdown analysis of incidents and accidents of all kinds, including fires. In its newly released revision, there are also additional specific requirements on the interpretation of fire frequencies and the single failure criterion.

The Nuclear Energy Ordinance requires the development and use of a Probabilistic Safety Analysis (PSA) for all relevant operating modes of the Swiss nuclear power plants (NPPs).

These requirements are further specified in two regulatory guidelines aimed at harmonizing the use and development of PSA.

Guideline ENSI-A05 Probabilistic Safety Analysis (PSA): Quality and Scope [ENSI-18a],

defines the quality and scope on a plant-specific Level 1 and Level 2 PSA. The guideline specifies also the requirements on the probabilistic fire analysis including acceptable criteria on the screening of fire scenarios.

Guideline ENSI-A06 Probabilistic Safety Analysis (PSA): Applications [ENSI-15] formalizes the requirements for applying PSA to NPPs. It defines general principles for all PSA applications and the scope of mandatory PSA applications. With the aim of identifying potential plant improvements, this guideline specifies the evaluation of the safety level, the balance of risk contributors, plant modifications (including technical specifications) and operational experience.

In case of a reportable event involving PSA-relevant systems, structures or components, such as a fire in the respective area would be, the incremental conditional core damage probability is to be reported.

3.12.2 New Reactors Since licenses related to new power reactors are currently not being sought in Switzerland, ENSI has not formulated a recent position on fire protection in such installations.

3.12.3 Supplementary Information

[HSK-03] Hauptabteilung für die Sicherheit der Kernanlagen (HSK), Sicherheitstechnische Anforderungen an den Brandschutz in Kernanlagen, HSK-R-50/d, March 2003, https://www.ensi.ch/de/dokumente/richtlinie-hsk-r-50-deutsch.

[SWT-01] Swiss Confederation: Federal Constitution, 101, 18 April 1999, https://www.admin.ch/opc/en/classified-compilation/19995395.

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[SWT-02] Federal Assembly of the Swiss Confederation, Nuclear Energy Act, 732.1, 21 March 2003, https://www.admin.ch/opc/en/classified-compilation/20010233.

[SWT-03] Federal Council, Nuclear Energy Ordinance, 732.11, 10 December 2004, https://www.admin.ch/opc/en/classified-compilation/20042217.

[SWT-04] Federal Department of the Environment, Transport, Energy and Communications (DETEC), Verordnung des UVEK über die Gefhrdungsannahmen und die Bewertung des Schutzes gegen Strflle in Kernanlagen, 732.112.2, 17 June 2009, https://www.admin.ch/opc/de/classified-compilation/20090231.

[SWT-05] Federal Assembly of the Swiss Confederation, Radiological Protection Act, 814.50, 22 March 1991, https://www.admin.ch/opc/en/classified-compilation/19910045.

[SWT-06] Swiss Federal Council, Radiological Protection Ordinance, 814.501, 22 June 1994, https://www.admin.ch/opc/en/classified-compilation/19940157.

[VKF-15] Vereinigung Kantonaler Feuerversicherungen (VKF), Brandschutzvorschriften 2015, https://www.bsvonline.ch/de/vorschriften/?req=Norm_&anchor.

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3.13 United Kingdom The information contained herein was accurate at the end of July 2018 and applies to commercial nuclear power generating stations in the United Kingdom.

3.13.1 Existing Reactors There is no difference in the legislative expectations placed on existing versus new reactors.

The methods and details of analysis that would be acceptable to the Office for Nuclear Regulation (ONR) may be different at various life-cycle stages, as an existing reactor may be able to offer operating experience and feedback as an alternative to detailed analysis which may not be possible for a new-build facility.

The licence conditions place requirements on the licensee to collect operational experience through incident reporting arrangements and to undertake Periodic Safety Reviews (PSR) to ensure that the safety case remains current and informed.

When making judgements, ONR recognises that there may have been changes in the legislative framework and relevant good practice since the construction of existing reactors and that it may be unreasonable to require the dutyholder to implement the current expectations of modern standards. In such cases the dutyholder is required to provide a demonstration that the cost (in terms of time, money or trouble) is grossly disproportionate when compared to the risk reduction gained; i.e. that it may not be proportionate to implement changes to meet the same xpectations that new reactors would be expected to achieve.

3.13.1.1 Overview of Legislation The legal framework for the nuclear industry in the UK is based on the Health and Safety at Work Act 1974 (HSWA74) [UK-01], the Energy Act 2013 (TEA13) [UK-02] and the Nuclear Installations Act 1965 (NIA65) [UK-03]. HSWA places duties on all employers, including those in the nuclear industry, to look after the health and safety of both their employees and the public.

However, because of the particular hazards associated with the nuclear industry, including the potential for accidents to cause widespread harm and social disruption, further legislation is also in place, specifically the NIA65. Additionally, there are provisions for nuclear regulations to be made under TEA13, as well as specific regulations under HSWA74 such as the Ionizing Radiations Regulations 2017 (IRR17) [UK-04] and Radiation (Emergency Preparedness and Public Information) Regulations 2001 (REPPIR) [UK-05].

TEA13 was introduced to establish a legislative framework for delivering secure, affordable and low carbon energy. Part 3 of TEA13 establishes the Office for Nuclear Regulation (ONR) as a body corporate and sets out the purposes of the ONR. The Act confers a variety of powers and duties onto the ONR. These reflect a number of the roles which the ONR is to perform. Its primary role is to regulate the nuclear industry in the areas set out in its five purposes: nuclear safety, nuclear site health and safety, nuclear security, nuclear safeguards and transport.

The primary conventional fire safety legislation for the UK includes the Regulatory Reform (Fire Safety) Order 2005 in England and Wales [UK-06], the Fire Safety (Scotland) Regulations 2006

[UK-07] and the Fire Safety Regulations (Northern Ireland) 2010 [UK-08]. Supporting the legislation there are a number of British Standards which, though they do not apply on a nuclear site, they do remain part of the basis of relevant good practice; one of the relevant main standards is BS 9999:2017 Code of practice for fire safety in the design, management and use of buildings [UK-09].

Nuclear power plants (NPPs) are required under the NIA65 to hold a nuclear site license. The nuclear site license granted by ONR is a legal document, issued for the full life cycle of the facility. It contains site-specific information, such as the licensee's address and the location of 75

the site and defines the number and type of installations permitted. Such installations include nuclear power stations, research reactors, nuclear fuel manufacturing and reprocessing, and the storage of radioactive matter in bulk.

A set of 36 Standard Conditions, covering design, construction, operation and decommissioning, is also attached to each license [UK-10]. These conditions require licensees to implement adequate arrangements to ensure compliance.

The UK operates a goal-setting approach to nuclear safety regulation and requires that the licensee or dutyholder demonstrate all risks are reduced so far as is reasonably practical (SFAIRP). Regulatory expectations are specified, and dutyholders are required to determine how best to achieve them and justify their chosen approach. This enables dutyholders to be innovative and flexible in how they achieve the high standards of nuclear safety and security required by implementing arrangements that meet their particular circumstances. It also strengthens accountability and encourages the adoption of relevant good practice and continuous improvement. The UK goal setting legal framework places the duty on the licensee to find the balance between nuclear and conventional safety requirements to comply with the law.

A key principle of the UKs approach is that nuclear licensees are required to build, operate and decommission nuclear sites in a way that ensures that risks are kept as low as reasonably practicable. This is referred to as the ALARP principle and requires licensees to demonstrate that they have done everything reasonably practicable to reduce risks. This requires them to balance the level of risk posed by their activities against the measures needed to control that risk in terms of money, time or trouble. However, they do not have to take action if those measures would be grossly disproportionate to the level of risk averted.

3.13.1.2 Regulatory Fire Protection Expectations The dutyholder is expected to have undertaken a systematic analysis of the hazards, potential consequences, fault progression and proposed protective measures for their installation. The safety case must demonstrate compliance with UK legislation, relevant regulations and the nuclear site licence conditions. In order to demonstrate that risks are ALARP, the dutyholder is expected to adopt relevant good practice.

ONR uses Safety Assessment Principles (SAPs) [ONR-14], together with supporting Technical Assessment Guides (TAGs) (which provide subject specific guidance and expectations), to guide their regulatory judgements when undertaking technical assessments of nuclear site licensees safety submissions. The SAPs were revised in 2014 to take account of learning from the lessons from Fukushima and work by the International Atomic Energy Agency (IAEA), in particular the development of IAEAs design standard on the safety of nuclear power plants (SSR-2/1). [IAEA-16]. Underpinning these is the legal duty on licensees to reduce risks so far as is reasonably practicable.

The SAPs and TAGs are developed to provide a concise summary of principles and guidance that dutyholders are expected to apply. They are derived from:

• Interpretation of British law;

• Requirements and guides of the International Atomic Energy Agency (IAEA);

• Western European Nuclear Regulators Association (WENRA) Reference Levels ;

• NUREG - if relevant and in line with ONR expectations;

• British standards; 76

• Recognized industry practice.

The key TAGs addressing fire include: TAG External Hazards [ONR-18a] and TAG 14 -

Internal Hazards [ONR-16a] with support from TAG Guidance on the Demonstration of ALARP (As Low As Reasonably Practicable) [ONR-18] and TAG-030 - Probabilistic Safety Analysis [ONR-16], where applicable.

ONRs regulatory expectation is that licensees should adopt a defense in depth, approach. This defense in depth should be secured by characteristics as near as possible to the top of the hierarchy of safety measures adopted in the design of the facilities against fire hazards with a focus on prevention, protection and mitigation.

Good engineering design should show that precedence has been given to fire prevention (e.g.

minimization of combustible inventories) and also how the design ensures that fires should they occur would not lead to unacceptable consequences. This involves limiting the severity of any fires (e.g. by early detection and extinguishing, and by the provision of suitably rated fire barriers that prevent fire spread). For any severe fires which do arise, the consequences on nuclear safety relevant structures, systems and components (SSCs) should be limited by design (e.g. by provision of redundant safety measures in segregated fire compartments).

Compartments formed by the installation of fire barriers should be rated to withstand total combustion of the fire load in the compartment. Where this is not practical due to conflicts with other plant design requirements, separation of the items important to safety could be achieved using an appropriate combination of limited combustibles, separation by distance, local passive fire barriers and fire detection and suppression systems.

The safety case should provide reference to surveys of combustible substances undertaken, which should be systematic and demonstrably complete. Transient fire loads that could be introduced either during construction, outages or at power modes should also be identified. It should be noted that all combustible inventories, including transient and protected combustible loads can contribute to a fully developed fire and the overall fire load.

The results of the fire hazard identification process should be documented to provide a basis for the hazard analysis required. In particular, fire hazards to items important to safety that may arise due to the failure of barriers and escalating fires should be identified using an appropriate systematic methodology, and the results documented.

The potential for fire initiation and growth and the possible consequences on items important to safety should be determined as part of the fire hazard analysis with the following key purposes:

• Determine consequences to SSC and them withstand, determine if further separation, isolation and redundancy is required;

• Determine performance requirements of fire safety measures;

• Specify the capacity and capability of the fire detection systems and any other active fire protection provisions;

• Test fire hazard identification and design substantiation assumptions and limitations; and

• Determine consequential effects from fires - e.g. flooding, explosion, dropped loads, etc.

The analysis approach must adequately address the inherent uncertainty associated with fire initiation and progression, and any reliance placed on the reliability of fire protection or mitigation measures, for example:

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• The analysis captures the outcome of design basis fires. Sensitivity studies may be required to establish the design basis fire.

• Generally, it should be expected that the analysis will be based on complete burn-out of all combustible loads including any protected loads.

• All SSCs in the fire compartment are lost in the fire.

• Fire analysis during maintenance operations, outages etc. should be developed and take into account the availability of fire barriers, status of doors/ hatches connecting different fire compartments etc.

Internal and external hazards, whilst usually assessed individually rarely occur in isolation. A single event may result in a number of hazards which occur simultaneously or in quick succession. Fire assessments will be expected to include reasonably foreseeable hazard combinations. There are three types of hazard combinations.

• Independent Hazards; when more than one internal and/or external hazard applies simultaneously. This can be the case, for example, of nominally frequent events such as internal fire and flooding when there is no causation link between them.

• Consequential Hazards: an internal or external hazard directly poses one or more additional hazards to plant and structures (e.g. an earthquake that causes a fire).

• Concurrent Hazards: a hazard or event results in multiple hazard(s), which occur simultaneously. An example of this would be an oil leak or flammable atmosphere leading to an explosion and or fire.

It is expected that the fire analysis should clearly identify the fire compartments and safety divisional areas of the design. It should also document the effects of fire scenarios on the nuclear safety relevant barriers and plant taking into account the most challenging plant state.

Typical features of adequate fire analysis generally include the following:

• The bounding combustible inventory within each fire compartment;

• For each compartment, the scenarios identified to present the most significant threat to the fire compartment barriers;

• Fire modelling;

• The time-temperature profiles;

• Substantiation of claimed barriers;

• Sensitivity analysis should be undertaken to show that there are sufficient safety margins;

• Application of relevant good practice, codes and standards; and

• Local fire effects.

Some common characteristics of design features and considerations in the design of safety measures against fire are covered in detail within TAG-14 [ONR-16a].

The hazards arising from fires, where either single or multiple measures are designed to prevent their escalation, should be quantified and assessed to verify the adequacy of the measures for preventing fire spread and maintaining the integrity of the safety systems delivering fundamental safety functions. As part of the assessment, the safety measures should be allocated an appropriate safety category and safety classification to clearly identify their role in ensuring 78

nuclear safety. The safety measures should be included in maintenance schedules and operating instructions, as appropriate. In particular, safety management procedures should be established for maintaining the integrity and reliability of fire barriers and any penetrations such as doors, cable and pipe conduit seals, heating, ventilation and air conditioning ducts and dampers, and the fire detection, alarm and extinguishing systems.

3.13.1.3 Post Fire Safe Shutdown Expectations It is an expectation that licensees have adequate measures, plans and equipment in place to safely recover from fire scenarios. By determining an adequate and bounding design basis fire, the design basis accident analysis contained within the safety case derives a reactor management philosophy and engineered controls that will manage the identified risks.

It is a requirement of the nuclear site licence (licence condition 11) that the licensee will make and implement adequate emergency arrangements. REPPIR [UK-05] requires that the nuclear and radiological consequences of severe accidents be considered on the surrounding population and that appropriate protective measures be identified and implemented.

Recent learning from Fukushima has resulted in all licensees reconsidering their emergency preparedness and resilience using a stress test exercise. Understanding of outputs from the stress tests resulted in increased regulatory expectations for provision of on-site and off-site back-up emergency equipment; review of human capability claims and closer scrutiny of combined internal and external hazards, including fire.

The existing nuclear power generation capability in the UK is delivered by fifteen operating reactors made up of seven sites containing two Advanced Gas Reactors (AGRs) each and one NPP utilizing one pressurized water reactor (PWR). The durations of fault sequences leading to core melt are substantially longer for AGRs (up to one to two days) than for PWRs (up to three hours), allowing more time for operator intervention. The post fire safe shutdown expectations therefore vary a little for the two reactor types, this must be considered within the safety case.

3.13.2 New Reactors Though there is no difference in the legislative expectations placed on existing versus new reactors, there is generally an expectation that the dutyholder will undertake an exhaustive safety analysis of their proposed installation and proactively demonstrate that modern standards have been met and that all reasonable measures have been taken to reduce risks so far as is reasonably practicable. This will generally be more onerous for the planned water-cooled reactors than is expected for the existing fleet.

As the UK legal framework is goal setting, the onus is on the dutyholder to consider all options and identify the most effective mechanisms to reduce risk; it is not sufficient to rely only on compliance with relevant standards.

ONR provides the Generic Design Assessment (GDA) process for proposed new reactors in the UK. This is a voluntary process that the UK government strongly encourages all prospective new licensees to follow. It takes place in advance of licensing and provides early guidance to help comply with UK legislation and demonstrate that management of risks is ALARP. This process provides advice on the licensability of a design before a potential licensee procures a site, thereby reducing financial risk to them.

GDA or assessments for new reactors will expect available operational experience of similar reactor designs already in operation elsewhere in the world to inform the current design and development of the safety case.

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3.13.3 Supplementary Information

[ONR-14] Office for Nuclear Regulation (ONR): Safety Assessment Principles (SAPs) for Nuclear Facilities, 2014 Edition, Revision 0, November 2014, http://www.onr.org.uk/saps/index.htm.

[ONR-16] Office for Nuclear Regulation (ONR): Probabilistic Safety Analysis, Nuclear Safety Technical Assessment Guide NS-TAST-GD-030, Revision 5, June 2016, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-030.pdf.

[ONR-16a] Office for Nuclear Regulation (ONR): Internal Hazards, Nuclear Safety Technical Assessment Guide NS-TAST-GD-014, Revision 4, September 2016, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-014.htm.

[ONR-18] Office for Nuclear Regulation (ONR): Guidance on the Demonstration of ALARP (As Low as Reasonably Practicable), Nuclear Safety Technical Assessment Guide NS-TAST-GD-005, Revision 9, March 2018, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-005.pdf.

[ONR-18a] Office for Nuclear Regulation (ONR): External Hazards, Nuclear Safety Technical Assessment Guide NS-TAST-GD-013, Revision 7, October 2018, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-013.htm.

[UK-01] Health and Safety at Work Act 1974 (HSWA74),

https://www.legislation.gov.uk/ukpga/1974/37.

[UK-02] The Energy Act 2013 (TEA13),

http://www.legislation.gov.uk/ukpga/2013/32/contents/enacted.

[UK-03] Nuclear Installations Act 1965 (NIA65),

https://www.legislation.gov.uk/ukpga/1965/57.

[UK-04] Ionizing Radiations Regulations 2017 (IRR17),

http://www.legislation.gov.uk/uksi/2017/1075/contents/made.

[UK-05] Radiation (Emergency Preparedness and Public Information) Regulations 2001 (REPPIR),

https://www.legislation.gov.uk/uksi/2001/2975/contents/made.

[UK-06] Regulatory Reform (Fire Safety) Order 2005, http://www.legislation.gov.uk/uksi/2005/1541/contents/made.

[UK-07] Fire Safety (Scotland) Regulations 2006, http://www.legislation.gov.uk/ssi/2006/456/contents/made.

[UK-08] Fire Safety Regulations (Northern Ireland) 2010, https://www.legislation.gov.uk/nisr/2010/325/contents/made.

[UK-09] BS 9999:2017: Code of practice for fire safety in the design, management and use of buildings, 2017.

[UK-10] Nuclear Site Licence Conditions, http://www.onr.org.uk/licensing.htm.

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3.14 United States of America 3.14.1 Existing Reactors The United States Nuclear Regulatory Commissions (NRC) Fire Protection Regulations abide by the Title 10, Energy, Code of Federal Regulations (CFR). The CFR is a codification of the general and permanent rules published in the Federal Register by the Executive departments and agencies of the United States Federal Government. The following references codify the United States commitments to fire protection in commercial NPPs.

The primary objectives of fire protection programs (FPPs) at U.S. commercial nuclear plants are to minimize both the probability of occurrence and the consequences of fire. To meet these objectives, the FPPs for operating nuclear power plants are designed to provide reasonable assurance, through defense in depth, that a fire will not prevent the necessary safe-shutdown functions from being performed and that radioactive releases to the environment in the event of a fire will be minimized.

During the initial implementation of the U.S. nuclear reactor program, the broad performance objectives of Title 10, Energy, Code of Federal Regulations (10 CFR), Criterion 3, Fire protection, of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities. formed the basis for regulatory acceptance of FPPs at nuclear power plants. Appendix A of 10 CFR Part 50 establishes the necessary design, fabrication, construction, testing, and performance requirements for structural, systems, and components (SSCs) important to safety. Criterion 3 addresses fire protection requirements and specifies, in part, that (1) SSCs important to safety must be designed and located to minimize the probability and effects of fires and explosions, (2) noncombustible and heat-resistant materials must be used wherever practical, and (3) fire detection and suppression systems must be provided to minimize the adverse effects of fires on SSCs important to safety. However, given the lack of detailed implementation guidance for this Criterion 3 during the early stages of nuclear power regulation, the NRC generally considered the level of fire protection acceptable if the facility complied with local fire codes and received an acceptable rating from its fire insurance underwriter. Thus, the fire protection features installed in early U.S. nuclear power plants were very similar to those installed in conventional fossil-fueled power generation stations.

The fire at the Browns Ferry Nuclear Power Plant, Unit 1, on March 22, 1975, was a pivotal event that brought fundamental change to fire protection and its regulation in the U.S. nuclear power industry. The investigations that followed the Browns Ferry fire identified significant deficiencies, in both the design of fire protection features and the licensees procedures for responding to a fire event. The investigators concluded that the occupant safety and property protection concerns of fire insurance underwriters did not sufficiently encompass nuclear safety issues, especially in terms of the potential for fire damage to cause the failure of redundant success paths of SSCs important for safe reactor shutdown.

One of the first actions the Commission took in response to the Browns Ferry fire was to impose technical specifications in accordance with 10 CFR 50.36, Technical specifications, for fire protection program. During 1976 to1977, each operating nuclear power plant was given a sample of the recommended standard technical specifications for fire protection systems and features.

In 1976, the NRC issued Branch Technical Position (BTP) Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants, which incorporated the recommendations from the Browns Ferry fire investigation and provided technical guidelines to assist licensees in preparing their FPPs. As part of this action, the NRC 81

staff asked each licensee to provide an analysis dividing the plant into distinct fire areas and demonstrating that redundant success paths of components required to achieve and maintain safe-shutdown conditions for the reactor were adequately protected from fire damage. However, the guidelines of BTP APCSB 9.5-1 applied only to those licensees that had filed for a construction permit after July 1, 1976.

Later in 1976, to establish defense-in-depth FPPs without significantly affecting the design, construction, or operation of existing plants that were either already operating or well past the design stage and into construction, the NRC modified the guidelines in APCSB 9.5-1 and issued Appendix A to BTP APCSB 9.5-1. This guidance provided acceptable alternatives in areas where strict compliance with BTP APCSB 9.5-1 would require significant physical modifications.

Additionally, the NRC informed each licensee that the staff would use the guidance in Appendix A to BTP APCSB 9.5-1 to analyze the consequences of a postulated fire within each area of the plant and asked licensees to provide results of a fire hazards analysis for each unit and the technical specifications for the present fire protection systems.

Early in 1977, each licensee responded with an FPP evaluation that included a fire hazards analysis. The staff reviewed these analyses and inspected operating reactors to examine the relationship of SSCs important to safety with fire hazards, potential consequences of fires, and fire protection features. After reviewing the licensees responses, the staff determined that additional guidance on the management and administration of FPPs was necessary and issued Generic Letter (GL) 77-02, Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance, which provided criteria used by the staff to review specific elements of a licensees FPP, including organization, training, combustible and ignition source controls, firefighting procedures, and quality assurance (QA). The BTP review process resolved many fire protection issues.

By 1980, most operating plants had completed their analyses and implemented much of the FPP guidance and recommendations. In most cases, the NRC found the licensees proposed modifications resulting from these analyses to be acceptable. In some instances, however, technical disagreements with the NRC staff led some licensees to oppose the adoption of certain specified fire protection recommendations, such as the requirements for fire brigade size and training, water supplies for fire suppression systems, alternative or dedicated shutdown capability, emergency lighting, qualifications of penetration seals used to enclose places where cables penetrated fire barriers, and the prevention of reactor coolant pump (RCP) oil system fires. Following deliberation, the Commission determined that, given the generic nature of some of the disputed issues, a rulemaking was necessary to ensure proper implementation of the NRCs fire protection requirements.

In late 1980, the NRC published the Fire protection rule, 10 CFR 50.48, which specified broad performance requirements, as well as Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, to 10 CFR Part 50, which contained detailed regulatory requirements for resolving the disputed issues.

As originally proposed, 10 CFR Part 50, Appendix R would have applied to all plants licensed before January 1, 1979, including those for which the staff had previously agreed that the fire protection features met the provisions of Appendix A to BTP APCSB 9.5-1.

After analyzing comments on the proposed rule, the Commission determined that only 3 of the 15 items in Appendix R were of such safety significance that they should apply to all plants (licensed before January 1, 1979), including those for which the staff had previously approved alternative fire protection actions. These three items are fire protection of safe-shutdown capability (including alternative or dedicated shutdown systems), emergency lighting, and the RCP oil collection system. Accordingly, the final rule required all reactors licensed to operate 82

before January 1, 1979, to comply with these three items, even if the NRC had previously approved alternative fire protection features in these areas.

In addition, the rule provided an exemption process. A licensee can request an exemption if the required fire protection feature to be exempted would not enhance fire protection safety in the facility, or if a modification to meet regulatory requirements might be detrimental to overall safety. During the initial backfit of the fire protection regulation, the NRC approved many plant-specific exemptions (i.e., alternative methods to achieve the underlying purpose of the regulation).

As licensees programs became more compliant with the fire protection regulations, the number of exemptions requested and approved decreased. Even so, several hundred exemptions to specific elements of the NRCs fire protection requirements were issued. This progression, the broad provisions of the General Design Criteria, the detailed implementing guidance, the plant-by-plant review, and finally the issuance and backfit of the fire protection regulations and the prescriptive requirements of Appendix R created a complex regulatory framework for fire protection in U.S. nuclear power plants licensed before 1979 and resulted in the issuance of additional guidelines, clarifications, and interpretations, primarily as Generic Letters (GLs).

The NRC does not require nuclear plants licensed after January 1, 1979, to meet the provisions of Appendix R unless directed to do so in specific license conditions. The NRC reviewed these nuclear plants using the guidelines of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (LWR Edition), which subsumed the criteria specified in Appendix R.

Following promulgation of 10 CFR 50.48 and Appendix R, the staff issued GL 81-12, Fire Protection Rule (45 FR 76602; November 19, 1980), to identify the information necessary to perform the reviews of licensee compliance with the alternative or dedicated shutdown requirements of Section III.G.3 of Appendix R. The guidance defined safe-shutdown objectives, reactor performance goals, necessary safe-shutdown systems and components, and associated circuit identification and analysis methods. GL 81-12 also asked licensees to develop technical specifications for safe-shutdown equipment not included in the existing plant technical specifications.

Most licensees requested and received additional time to perform their reanalysis, propose modifications to improve post-fire safe-shutdown capability, and identify exemptions for certain fire protection configurations. In reviewing some exemption requests, the staff noted that a number of licensees had significantly different interpretations of certain requirements. The staff identified these differences in the draft safety evaluation reports (SERs) and discussed them on several occasions with the cognizant licensees. These discussions culminated in the issuance of GL 83-33, NRC Positions on Certain Requirements of Appendix R to 10 CFR 50..

Following inspections of operating plants, which identified a number of significant items of noncompliance and disagreements regarding the implementation of interpretations provided in GL 83-33, the industry requested interpretations of certain Appendix R requirements and prepared a list of questions to be discussed. The NRC responded by holding workshops to assist the industry in understanding the NRCs requirements and to improve the staffs understanding of the industrys concerns. The staff issued GL 86-10, Implementation of Fire Protection Requirements, as a result of these interactions.

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GL 86-10 provided: (1) detailed interpretations of Appendix R requirements including frequently asked questions and answers; and (2) the means by which a licensee may make changes to the approved fire protection program without prior approval of the Commission. To accomplish the latter, the licensee had to: (1) incorporate into its Updated Final Safety Analysis Report their fire protection program approved by the NRC; (2) adopt a standard license condition cited in GL 86-10; and (3) ensure that self-approved changes to the fire protection program would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. With the issuance of GL 86-10, the term 86-10 evaluation entered the fire protection lexicon. For those licensees willing to incorporate their fire protection programs into their Updated Final Safety Analysis Report and adopt the standard license condition cited in GL 86-10, it provided a less burdensome pathway to make simple changes to their fire protection programs without submitting license amendment/exemption requests provided the change had no adverse effect.

Under GL 86-10 guidance, most technical specification requirements for the operability of the fire protection systems and features and the associated compensatory measures were transferred from the technical specifications to documents referred to in the Updated Final Safety Analysis Report, such as a Technical Requirements Manual 3. In GL 86-10 the 2F Commission recommended that licensees apply for an amendment to their operating licenses to remove fire protection technical specifications and adopt a standard license conditions for fire protection.

In addition, the Commission concluded that a standard license condition, requiring compliance with the provisions of the fire protection program as described in the Updated Final Safety Analysis Report, should be used to ensure uniform enforcement of the fire protection requirements.

GL 88-12, Removal of Fire Protection Requirements from Technical Specifications, gave licensees additional guidance for implementing the standard license condition and relocating the technical specifications associated with fire detection and suppression, fire barriers, and fire brigade staffing. Licensees were to retain the technical specifications associated with safe-shutdown equipment and the administrative controls related to fire protection under the guidance of the generic letter.

Beginning in the late 1990s, the Commission provided the NRC staff with guidance for identifying and assessing performance-based approaches to regulation. In SECY-98-058, Development of a Risk-Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants, the NRC staff proposed to the Commission that the staff work with the National Fire Protection Association (NFPA) and the industry to develop a performance-based, risk-informed consensus standard for fire protection for nuclear power plants that could be endorsed in a rulemaking. In 2001, NFPA published the 2001 edition of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, as an American National Standard for performance-based fire protection for light-water nuclear power plants. In 2004, the Commission amended its fire protection requirements in 10 CFR 50.48 to add 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805. which incorporates by reference the 2001 edition of NFPA 805, with certain exceptions, and allows 3 The Technical Requirements Manual is a document that is part of the operating nuclear power plants licensing basis, but any changes to it do not have to be approved in advance by the NRC (10 CFR 50.36). In other words, a license amendment request is not needed to modify the Technical Requirements Manual.

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licensees to apply for a license amendment to comply with the new rule. NFPA has issued subsequent editions of NFPA 805, but the regulation does not endorse them.

3.14.1.1 Overview Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities.

The regulations in this part provide for the licensing of production and utilization facilities pursuant to the Atomic Energy Act of 1954, as amended and Title II of the Energy Reorganization Act of 1974.

1. Criterion 3, Fire protection, of Appendix A, General Design Criteria for Nuclear Power Plant, to 10 CFR Part 50. [NRC-01] Criterion 3 specifies the contents of a fire protection plan and lists the basic fire protection guidelines for the plan.

a. Used in the 1970s prior to the Browns Ferry Fire of 1975 (NUREG/KM-002), this document was used as justification by the U.S. Atomic Energy Commission (AEC) and U.S. NRC for acceptance of fire protection programs at nuclear power plants.

b. It is based on very broad performance objectives.

2. 10 CFR 50.48 and Appendix R [NRC-02]

a. 10 CFR 50.48, Fire protection, was issued in 1980 and contained broad deterministic performance requirements.

b. And Appendix R to 10 CRF Part 50 was also issued and provided specific and detailed requirements for addressing disputed issues.

3. 10 CFR 50.48(c) (69 FR 33550, June 16, 2004)

a. This rule was published on 16 June 2004.

b. Allows the voluntary adoption of probabilistic or risk-informed performance-based fire protection program in accordance with:

i. National Fire Protection Association (NFPA) 805: Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition

[NFPA-01]

4. 10 CFR 50.48(f) requires licensees that have certified the permanent cessation of operations and the removal of fuel from the reactor vessel to maintain a fire protection program to address the potential for fires that could result in a radiological hazard.

Table 2 Summary of the U.S. Operating NPPs Fire Protection Regulations Plants Licensed Before January 1, 1979 Plants Licensed After January 1, 1979 Pre-1979 Plants Post-1979 Plants Plants licensed to operate before January 1, Plants licensed to operate after January 1, 1979, must meet some portions of the 1979, must comply with 10 CFR 50.48(a), as requirements of Appendix R, to 10 CFR Part well as any plant specific fire protection 50, except to the extent provided in 10 CFR license condition. The fire protection license 50.48(b). These plants were also required to condition for these plants typically references meet the general fire protection criteria from NRC safety evaluation reports as the product Appendix A to BTP APCSB 9.5-1 to which of initial licensing reviews against Appendix A they committed and referenced in their safety to BTP APCSB 9.5-1 and the criteria of evaluation report. certain sections of 10 CFR 50, Appendix R or 85

NUREG-0800, SRP and BTP Chemical and Engineering Branch (CMEB) 9.5.1, Revision 3, which includes similar criteria specified in 10 CFR 50, Appendix R.

Table 3 Summary of U.S. Regulations for Pre-1979, Post-1979, and NFPA 805 Plants Pre-79 Plants Post-79 Plants NFPA 805 Plants Operating Plants Licensed Operating Plants Licensed Before January 1979 After January 1979 10 CFR 50.48(a) 10 CFR 50.48(a) 10 CFR 50.48(a) 10 CFR 50.48(b) or Appendix Appendix A to 10 CFR Part 10 CFR 50.48(c)

R 50, Criterion 3 Appendix A to 10 CFR Part Appendix A to 10 CFR Part 50, Criterion 3 50, Criterion 3 Applicable Appendix R NUREG-0800, Standard NUREG-0800, Standard Sections III.G, III.J, and III.O Review Plan, Section 9.5.1.1 Review Plan, Section 9.5.1.2 only Appendix A to BTP APCSB BTP CMEB 9.5-1 (Includes Regulatory Guide 1.205 9.5-1 Appendix R Requirements Regulatory Guide 1.200 Table 4 Comparison of U.S. NRC Fire Protection Program Features Pre-1979 Post-1979 10 CFR 50.48(c) - NFPA 805

• Appendix A to BTP 9.5-1

• NUREG-0800, SRP 9.5.1

• NFPA 805 License

• Fire Protection Safety Revision 3 Amendment Request Review to Appendix A to

• Fire Protection Submittal

• NFPA 805 SER

• 9.5-1

• Fire Protection SER

• Fire Protection License

• Fire Protection SER

• Fire Protection License Condition

• Fire Protection TS Condition (replaced by TRM IAW GL

• GL 86-10 88-12)

Plants licensed prior to 1985

• Fire Protection License likely had Fire Protection TS.

Condition These plants had TS

• Appendix R replaced by TRM IAW GL 88-

• Appendix R SER, 12.

Alternative Shutdown

• GL 86-10

• Exemptions 3.14.1.2 Supplementary Material

1. Appendix A to Part 50--General Design Criteria for Nuclear Power Plants- Criterion 3 Fire Protection [NRC-01]

Criterion 3 - Fire protection. Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the 86

probability and effect of fires and explosions. Non-combustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

2. § 50.48 Fire protection [NRC-04]

a. (1) Each holder of an operating license issued under this part or a combined license issued under part 52 of this chapter must have a fire protection plan that satisfies Criterion 3 of Appendix A to this part. This fire protection plan must:

i. Describe the overall fire protection program for the facility; ii. Identify the various positions within the licensee's organization that are responsible for the program; iii. State the authorities that are delegated to each of these positions to implement those responsibilities; and iv. Outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage.

(2) The plan must also describe specific features necessary to implement the program described in paragraph (a)(1) of this section such as

i. Administrative controls and personnel requirements for fire prevention and manual fire suppression activities; ii. Automatic and manually operated fire detection and suppression systems; and iii. The means to limit fire damage to structures, systems, or components important to safety so that the capability to shut down the plant safely is ensured.

(3) The licensee shall retain the fire protection plan and each change to the plan as a record until the Commission terminates the reactor license. The licensee shall retain each superseded revision of the procedures for 3 years from the date it was superseded.

(4) Each applicant for a design approval, design certification, or manufacturing license under part 52 of this chapter must have a description and analysis of the fire protection design features for the standard plant necessary to demonstrate compliance with Criterion 3 of Appendix A to this part.

b. Appendix R to this part establishes fire protection features required to satisfy Criterion 3 of Appendix A to this part with respect to certain generic issues for nuclear power plants licensed to operate before January 1, 1979.

(1) Except for the requirements of Sections III.G, III.J, and III.O, the provisions of Appendix R to this part do not apply to nuclear power plants licensed to operate before January 1, 1979, to the extent that

i. Fire protection features proposed or implemented by the licensee have been accepted by the NRC staff as satisfying the provisions of Appendix A to Branch Technical Position (BTP) APCSB 9.5-1 [NRC-05] reflected in NRC fire protection safety evaluation reports issued before the effective date of February 19, 1981; or 87

ii. Fire protection features were accepted by the NRC staff in comprehensive fire protection safety evaluation reports issued before Appendix A to Branch Technical Position (BTP) APCSB 9.5-1 was published in August 1976 [NRC-04].

(2) With respect to all other fire protection features covered by Appendix R, all nuclear power plants licensed to operate before January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections III.G, III.J, and III.O.

c. National Fire Protection Association Standard NFPA 805 [NFPA-01]

(1) Approval of incorporation by reference. National Fire Protection Association (NFPA)

Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805 [NFPA-01]), which is referenced in this section, was approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51. Copies of NFPA 805 may be purchased from the NFPA Customer Service Department, 1 Batterymarch Park, P.O.

Box 9101, Quincy, MA 02269-9101 and in PDF format through the NFPA Online Catalog (http://www.nfpa.org) or by calling 1-800-344-3555 or (617) 770-3000. Copies are also available for inspection at the NRC Library, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852-2738, and at the NRC Public Document Room, Building One White Flint North, Room O1-F15, 11555 Rockville Pike, Rockville, Maryland 20852-2738.

Copies are also available at the National Archives and Records Administration (NARA).

For information on the availability of this material at NARA, call (202) 741-6030, or go to:

http://www.archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html.

(2) Exceptions, modifications, and supplementation of NFPA 805. As used in this section, references to NFPA 805 are to the 2001 Edition [NFPA-01], with the following exceptions, modifications, and supplementation:

i. Life Safety Goal, Objectives, and Criteria. The Life Safety Goal, Objectives, and Criteria of Chapter 1 are not endorsed.

ii. Plant Damage/Business Interruption Goal, Objectives, and Criteria. The Plant Damage/Business Interruption Goal, Objectives, and Criteria of Chapter 1 are not endorsed.

iii. Use of feed-and-bleed. In demonstrating compliance with the performance criteria of Sections 1.5.1(b) and (c), a high-pressure charging/injection pump coupled with the pressurizer power-operated relief valves (PORVs) as the sole fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability (i.e., feed-and-bleed) for pressurized-water reactors (PWRs) is not permitted.

iv. Uncertainty analysis. An uncertainty analysis performed in accordance with Section 2.7.3.5 is not required to support deterministic approach calculations.

v. Existing cables. In lieu of installing cables meeting flame propagation tests as required by Section 3.3.5.3, a flame-retardant coating may be applied to the electric cables, or an automatic fixed fire suppression system may be installed to provide an equivalent level of protection. In addition, the italicized exception to Section 3.3.5.3 is not endorsed.

vi. Water supply and distribution. The italicized exception to Section 3.6.4 is not endorsed. Licensees who wish to use the exception to Section 3.6.4 must submit 88

a request for a license amendment in accordance with paragraph (c)(2)(vii) of this section.

vii. Performance-based methods. Notwithstanding the prohibition in Section 3.1 against the use of performance-based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard. Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under § 50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach;

a. Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;

b. Maintains safety margins; and

c. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

(3) Compliance with NFPA 805.

i. A licensee may maintain a fire protection program that complies with NFPA 805 as an alternative to complying with paragraph (b) of this section for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants licensed to operate after January 1, 1979. The licensee shall submit a request to comply with NFPA 805 in the form of an application for license amendment under § 50.90. The application must identify any orders and license conditions that must be revised or superseded and contain any necessary revisions to the plant's technical specifications and the bases thereof. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the licensee has identified orders, license conditions, and the technical specifications that must be revised or superseded, and that any necessary revisions are adequate. Any approval by the Director or the designee must be in the form of a license amendment approving the use of NFPA 805 together with any necessary revisions to the technical specifications.

ii. The licensee shall complete its implementation of the methodology in Chapter 2 of NFPA 805 (including all required evaluations and analyses) and, upon completion, modify the fire protection plan required by paragraph (a) of this section to reflect the licensee's decision to comply with NFPA 805, before changing its fire protection program or nuclear power plant as permitted by NFPA 805.

(4) Risk-informed or performance-based alternatives to compliance with NFPA 805. A licensee may submit a request to use risk-informed or performance-based alternatives to compliance with NFPA 805. The request must be in the form of an application for license amendment under § 50.90 of this chapter. The Director of the Office of Nuclear Reactor Regulation, or designee of the Director, may approve the application if the Director or designee determines that the proposed alternatives:

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i. Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; ii. Maintain safety margins; and iii. Maintain fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

d. [Reserved]. (As it is used to label a section of the U.S. Code or Code of Federal Regulations, the word 'reserved' means that the section has been saved as 'empty space' to be used later that is, that the section has been 'reserved' for later use)

e. [Reserved]. (As it is used to label a section of the U.S. Code or Code of Federal Regulations, the word 'reserved' means that the section has been saved as 'empty space' to be used later - that is, that the section has been 'reserved' for later use)

f. Licensees that have submitted the certifications required under § 50.82(a)(1) shall maintain a fire protection program to address the potential for fires that could cause the release or spread of radioactive materials (i.e., that could result in a radiological hazard).

A fire protection program that complies with NFPA 805 shall be deemed to be acceptable for complying with the requirements of this paragraph.

(1) The objectives of the fire protection program are to

i. Reasonably prevent these fires from occurring; ii. Rapidly detect, control, and extinguish those fires that do occur and that could result in a radiological hazard; and iii. Ensure that the risk of fire induced radiological hazards to the public, environment and plant personnel is minimized.

(2) The licensee shall assess the fire protection program on a regular basis. The licensee shall revise the plan as appropriate throughout the various stages of facility decommissioning.

(3) The licensee may make changes to the fire protection program without NRC approval if these changes do not reduce the effectiveness of fire protection for facilities, systems, and equipment that could result in a radiological hazard, taking into account the decommissioning plant conditions and activities.

[65 FR 38190, June 20, 2000; 69 FR 33550, June 16, 2004; 72 FR 49495, Aug. 28, 2007]

3. Appendix R to Part 50 Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 [NRC-02]

I. Introduction and Scope This appendix applies to licensed nuclear power electric generating stations that were operating prior to January 1, 1979, except to the extent set forth in § 50.48(b) of this part.

With respect to certain generic issues for such facilities it sets forth fire protection features required to satisfy Criterion 3 of Appendix A to this part.

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."

When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boiloff.

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The phrases "important to safety," or "safety-related," will be used throughout this Appendix R as applying to all safety functions. The phrase "safe shutdown" will be used throughout this appendix as applying to both hot and cold shutdown functions.

Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under postfire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. Three levels of fire damage limits are established according to the safety functions of the structure, system, or component:

Table 5 U.S. 10 CFR 50 Appendix R Fire Damage Safety function Fire damage limits Hot Shutdown One train of equipment necessary to achieve hot shutdown from either the control room or emergency control station(s) must be maintained free of fire damage by a single fire, including an exposure fire. (An exposure fire is a fire in a given area that involves either in situ or transient combustibles and is external to any structures, systems, or components located in or adjacent to that same area. The effects of such fire (e.g., smoke, heat, or ignition) can adversely affect those structures, systems, or components important to safety. Thus, a fire involving one train of safe shutdown equipment may constitute an exposure fire for the redundant train located in the same area, and a fire involving combustibles other than either redundant train may constitute an exposure fire to both redundant trains located in the same area.)

Cold Shutdown Both trains of equipment necessary to achieve cold shutdown may be damaged by a single fire, including an exposure fire, but damage must be limited so that at least one train can be repaired or made operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> using onsite capability.

Design Basis Accidents Both trains of equipment necessary for mitigation of consequences following design basis accidents may be damaged by a single exposure fire.

The most stringent fire damage limit shall apply for those systems that fall into more than one category. Redundant systems used to mitigate the consequences of other design basis accidents but not necessary for safe shutdown may be lost to a single exposure fire.

However, protection shall be provided so that a fire within only one such system will not damage the redundant system.

II. General Requirements A. Fire protection program. A fire protection program shall be established at each nuclear power plant. The program shall establish the fire protection policy for the protection of structures, systems, and components important to safety at each plant and the procedures, equipment, and personnel required to implement the program at the plant site.

The fire protection program shall be under the direction of an individual who has been delegated authority commensurate with the responsibilities of the position and who has available staff personnel knowledgeable in both fire protection and nuclear safety.

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The fire protection program shall extend the concept of defense-in-depth to fire protection in fire areas important to safety, with the following objectives:

1. To prevent fires from starting;

2. To detect rapidly, control, and extinguish promptly those fires that do occur;

3. To provide protection for structures, systems, and components important to safety so that a fire that is not promptly extinguished by the fire suppression activities will not prevent the safe shutdown of the plant.

B. Fire hazards analysis. A fire hazards analysis shall be performed by qualified fire protection and reactor systems engineers to (1) consider potential in situ and transient fire hazards; (2) determine the consequences of fire in any location in the plant on the ability to safely shut down the reactor or on the ability to minimize and control the release of radioactivity to the environment; and (3) specify measures for fire prevention, fire detection, fire suppression, and fire containment and alternative shutdown capability as required for each fire area containing structures, systems, and components important to safety in accordance with NRC guidelines and regulations.

C. Fire prevention features. Fire protection features shall meet the following general requirements for all fire areas that contain or present a fire hazard to structures, systems, or components important to safety.

1. In situ fire hazards shall be identified and suitable protection provided.

2. Transient fire hazards associated with normal operation, maintenance, repair, or modification activities shall be identified and eliminated where possible. Those transient fire hazards that cannot be eliminated shall be controlled and suitable protection provided.

3. Fire detection systems, portable extinguishers, and standpipe and hose stations shall be installed.

4. Fire barriers or automatic suppression systems or both shall be installed as necessary to protect redundant systems or components necessary for safe shutdown.

5. A site fire brigade shall be established, trained, and equipped and shall be on site at all times.

6. Fire detection and suppression systems shall be designed, installed, maintained, and tested by personnel properly qualified by experience and training in fire protection systems.

7. Surveillance procedures shall be established to ensure that fire barriers are in place and that fire suppression systems and components are operable.

D. Alternative or dedicated shutdown capability. In areas where the fire protection features cannot ensure safe shutdown capability in the event of a fire in that area, alternative or dedicated safe shutdown capability shall be provided.

III. Specific Requirements A. Water supplies for fire suppression systems. Two separate water supplies shall be provided to furnish necessary water volume and pressure to the fire main loop.

Each supply shall consist of a storage tank, pump, piping, and appropriate isolation and control valves. Two separate redundant suctions in one or more intake structures from a large body of water (river, lake, etc.) will satisfy the requirement for two separated water storage tanks. These supplies shall be separated so that a failure of one supply will not result in a failure of the other supply.

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Each supply of the fire water distribution system shall be capable of providing for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> the maximum expected water demands as determined by the fire hazards analysis for safety-related areas or other areas that present a fire exposure hazard to safety- related areas.

When storage tanks are used for combined service-water/fire-water uses the minimum volume for fire uses shall be ensured by means of dedicated tanks or by some physical means such as a vertical standpipe for other water service. Administrative controls, including locks for tank outlet valves, are unacceptable as the only means to ensure minimum water volume.

Other water systems used as one of the two fire water supplies shall be permanently connected to the fire main system and shall be capable of automatic alignment to the fire main system. Pumps, controls, and power supplies in these systems shall satisfy the requirements for the main fire pumps. The use of other water systems for fire protection shall not be incompatible with their functions required for safe plant shutdown. Failure of the other system shall not degrade the fire main system.

B. Sectional isolation valves. Sectional isolation valves such as post indicator valves or key operated valves shall be installed in the fire main loop to permit isolation of portions of the fire main loop for maintenance or repair without interrupting the entire water supply.

C. Hydrant isolation valves. Valves shall be installed to permit isolation of outside hydrants from the fire main for maintenance or repair without interrupting the water supply to automatic or manual fire suppression systems in any area containing or presenting a fire hazard to safety-related or safe shutdown equipment.

D. Manual fire suppression. Standpipe and hose systems shall be installed so that at least one effective hose stream will be able to reach any location that contains or presents an exposure fire hazard to structures, systems, or components important to safety.

Access to permit effective functioning of the fire brigade shall be provided to all areas that contain or present an exposure fire hazard to structures, systems, or components important to safety.

Standpipe and hose stations shall be inside PWR containments and BWR containments that are not inerted. Standpipe and hose stations inside containment may be connected to a high quality water supply of sufficient quantity and pressure other than the fire main loop if plant-specific features prevent extending the fire main supply inside containment. For BWR drywells, standpipe and hose stations shall be placed outside the dry well with adequate lengths of hose to reach any location inside the dry well with an effective hose stream.

E. Hydrostatic hose tests. Fire hose shall be hydrostatically tested at a pressure of 150 psi or 50 psi above maximum fire main operating pressure, whichever is greater. Hose stored in outside hose houses shall be tested annually. Interior standpipe hose shall be tested every three years.

F. Automatic fire detection. Automatic fire detection systems shall be installed in all areas of the plant that contain or present an exposure fire hazard to safe shutdown or safety-related systems or components. These fire detection systems shall be capable of operating with or without offsite power.

G. Fire protection of safe shutdown capability.

1. Fire protection features shall be provided for structures, systems, and components important to safe shutdown. These features shall be capable of limiting fire damage so that:

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a. One train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) is free of fire damage; and

b. Systems necessary to achieve and maintain cold shutdown from either the control room or emergency control station(s) can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

2. Except as provided for in paragraph G.3 of this section, where cables or equipment, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:

a. Separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier;

b. Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area; or

c. Enclosure of cable and equipment and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating, In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area; Inside non-inerted containments one of the fire protection means specified above or one of the following fire protection means shall be provided:

d. Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards;

e. Installation of fire detectors and an automatic fire suppression system in the fire area; or

f. Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energy shield.

3. Alternative or dedicated shutdown capability and its associated circuits,1 independent of cables, systems or components in the area, room, zone under consideration should be provided:

a. Where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section; or

b. Where redundant trains of systems required for hot shutdown located in the same fire area may be subject to damage from fire suppression activities or from the rupture or inadvertent operation of fire suppression systems. In addition, fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under consideration.

H. Fire brigade. A site fire brigade trained and equipped for fire fighting shall be established to ensure adequate manual fire fighting capability for all areas of the plant containing structures, systems, or components important to safety. The fire brigade shall be at least five members on each shift. The brigade leader and at least two brigade members shall have sufficient training in or knowledge of plant safety-related systems to understand the effects of fire and fire suppressants on safe shutdown capability. The qualification of fire brigade members shall include an annual physical examination to determine their ability to perform 94

strenuous fire fighting activities. The shift supervisor shall not be a member of the fire brigade. The brigade leader shall be competent to assess the potential safety consequences of a fire and advise control room personnel. Such competence by the brigade leader may be evidenced by possession of an operator's license or equivalent knowledge of plant safety-related systems.

The minimum equipment provided for the brigade shall consist of personal protective equipment such as turnout coats, boots, gloves, hard hats, emergency communications equipment, portable lights, portable ventilation equipment, and portable extinguishers. Self-contained breathing apparatus using full-face positive-pressure masks approved by NIOSH (National Institute for Occupational Safety and Health approval formerly given by the U.S.

Bureau of Mines) shall be provided for fire brigade, damage control, and control room personnel. At least 10 masks shall be available for fire brigade personnel. Control room personnel may be furnished breathing air by a manifold system piped from a storage reservoir if practical. Service or rated operating life shall be a minimum of one-half hour for the self-contained units.

At least a 1-hour supply of breathing air in extra bottles shall be located on the plant site for each unit of self-contained breathing apparatus. In addition, an onsite 6-hour supply of reserve air shall be provided and arranged to permit quick and complete replenishment of exhausted air supply bottles as they are returned. If compressors are used as a source of breathing air, only units approved for breathing air shall be used and the compressors shall be operable assuming a loss of offsite power. Special care must be taken to locate the compressor in areas free of dust and contaminants.

I. Fire brigade training. The fire brigade training program shall ensure that the capability to fight potential fires is established and maintained. The program shall consist of an initial classroom instruction program followed by periodic classroom instruction, fire fighting practice, and fire drills:

1. Instruction

a. The initial classroom instruction shall include:

(1) Indoctrination of the plant fire fighting plan with specific identification of each individual's responsibilities.

(2) Identification of the type and location of fire hazards and associated types of fires that could occur in the plant.

(3) The toxic and corrosive characteristics of expected products of combustion.

(4) Identification of the location of fire fighting equipment for each fire area and familiarization with the layout of the plant, including access and egress routes to each area.

(5) The proper use of available fire fighting equipment and the correct method of fighting each type of fire. The types of fires covered should include fires in energized electrical equipment, fires in cables and cable trays, hydrogen fires, fires involving flammable and combustible liquids or hazardous process chemicals, fires resulting from construction or modifications (welding), and record file fires.

(6) The proper use of communication, lighting, ventilation, and emergency breathing equipment.

(7) The proper method for fighting fires inside buildings and confined spaces.

(8) The direction and coordination of the fire fighting activities (fire brigade leaders only).

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(9) Detailed review of fire fighting strategies and procedures.

(10) Review of the latest plant modifications and corresponding changes in fire fighting plans.

Note: Items (9) and (10) may be deleted from the training of no more than two of the non-operations personnel who may be assigned to the fire brigade.

b. The instruction shall be provided by qualified individuals who are knowledgeable, experienced, and suitably trained in fighting the types of fires that could occur in the plant and in using the types of equipment available in the nuclear power plant.

c. Instruction shall be provided to all fire brigade members and fire brigade leaders.

d. Regular planned meetings shall be held at least every 3 months for all brigade members to review changes in the fire protection program and other subjects as necessary.

e. Periodic refresher training sessions shall be held to repeat the classroom instruction program for all brigade members over a two- year period. These sessions may be concurrent with the regular planned meetings.

2. Practice Practice sessions shall be held for each shift fire brigade on the proper method of fighting the various types of fires that could occur in a nuclear power plant. These sessions shall provide brigade members with experience in actual fire extinguishment and the use of emergency breathing apparatus under strenuous conditions encountered in fire fighting.

These practice sessions shall be provided at least once per year for each fire brigade member.

3. Drills

a. Fire brigade drills shall be performed in the plant so that the fire brigade can practice as a team.

b. Drills shall be performed at regular intervals not to exceed 3 months for each shift fire brigade. Each fire brigade member should participate in each drill, but must participate in at least two drills per year.

A sufficient number of these drills, but not less than one for each shift fire brigade per year, shall be unannounced to determine the fire fighting readiness of the plant fire brigade, brigade leader, and fire protection systems and equipment. Persons planning and authorizing an unannounced drill shall ensure that the responding shift fire brigade members are not aware that a drill is being planned until it is begun. Unannounced drills shall not be scheduled closer than four weeks.

At least one drill per year shall be performed on a "back shift" for each shift fire brigade.

c. The drills shall be preplanned to establish the training objectives of the drill and shall be critiqued to determine how well the training objectives have been met. Unannounced drills shall be planned and critiqued by members of the management staff responsible for plant safety and fire protection. Performance deficiencies of a fire brigade or of individual fire brigade members shall be remedied by scheduling additional training for the brigade or members. Unsatisfactory drill performance shall be followed by a repeat drill within 30 days.

d. At 3-year intervals, a randomly selected unannounced drill must be critiqued by qualified individuals independent of the licensee's staff. A copy of the written report from these individuals must be available for NRC review and shall be retained as a record as specified in section III.I.4 of this appendix.

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e. Drills shall as a minimum include the following:

(1) Assessment of fire alarm effectiveness, time required to notify and assemble fire brigade, and selection, placement and use of equipment, and fire fighting strategies.

(2) Assessment of each brigade member's knowledge of his or her role in the fire fighting strategy for the area assumed to contain the fire. Assessment of the brigade member's conformance with established plant fire fighting procedures and use of fire fighting equipment, including self-contained emergency breathing apparatus, communication equipment, and ventilation equipment, to the extent practicable.

(3) The simulated use of fire fighting equipment required to cope with the situation and type of fire selected for the drill. The area and type of fire chosen for the drill should differ from those used in the previous drill so that brigade members are trained in fighting fires in various plant areas. The situation selected should simulate the size and arrangement of a fire that could reasonably occur in the area selected, allowing for fire development due to the time required to respond, to obtain equipment, and organize for the fire, assuming loss of automatic suppression capability.

(4) Assessment of brigade leader's direction of the fire fighting effort as to thoroughness, accuracy, and effectiveness.

4. Records Individual records of training provided to each fire brigade member, including drill critiques, shall be maintained for at least 3 years to ensure that each member receives training in all parts of the training program. These records of training shall be available for NRC review.

Retraining or broadened training for fire fighting within buildings shall be scheduled for all those brigade members whose performance records show deficiencies.

J. Emergency lighting. Emergency lighting units with at least an 8-hour battery power supply shall be provided in all areas needed for operation of safe shutdown equipment and in access and egress routes thereto.

K. Administrative controls. Administrative controls shall be established to minimize fire hazards in areas containing structures, systems, and components important to safety.

These controls shall establish procedures to:

1. Govern the handling and limitation of the use of ordinary combustible materials, combustible and flammable gases and liquids, high efficiency particulate air and charcoal filters, dry ion exchange resins, or other combustible supplies in safety-related areas.

2. Prohibit the storage of combustibles in safety-related areas or establish designated storage areas with appropriate fire protection.

3. Govern the handling of and limit transient fire loads such as combustible and flammable liquids, wood and plastic products, or other combustible materials in buildings containing safety-related systems or equipment during all phases of operating, and especially during maintenance, modification, or refueling operations.

4. Designate the onsite staff member responsible for the inplant fire protection review of proposed work activities to identify potential transient fire hazards and specify required additional fire protection in the work activity procedure.

5. Govern the use of ignition sources by use of a flame permit system to control welding, flame cutting, brazing, or soldering operations. A separate permit shall be issued for each area where work is to be done. If work continues over more than one shift, the permit shall 97

be valid for not more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the plant is operating or for the duration of a particular job during plant shutdown.

6. Control the removal from the area of all waste, debris, scrap, oil spills, or other combustibles resulting from the work activity immediately following completion of the activity, or at the end of each work shift, whichever comes first.

7. Maintain the periodic housekeeping inspections to ensure continued compliance with these administrative controls.

8. Control the use of specific combustibles in safety-related areas. All wood used in safety-related areas during maintenance, modification, or refueling operations (such as lay-down blocks or scaffolding) shall be treated with a flame retardant. Equipment or supplies (such as new fuel) shipped in untreated combustible packing containers may be unpacked in safety-related areas if required for valid operating reasons. However, all combustible materials shall be removed from the area immediately following the unpacking. Such transient combustible material, unless stored in approved containers, shall not be left unattended during lunch breaks, shift changes, or other similar periods. Loose combustible packing material such as wood or paper excelsior, or polyethylene sheeting shall be placed in metal containers with tight-fitting self-closing metal covers.

9. Control actions to be taken by an individual discovering a fire, for example, notification of control room, attempt to extinguish fire, and actuation of local fire suppression systems.

10. Control actions to be taken by the control room operator to determine the need for brigade assistance upon report of a fire or receipt of alarm on control room annunciator panel, for example, announcing location of fire over PA system, sounding fire alarms, and notifying the shift supervisor and the fire brigade leader of the type, size, and location of the fire.

11. Control actions to be taken by the fire brigade after notification by the control room operator of a fire, for example, assembling in a designated location, receiving directions from the fire brigade leader, and discharging specific fire fighting responsibilities including selection and transportation of fire fighting equipment to fire location, selection of protective equipment, operating instructions for use of fire suppression systems, and use of preplanned strategies for fighting fires in specific areas.

12. Define the strategies for fighting fires in all safety-related areas and areas presenting a hazard to safety-related equipment. These strategies shall designate:

a. Fire hazards in each area covered by the specific prefire plans.

b. Fire extinguishants best suited for controlling the fires associated with the fire hazards in that area and the nearest location of these extinguishants.

c. Most favorable direction from which to attack a fire in each area in view of the ventilation direction, access hallways, stairs, and doors that are most likely to be free of fire, and the best station or elevation for fighting the fire. All access and egress routes that involve locked doors should be specifically identified in the procedure with the appropriate precautions and methods for access specified.

d. Plant systems that should be managed to reduce the damage potential during a local fire and the location of local and remote controls for such management (e.g., any hydraulic or electrical systems in the zone covered by the specific fire fighting procedure that could increase the hazards in the area because of overpressurization or electrical hazards).

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e. Vital heat-sensitive system components that need to be kept cool while fighting a local fire. Particularly hazardous combustibles that need cooling should be designated.

f. Organization of fire fighting brigades and the assignment of special duties according to job title so that all fire fighting functions are covered by any complete shift personnel complement. These duties include command control of the brigade, transporting fire suppression and support equipment to the fire scenes, applying the extinguishant to the fire, communication with the control room, and coordination with outside fire departments.

g. Potential radiological and toxic hazards in fire zones.

h. Ventilation system operation that ensures desired plant air distribution when the ventilation flow is modified for fire containment or smoke clearing operations.

i. Operations requiring control room and shift engineer coordination or authorization.

j. Instructions for plant operators and general plant personnel during fire.

L. Alternative and dedicated shutdown capability.

1. Alternative or dedicated shutdown capability provided for a specific fire area shall be able to (a) achieve and maintain subcritical reactivity conditions in the reactor; (b) maintain reactor coolant inventory; (c) achieve and maintain hot standby conditions for a PWR (hot shutdown for a BWR); (d) achieve cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and (e) maintain cold shutdown conditions thereafter. During the postfire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of normal a.c. power, and the fission product boundary integrity shall not be affected; i.e., there shall be no fuel clad damage, rupture of any primary coolant boundary, of rupture of the containment boundary.

2. The performance goals for the shutdown functions shall be:

a. The reactivity control function shall be capable of achieving and maintaining cold shutdown reactivity conditions.

b. The reactor coolant makeup function shall be capable of maintaining the reactor coolant level above the top of the core for BWRs and be within the level indication in the pressurizer for PWRs.

c. The reactor heat removal function shall be capable of achieving and maintaining decay heat removal.

d. The process monitoring function shall be capable of providing direct readings of the process variables necessary to perform and control the above functions.

e. The supporting functions shall be capable of providing the process cooling, lubrication, etc., necessary to permit the operation of the equipment used for safe shutdown functions.

3. The shutdown capability for specific fire areas may be unique for each such area, or it may be one unique combination of systems for all such areas. In either case, the alternative shutdown capability shall be independent of the specific fire area(s) and shall accommodate postfire conditions where offsite power is available and where offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Procedures shall be in effect to implement this capability.

4. If the capability to achieve and maintain cold shutdown will not be available because of fire damage, the equipment and systems comprising the means to achieve and maintain the hot standby or hot shutdown condition shall be capable of maintaining such conditions until cold shutdown can be achieved. If such equipment and systems will not be capable of being powered by both onsite and offsite electric power systems because of fire damage, an 99

independent onsite power system shall be provided. The number of operating shift personnel, exclusive of fire brigade members, required to operate such equipment and systems shall be on site at all times.

5. Equipment and systems comprising the means to achieve and maintain cold shutdown conditions shall not be damaged by fire; or the fire damage to such equipment and systems shall be limited so that the systems can be made operable and cold shutdown can be achieved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Materials for such repairs shall be readily available on site and procedures shall be in effect to implement such repairs. If such equipment and systems used prior to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the fire will not be capable of being powered by both onsite and offsite electric power systems because of fire damage, an independent onsite power system shall be provided. Equipment and systems used after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be powered by offsite power only.

6. Shutdown systems installed to ensure postfire shutdown capability need not be designed to meet seismic Category I criteria, single failure criteria, or other design basis accident criteria, except where required for other reasons, e.g., because of interface with or impact on existing safety systems, or because of adverse valve actions due to fire damage.

7. The safe shutdown equipment and systems for each fire area shall be known to be isolated from associated non-safety circuits in the fire area so that hot shorts, open circuits, or shorts to ground in the associated circuits will not prevent operation of the safe shutdown equipment. The separation and barriers between trays and conduits containing associated circuits of one safe shutdown division and trays and conduits containing associated circuits or safe shutdown cables from the redundant division, or the isolation of these associated circuits from the safe shutdown equipment, shall be such that a postulated fire involving associated circuits will not prevent safe shutdown. An acceptable method of complying with this alternative would be to meet Regulatory Guide 1.75 position 4 related to associated circuits and IEEE Std 384-1974 (Section 4.5) where trays from redundant safety divisions are so protected that postulated fires affect trays from only one safety division.

M. Fire barrier cable penetration seal qualification. Penetration seal designs must be qualified by tests that are comparable to tests used to rate fire barriers. The acceptance criteria for the test must include the following:

1. The cable fire barrier penetration seal has withstood the fire endurance test without passage of flame or ignition of cables on the unexposed side for a period of time equivalent to the fire resistance rating required of the barrier;

2. The temperature levels recorded for the unexposed side are analyzed and demonstrate that the maximum temperature is sufficiently below the cable insulation ignition temperature; and

3. The fire barrier penetration seal remains intact and does not allow projection of water beyond the unexposed surface during the hose stream test.

N. Fire doors. Fire doors shall be self-closing or provided with closing mechanisms and shall be inspected semiannually to verify that automatic hold-open, release, and closing mechanisms and latches are operable. One of the following measures shall be provided to ensure they will protect the opening as required in case of fire:

1. Fire doors shall be kept closed and electrically supervised at a continuously manned location;

2. Fire doors shall be locked closed and inspected weekly to verify that the doors are in the closed position; 100

3. Fire doors shall be provided with automatic hold-open and release mechanisms and inspected daily to verify that doorways are free of obstructions; or

4. Fire doors shall be kept closed and inspected daily to verify that they are in the closed position. The fire brigade leader shall have ready access to keys for any locked fire doors.

Areas protected by automatic total flooding gas suppression systems shall have electrically supervised self-closing fire doors or shall satisfy option 1 above.

O. Oil collection system for reactor coolant pump. The reactor coolant pump shall be equipped with an oil collection system if the containment is not inerted during normal operation. The oil collection system shall be so designed, engineered, and installed that failure will not lead to fire during normal or design basis accident conditions and that there is reasonable assurance that the system will withstand the Safe Shutdown Earthquake. See Regulatory Guide 1.29"Seismic Design Classification" paragraph C.2.

Such collection systems shall be capable of collecting lube oil from all potential pressurized and unpressurized leakage sites in the reactor coolant pump lube oil systems. Leakage shall be collected and drained to a vented closed container that can hold the entire lube oil system inventory. A flame arrester is required in the vent if the flash point characteristics of the oil present the hazard of fire flashback. Leakage points to be protected shall include lift pump and piping, overflow lines, lube oil cooler, oil fill and drain lines and plugs, flanged connections on oil lines, and lube oil reservoirs where such features exist on the reactor coolant pumps. The drain line shall be large enough to accommodate the largest potential oil leak.

[45 FR 76611, Nov. 19, 1980; 46 FR 44735, Sept. 8, 1981, as amended at 53 FR 19251, May 27, 1988; 65 FR 38191, June 20, 2000; 77 FR 39907, Jul. 6, 2012]

3.14.2 New Reactors New reactor designs integrate fire protection requirements, including the protection of safe shutdown capability and the prevention of radiological release, into the planning and design phase for the plant. In addition, new reactor designs should minimize or eliminate the use of alternative/dedicated shutdown systems and should only rely on such systems when it is not feasible to provide the required protection for redundant safe shutdown systems, such as in the main control room.

3.14.2.1 Overview

1. Criterion 3, Fire protection, of Appendix A, to 10 CFR Part [NRC-01]

a. Criterion GDC 3 addresses fire protection requirements and specifies, in part, that (1) systems, structures and components (SSCs) important to safety must be designed and located to minimize the probability and effects of fires and explosions, (2) non-combustible and heat-resistant materials must be used wherever practical, and (3) fire detection and suppression systems must be provided to minimize the adverse effects of fires on SSCs important to safety.

2. 10 CFR 50.48(a) [NRC-04]

a. The fire protection program for new reactor plants are subject to 10 CFR 50.48(a) and the criteria for enhanced fire protection.

3. Regulatory Guide 1.189 [NRC-07]

a. Enhanced Fire Protection Criteria: New reactor designs should ensure that safe shutdown can be achieved assuming that all equipment in any one fire area will be rendered inoperable by fire and that re-entry into the fire area for repairs and operator actions is not possible. The control room should be evaluated to ensure that the effects of fire do not adversely affect the ability to achieve and maintain safe shutdown. New 101

reactors should provide fire protection for redundant shutdown systems in the reactor containment building that will ensure, to the extent practicable, that one shutdown division will be free of fire damage. Additionally, new reactor designs should ensure that smoke, hot gases, or the fire suppressant will not migrate into other fire areas to the extent that they could adversely affect safe-shutdown capabilities, including operator actions

b. The enhanced fire protection criteria for advanced light-water reactor designs in Commission Papers (SECY) SECY 90-016 (ML003707849), SECY 93-087 (ML003708021), and SECY 94-084 (ML003708068). SECY 90-016 establishes enhanced fire protection criteria for the evolutionary advanced light-water reactor designs to ensure that post-fire safe-shutdown can be achieved assuming all cable and equipment in any one fire area is rendered inoperable as a result of fire damage and that re-entry into the fire area by plant personnel for repairs or operator actions is not possible. Passive Plant Safe Shutdown Condition: The passive decay heat removal systems should be capable of achieving and maintaining 215.6ºC (420ºF) or below for non-loss of coolant accident (LOCA) events. This safe shutdown condition is predicated on demonstration of acceptable passive safety system performance.

3.14.3 Supplementary Information

[NRC-01] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 Fire Protection, August 2007, Appendix A to Part 50 - General Design Criteria for Nuclear Power Plants, (updated) January 1, 2011, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appa.html.

[NRC-02] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 Fire Protection, Appendix R to 10 CFR Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979", updated 2017, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appr.html.

[NRC-04] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 Fire Protection, August 2007, http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0048.html.

[NRC-05] United States Nuclear Regulatory Commission (NRC): Appendix A to Branch Technical Position (BTP) APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants, 1976, http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/,

br0361/s1/apcsb95-1.pdf.

[NRC-07] United States Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation: Fire Protection for Nuclear Power Plants, Regulatory Guide 1.189, Rev. 2, Washington, DC, USA, October 2009, https://www.nrc.gov/docs/ML0925/ML092580550.pdf.

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4 REGULATORY QUESTIONAIRE 4.1 General Comparisons In order to provide a uniform comparison between the high-level regulations on a country by country basis, a simplified questionnaire intended to capture the larger themes and prevailing practices of many countries with regard to fire safety regulatory schemes was provided to the National Coordinator(s) of each country. The four questions asked and the respective responses from the 14 FIRE Database Project member countries are provided in the following paragraphs.

4.2 Regulation Questionnaire Form The following survey has been answered by all FIRE Database Project member countries in order to document the regulatory information in a simple manner.

4.2.1 Questions about the use of FIRE PSA (PRA) in FIRE member countries Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

Q.2 Are Fire PRAs used to support license applications?

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

4.2.2 High Level Questionnaire Summary Results Table 6 FIRE Member Countries - High Level Questionnaire Summary Results Country Prescriptive / PSA Supports Licensing Methodology Risk-Informed Actions Framework Belgium Prescriptive & Yes NUREG/CR-6850 risk-informed [NRC-08]

Canada Prescriptive & Yes NUREG/CR-6850 risk-informed [NRC-08]

Czech Republic Prescriptive & Yes NUREG/CR-6850 risk-informed [NRC-08]

Finland Prescriptive & Yes Partially applied risk-informed NUREG/CR-6850

[NRC-08]

France Prescriptive & Yes INB order [FRA-02]

Risk-Informed Germany Prescriptive & Yes German PSA Guide risk-informed; [BMU-05] and however, Supporting Technical Documents on PSA Methods and Data 103

Country Prescriptive / PSA Supports Licensing Methodology Risk-Informed Actions Framework traditionally mainly [FAK-05], [FAK-05a],

prescriptive [FAK-16]

Japan Prescriptive No NUREG/CR-6850

[NRC-08]

Korea Prescriptive No NUREG/CR-6850

[NRC-08]

The Netherlands Prescriptive & Yes NUREG/CR-6850 risk-informed [NRC-08]

Spain Prescriptive & Only under performance- NUREG/CR-6850 risk-informed based methodology [NRC-08]

accepted as a valid way to meet fire protection regulation requirements Sweden Prescriptive Yes Not directly, but influenced by NUREG/CR-6850

[NRC-08]

Switzerland Prescriptive & Yes ENSI-A05 [ENSI-18a]

risk-informed United Kingdom Prescriptive & To a limited extent; ONR SAPs and the risk-informed for new reactors, Fire PSA ONR TAG on PSA, would be expected prior to NS-TAST-GD-030 operation (TAG-030) -

Probabilistic Safety Analysis [ONR-16]

United States Prescriptive & Yes NUREG/CR-6850 risk-informed [NRC-08]

4.3 Survey Responses 4.3.1 Belgium Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

Both; Royal Decree (RD) of 30/11/2011 transposes the WENRA 2008 Safety Reference Levels in the Belgian regulatory framework. Art. 17.3 of the RD requires that a deterministic fire risk analysis has to be performed in order to demonstrate that safety objectives are met, fire protection systems are properly designed and that all administrative provisions were correctly identified. In addition, Art. 32.2 requires that a fire probabilistic risk analysis (Fire PRA) to be performed to complement the deterministic risk analysis, for power reactors only.

Q.2 Are Fire PRAs used to support license applications?

Because new nuclear installation must comply with the Royal Decree of 30/11/2011, Fire PSA can be considered as a requirement for a license application of new nuclear power reactors 104

(Fire PSA is not required for other type on nuclear installations, such as research reactors).

However, there is currently no plan for a license application of nuclear power reactor.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

The Belgian regulatory body, nuclear safety authority and TSO, ensures an on-line review of the execution of the Fire PSA study, this includes establishment of the methodology, compliance with the guidance and justification of potential deviations, as well as the analysis and discussion of the results, including CDF quantification and physical improvement to the fire safety of the installation.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

NUREG/CR-6850 is used as the methodological guidance for Fire PSA in Belgium. This is not enforced by the Belgian regulation but acknowledged by the regulatory body as the state-of-the-art methodology to be followed. Licensees may propose some deviation from this methodological framework, which has to be reviewed and accepted by the regulatory body.

4.3.2 Canada Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

In Canada the regulations for fire protection is based on prescriptive (deterministic) and risk-informed assessment principals. The licensee Power Reactor Operating Licence (PROL) includes a license condition under Safety and Control Area of Emergency Management and Fire Protection: The licensee shall implement and maintain a fire protection program. The fire Protection program is required to meet Canadian Standards Association (CSA) N293-12 Fire protection for nuclear power plants [CSA-12]. CSA N293 can be thought of as an objective based standard (consistent with the Canadian Building and Fire code approach). CSA N293 permits the use of performance-based approaches but does not prescribe in similar fashion to NFPA 805 [NFPA-01] a risk-based fire protection program. CSA N293 establishes the fire protection requirements for the design construction, commissioning, operation, and decommissioning of nuclear power plants to address the fire protection goals and objectives.

CSA N293 requirements include:

- Design requirements (e.g., fire detection and alarm system, fire suppression, fire resistance rating of building structures, building materials, egress);

- Operational requirements (e.g., control of ignition sources, ITM of fire protection features, control of flammable, combustible materials);

- Fire Protection Program requirements;

- Fire safety assessment requirements (e.g., code compliance, fire hazard assessments, fire safe shutdown analysis); and

- Fire response and decommissioning.

Q.2 Are Fire PRAs used to support license applications?

Yes. The licensee Power Reactor Operating Licence (PROL) includes a license condition with regard to Safety Analysis and the licensees are required to comply with regulatory document REGDOC-2.4.2 Probabilistic Safety Assessment (PSA) for Nuclear Power Plants [CNSC-14a].

REGDOC-2.4.2 requires the PSA to include both internal and external events.

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Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

PSA for internal and external events are required to meet the requirement of the regulatory document REGDOC-2.4.2 Probabilistic Safety Assessment (PSA) for Nuclear Power Plants

[CNSC-14a]. This regulatory document requires the licensees to seek CNSC acceptance of the PSA methodology before the conduct of the PSA. This provides CNSC staff assurance that the PSA is developed according to an accepted methodology. In addition, CNSC staff use ASME/ANS PRA Standard, CSA N290.17.17 Probabilistic safety assessment for nuclear power plants [CSA-17], NUREG (e.g., NUREG/CR-6850 [NRC-08] for fire) IAEA documents (such as SSG-3 [IAEA-10], SSG-4 [IAEA-10a], IAEA-TECDOC-1135 [IAEA-00d], and IAEA-TECDOC-1229 [IAEA-01]) to review the quality of Fire PSA.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

The methodology used for Fire PSA for the Canadian nuclear power plants is generally developed based on NUREG/CR-6850 methodology [NRC-08]. The limitation is the applicability of fire data/ fire initiating event frequencies, definition of fire zones given the difference in CANDU design and layout, fire scenario modelling (e.g., oil, hydrogen, multi-cable tray scenarios) and the whole site multi-unit PSA issues.

4.3.3 Czech Republic Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

Until 2016, fire legislation was based on prescriptive principals. The new Atomic Act - Act No.

263/2016 Coll and related SÚJB Decree No. 162/2016 Coll., Requirements for safety assessment under the Atomic Act require the use the fire protection regulations based both on prescriptive and probabilistic methods.

Q.2 Are Fire PRAs used to support license applications?

Yes, complete Level 1 and limited scope Level 2 PSA (Fire PSA is included like internal hazards) for power operation as well as for low power and shutdown plant operational states) or applications are used in the frame of regulatory oversight in addition to deterministic analyses.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

PSA is reviewed by independent review team during the Periodic Safety Reviews (PSR). The methodology and criteria for chapter 6 (PSA evaluation) were developed in UJV Rez. PSR is performed every ten years (PSR 30 - last revision at the Dukovany NPP was made in 2015 and PSR 20 - last revision at the Temelin NPP was made in 2018).

At the Dukovany NPP the IPERS mission took place in 1998, focusing on first level PSA study, in order to assess the study and propose specific proposals for its improvement. At present, this activity builds IAEA TSR-PSA review team. Last mission was performed at NPP Dukovany in 2016. Some recommendations were also concerned to quality and scope of Fire PSA. All recommendations were analyzed in detail and adopted recommendations were included into the PSA model and documents.

At the Temelin NPP a mission on the PSA study took place in 1995 and 1996. The mission concluded that Temelín NPP carefully adopted PSA methodology and the results confirmed a high level of power plant safety in spite of conservative assumptions. In 2003, the IPSART mission re-examined the previous verifications and focused in detail on updated models of 106

probabilistic safety assessment of the current design and operation of the power plant. The visit of IAEA TSR-PSA review team is in plan.

In addition, the Czech regulatory body, NPPs and TSO ensure an on-line review of the execution of the Fire PSA study, this includes establishment of the methodology, compliance with the guidance.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

Fire PSA is implemented like one of internal hazard into PSA by UJV Rez. The used approach is based on NUREG/CR-6850 [NRC-08]. For some aspects (for example - quantification of spurious signals probability) the support of another guidelines is partially used (NUREG/CR-7150 [NRC-16], NEI 00-01 [NEI-05].

The issues we have could be the effective selection of the most important combination of spurious signals (multi-cable tray scenarios) leading to inadvertent behavior of equipment (how to cut off most insignificant combinations and still stay on the conservative side), smoke impact on electronics, uncertainty of fires in the I&C cabinet rooms and MCR (only the spatial separation of redundant lines), modelling of large fires in one compartment (oil fire in turbine hall) and influence on steel construction (pillars, beams), hot gas layer phenomena.

4.3.4 Finland Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

In Finland the regulations for fire protection is based on prescriptive (deterministic) principles.

STUK YVL Guide B.8 [FIN-19].

- It shall be demonstrated by means of the containment fire hazard analysis that, despite containment fires, the reactor can be shut down and cooled, and residual heat can be removed without compromising containment integrity.

- It shall be demonstrated by a fire hazard analysis of the control room that control of the necessary safety functions can be executed in the event of a fire in the control room or in any other fire compartment.

- In connection with the design of the I&C systems of the nuclear power plant, the influence of fires on the functioning of safety significant I&C systems shall be analyzed, including the effects of fire-induced temperature rise and combustion gases on equipment and the reflection of disturbances and failures thereof on the execution of safety functions.

Q.2 Are Fire PRAs used to support license applications?

Yes, per STUK YVL Guide A.7 (PRA) [FIN-20] PRA is mandatory (full scope PRA Level 1 &

2; power operation as well as low power and shutdown modes).

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

International review guidelines were partially utilized, when STUK developed own review guideline for the whole PRA. The electronic PRA models are also provided to STUK for review and to be able to perform quantification by STUK (e.g. sensitivity and uncertainty analyses).

Under the review of Fire-PRAs, small and medium scale fire research has been ordered by a TSO (VTT) especially considering cable fires and I&C cabinet fires. Fire simulations have also been ordered by VTT considering cable fires, I&C cabinet fires and large oil pool fires. The 107

outcome of VTTs work has been taken into account when updating Fire PRAs. As stated in the YVL Guide A.7 (items 501, 506, and 508):

• STUK oversees the licensees risk management by reviewing the associated documents, models, analyses, guidelines and applications and by performing verification analyses. STUK makes inspection visits to nuclear power plants and organisations involved in the implementation of the PRA. STUK may commission work supporting the review of the PRA from external expert organisations.

• STUK reviews in the extent necessary the updates to the PRA and its applications submitted for information during the nuclear power plants operation and, where necessary, makes a decision on them. Always in connection with a Periodic Safety Review, STUK performs an extensive review of the adequacy of the PRA and its applications.

• In reviewing the PRA and its applications, STUK qualitatively and quantitatively assesses the adequacy of the quality and scope of the power utilitys PRA and its applications.

- The qualitative review assesses whether the data, methods and their results are justified and acceptable and checks the modelling of i.e. initiating events, safety systems, auxiliary systems and operator actions.

- The quantitative review assesses the most important numerical results, the computation of accident sequences and the associated uncertainty and sensitivity analyses.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

Fire PRA was started in Finland in the 1980s utilizing available international fire data and some foreign Fire PRAs as examples. EPRI/NUREG reports (fire frequencies) have been partially utilized when the Fire PRAs have been developed further through the years. In most cases conservative assumptions are applied: fire may impact all vulnerable equipment inside the compartment (no further modelling). Some expert judgements on fire propagation (also utilizing fire event data or experimental fire test results) have been applied. Fire simulations have been done for the most important compartments only (e.g. cable fires, I&C cabinet fires and large oil pool fire in the turbine hall).

Especially severe fire events around the world are important to realize possible consequences of fires and to avoid optimistic assumptions/limitations (e.g. to realize possible scope of direct, indirect and consequential failures).

STUK does not specify methodologies to be used for PRA.

Smoke impact on electronics is a potential uncertainty issue considering fires in the I&C cabinet rooms.

4.3.5 France Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The order laying down general rules for Basic Nuclear Installation (BNI), said Order of 7 February, 2012 setting the general rules relative to basic nuclear installations [FRA-02],was published in the Official Journal on 7 February 2012. It is a major element of this approach. This includes in particular the French law rules corresponding to international best 108

practices. The provisions of the order BNI are mainly dealing in the organization and responsibilities of the Basic Nuclear Installation operators, demonstration nuclear safety, nuisance control and their impact on health and the environment, waste management and the preparation and management of emergency situations.

The ASN resolution No 2014-DC-0417 of 28th January 2014 concerning the rules applicable to BNI with regard to management of fire risks [FRA-03] concerning the rules applicable to BNI with regard to management of fire risks is applicable since 1st July 2014.

Q.2 Are Fire PRAs used to support license applications?

Yes, in France, the licensee may use PSA to complete the nuclear safety demonstration, but he must have first a deterministic approach. By instance, EDF has used PSA to support his demonstration about the EPR.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

EDF Fire PSA is reviewed by IRSN during the Periodic Safety Reviews. IRSN also develops independent focused Fire PSA, which allows to compare results and main assumptions.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

EDF and IRSN Fire PSA methods are inspired by NUREG/CR-6850 [NRC-08] recommended methods. Specific methodological aspects were adapted by IRSN/EDF in order to better fit specific needs (for example to focus the Fire PSA on specific aspects, human errors evaluation taking into account fire procedures; etc.) or operating experience (for example, fire spreading probability assessment, cable fire initiation, etc.).

4.3.6 Germany Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The NPP fire protection regulations in Germany are both, however traditionally the regulation is mainly prescriptive.]

The most recent German nuclear regulations, the so-called Safety Requirements for Nuclear Power Plants [BMU-15] covering also (in a separate Appendix) internal hazards such as plant internal fires are in general prescriptive based on deterministic safety principles. However, these requirements also include probabilistic safety assessment supplementing the deterministic assessment. The subordinate nuclear fire protection standards KTA-2101, Part 1 to 3 [KTA-15],

[KTA-15a], [KTA-15b] are again mainly prescriptive, but use probabilistic arguments and evidence given by risk-informed approaches.

Q.2 Are Fire PRAs used to support license applications?

Yes, either complete Level 1 Fire PSA (for power operation as well as for low power and shutdown plant operational states) or application case based probabilistic studies are used in the frame of regulatory oversight in addition to deterministic analyses. The main benefit of such analyses is to compare e.g. the existing situation in a plant and to demonstrate if and in how far a proposed modification (not only of plants SSC but also of procedures or both) will increase or decrease the plant safety with respect to fires.

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Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

On behalf of the local state regulatory body in charge of nuclear oversight of the plant under investigation experts from the corresponding technical expert/safety organization(s), such as TÜV and/or GRS review the Fire PSA carried out by the licensee or its consultants in detail, performing as far as necessary also own calculations. These reviews are based on the German PSA Guide [BMU-05] and its technical supplements on PSA methods and data [FAK-05], [FAK-05a], [FAK-16] considering also international guidance, e.g. from IAEA [IAEA-10].

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

For Fire PSA to be carried out in Germany, in particular in the frame of the Periodic Safety Reviews (PSR), the German PSA Guide [BMU-05] and its Supporting Technical Documents on PSA Methods and Data [FAK-05], [FAK-05a], [FAK-16] are used providing state-of-the-art guidance on Fire PSA methods and data to be used. These guidance documents do not provide detailed guidance in how far detailed fire modelling needs to be performed and which fire simulation codes are suitable for the scenarios to be investigated in detail. However, uncertainty and sensitivity analyses have to be carried out, e.g. by the GRS code SUSA, in order to give indications on the level of conservatism of the Fire PSA.

4.3.7 Japan Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The regulations are prescriptive. NRA has established the NRA Ordinance on Standards for the Location, Structures and Equipment of Commercial Power Reactors Article 8 (Design Considerations against Fire) [JPN-04] for fire protection. As for Article 8, the nuclear reactor facilities shall be designed such that their safety will not be impaired by fire considering protective measures for preventing, detecting and fire suppression, and mitigating its effect, independently. These protective measures shall also be designed such as not to impair the required functions of SSCs with safety functions as a result of their failure or malfunction.

Q.2 Are Fire PRAs used to support license applications?

No. Fire PRAs are not used.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

Fire PRAs are not used to support the license amendment.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

Examples of guidance are provided in the following; however, how to use it is not yet decided at present.

- Fire PRA Guide for Japanese nuclear industry developed by Nuclear Risk Research Center (NRRC) of Central Research Institute of the Electric Power Industry (CRIEPI),

- Implementation standard concerning the internal fire probabilistic risk assessment of nuclear power plants developed by the Atomic Energy Society of Japan,

- EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities (NUREG/CR-6850, EPRI 101-1989, Volumes 1 and 2), developed by U.S. NRC [NRC-08].

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4.3.8 Korea Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The NPP fire protection regulations in Korea are prescriptive.

Fire protection regulations for NPPs in Korea are based on a prescriptive regulatory guideline, KINS/RG-N10.06 (Fire protection in Nuclear Power Plants) [KOR-06], which includes comprehensive requirements related to fire protection plan with fire hazard / fire safe shutdown analysis. NFPA 803 [NFPA-98] and NFPA 804 [NFPA-15] are used as a regulatory requirement

/plant-specific license condition by KINS/RG-N10.06 [KOR-06]. After approval of implementing NFPA 805 [NFPA-01], one-point exemption for a specific requirement of fire protection is accepted by regulatory body, Nuclear Safety and Secure Commission, using performance-based risk-informed methods, NFPA 805.

Q.2 Are Fire PRAs used to support license applications?

The results or insight of Fire PRA are not yet used to support license applications in Korea, but Fire PRA are only performed as a regulatory process of Periodic Safety Review and severe accident policy in Korea.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

The detailed method to ensure the quality of Fire PRA is not yet established in Korea. But international requirements such as NUREG and IAEA documents will be used to establish the guideline for reviewing Fire PRA submittal. Plant walk-down and a few sample verification reviews for Fire PRA model will be included in detailed method in Korea.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

NUREG/CR-6850 (EPRI 1011989) [NRC-08] methods will be used as the baseline methodological guidance in Korea.

4.3.9 The Netherlands Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The NPP fire protection regulations in the Netherlands are both, prescriptive and performance-based. There is a uniform regulatory approach harmonizing all parts of fire protection standards, based mainly on (prescriptive) deterministic requirements also considering probabilistic requirements, for the assessment of reliability.

Q.2 Are Fire PRAs used to support license applications?

Yes; the licensee includes a license condition with regard to Safety Analysis and the licensees are required to comply with IAEA documents for Probabilistic Safety Assessment for Nuclear Power Plants [IAEA-10], [IAEA-10a]. The Borssele NPP has also a full scope Level 3 Probabilistic Safety Assessment (PSA) which includes internal and external hazard initiators.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

Licensees are requested to carry out the Periodic Safety Review (PSR) to incorporate the state-of-the-art knowledge into the plant design, operation and maintenance activities. A review of 111

operational safety aspects must be performed once every two years, whilst a more comprehensive safety review must be conducted once every ten years.

Upon request of ANVS, in-depth international team reviews are also carried out by bodies such as the IAEA (OSART, Fire Safety, IPSART [NED-12], [NED-13], etc.). ANVS carries out inspections or team audits from time to time. In addition, the Borssele nuclear plant itself carries out self-assessments at regular intervals and invites others like WANO to perform assessments.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

NUREG/CR-6850 [NRC-08] is used as the methodological guidance for Fire PSA in the Netherlands. This is not enforced by the regulation but acknowledged by the regulatory body.

Development of probabilistic-logic model and implementation of quantitative calculations are performed using the software WinNUPRA 4.0 [SCI-06] applying event/fault tree methodology.

4.3.10 Spain Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The regulations in Spain are prescriptive/both.

Currently, Safety Instruction IS-30 [SPN-05] is the regulatory standard for fire protection in nuclear power plants and is a deterministic regulation endorsing 10 CFR 50.48 [NRC-04], [NRC-07] requirements. However, the IS-30 itself includes the statement that some of its provisions may be accomplished under risk-informed performance-based analysis carried out under a methodology accepted by the regulator, i.e. the NFPA-805 [NFPA-01]. This approach requires a license amendment application from the licensee.

Q.2 Are Fire PRAs used to support license applications?

Fire PRAs for all operation modes are required in all power stations. Additionally, for licensees who have used these analyses in support of their performance-based applications to meet IS-30

[SPN-05] provisions these PSAs are required to meet the NUREG/CR-6850 methodology

[NRC-08] or at power operating mode.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

The review process is under a step-by-step implementation of standardized quality features, as ASME/ANS RA-Sa-2009 [ANS-09] shows. On top of that, RG 1.200 [NRC-12] is used as a guidance to determine the quality level of the PSA. Additionally, CSN performs an independent review of the Fire PRA. In this process, particular attention is given to the conclusions from the peer-review process from an independent expert panel that is also required to plants transitioning to NFPA 805 [NFPA-01].

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

As answered above, NUREG/CR-6850 [NRC 08] is being adopted as the standard Fire PRA methodology.

4.3.11 Sweden Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The NPP fire protection regulations in Sweden are prescriptive (deterministic).

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SSMFS 2008:17 [SSM-09a-] specifies amongst others that fire protection provisions in nuclear facilities have to ensure that:

- the design of a facility are able to withstand a fire event and its consequences;

- fire events included in the deterministic safety analysis are divided into a limited number of event classes with specified analysis assumptions and acceptance criteria.

SSMFS 2008:1 [SSM-09] specifies amongst others that fire protection provisions in nuclear facilities have to ensure that the capacity of a facilitys barriers and defense-in-depth system to prevent radiological accidents and mitigate the consequences in the event of an accident are analyzed using deterministic methods before the facility is constructed or modified and taken into operation.

Q.2 Are Fire PRAs used to support license applications?

Yes, SSMFS 2008:1 [SSM-09] specifies that in addition to deterministic analyses the facility shall be analyzed using probabilistic methods in order to obtain as comprehensive view as possible of safety. Consequently, fire needs to be included in the scope of the licensees' PSAs.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

SSM reviews the PSAs rather briefly. SSM make sure that the licensees have the necessary conditions (competence, suitability, adequate time and resources etc.) needed to ensure the quality.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

SSM does not require use of a specific guide or method. The used methods are conservative and bounding (room based, all equipment assumed to fail). The licensees state that parts of the Fire PSA methods are influenced by NUREG/CR-6850 [NRC-08].

4.3.12 Switzerland Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The regulations in Switzerland are prescriptive.

Fire protection in nuclear installations in Switzerland is governed by guideline HSK-R-50 Requirements Important to Safety for Fire Protection in Nuclear Installations [HSK-03], and rules on the detailed arrangements of fire protection measures are developed by the cantonal fire insurers. However, Fire PSA is a mandatory part of the Level 1 and Level 2 PSAs for developed for NPPs. The applications of the risk measures developed, include among others, restrictions on their total magnitude and requirements for a balanced risk profile. These provisions limit the permissible risk of fire to nuclear safety, even though they may not specifically target fire.

Q.2 Are Fire PRAs used to support license applications?

There are applications of CDF, FDF and LERF for license amendments such as changes of SSCs or revisions of Technical Specifications with regard to completion times and permissible maintenance configurations during operation. While these do not imply specific requirements regarding fire protection, the Fire PSA does constitute an integral part of the PSAs performed to derive those risk measures, so any effect of fire risk in licensing applications needs to be addressed.

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Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

ENSI reviews the PSAs of the NPPs in their entirety in the context of the PSRs due once in a decade. Where the licensee chooses to implement significant changes in the modelling, these are typically reviewed directly after their submittal. Thus, at the time of submittal of license amendment request only model changes directly related to the proposed amendment need to be reviewed.

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

All PSAs for Swiss NPPs are developed according to guideline ENSI-A05 [ENSI-18a]. As far as development and modelling of scenarios is concerned, licensees are free in their choice of methodology provided that the suggested solution ensures a level of quality as required in the guideline ENSI-A05 [ENSI-18a]. Most of the licensees opt for NUREG/CR-6850 [NRC-08] and the more recent work under the auspices of EPRI and the U.S.NRC to clarify and update the guidance.

4.3.13 United Kingdom Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The NPP fire regulations in the United Kingdom (UK) are both, prescriptive and performance-based.

NFPA 805 [NFPA-01] is not used as a regulatory requirement or for licensing basis in the UK. In general, the UK looks toward both British Standards and to a goal setting risk-informed method for NPP fire protection regulation.

Some aspects of fire safety on an NPP, such as those which relate to conventional fire safety may have some prescriptive elements. Where nuclear fire and conventional fire safety clash, the consequences and risks are compared, and measures put in place to ensure all fire risks are reduced so far as is reasonably practical (SFAIRP). An analogous term, as low as reasonably practicable (ALARP), is frequently used in the nuclear context.

UK regulation is primarily goal setting in nature and requires that the licensee demonstrate ALARP through adoption of relevant good practice (RGP) and any other risk reduction methods as may be applicable.

The primary conventional fire safety legislation for the UK includes the Regulatory Reform (Fire Safety) Order 2005 [UK-06] in England and Wales, the Fire Safety (Scotland) Regulations 2006

[UK-07] and the Fire Safety Regulations (Northern Ireland) 2010 [UK-08].

The Office for Nuclear Regulation (ONR) establishes guidance in the form of published Safety Assessment Principles (SAPs) and Technical Assessment Guides (TAGs).

ONR SAPs expect the use of risk-based approaches but do not prescribe a risk-based fire protection program as specified in NFPA 805 [NFPA-01]. Relevant British Standards, which may be prescriptive in nature, are used as a baseline for conventional fire safety aspects from which RGP is determined.

Q.2 Are Fire PRAs used to support license applications?

To a limited extent: the station safety reports prepared at the time of licensing for the more modern UK NPPs have PSAs (probabilistic safety assessments) that include risk estimates for hazards, including fire. These have been maintained throughout station life.

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The older AGR (advanced gas cooled reactors) include an estimate of the risk from fire and are not modern Fire PSAs, but a probabilistic representation of the deterministic safety case.

Nevertheless, a pilot Fire PSA was carried out for one of the AGRs in the middle 2000s. This was extensive and included cable tracing and confirmation of the fire zones and their interfaces and penetrations. It was carried out against current IAEA Fire PSA guidance. The exercise gave some learning points but was largely confirmatory of the soundness of the deterministic fire safety case.

New build reactors are expected to produce a PSA consistent with modern standards, including a Fire PSA. This is often developed before construction but must be before operation.

In general, evidence that relevant good practice has been adopted with respect to the NUREG guidance, Fire PSA standard or the ONR TAG on PSA, NS-TAST-GD-030 (TAG-030) -

Probabilistic Safety Analysis [ONR-16] must support license applications. Other TAGs which relate to fire aspects include: NS-TAST-GD-013 (TAG-13) - External Hazards [ONR-16a] and NS-TAST-GD-014 (TAG 14) - Internal Hazards [ONR-18a].

It is not the intention of Appendix 1 of TAG-030 to prescribe specific methods and approaches for conducting PSA for NPPs. Dutyholders may choose to use alternative methods to those covered in this Appendix as long as they are shown to lead to equally valid outcomes. In cases where the PSA or specific areas of it have been undertaken using alternative approaches, ONR will review on a case-by-case basis and judge each on its own merits.

The site license conditions give a legal framework which can be drawn on in assessment and are, in general, set out in the form of requiring the licensee to make adequate arrangements, in the interests of safety, to secure certain objectives. The principal license conditions (LCs) relevant to PSA are LC14, LC23, LC27 and LC28.

LC14 requires the licensee to make and implement adequate arrangements for the production and assessment of safety cases. Normally, the licensees safety case will need to contain PSA as well as deterministic analysis.

LC23 requires that the safety case identifies the conditions and limits necessary in the interest of safety. The SAPs, which convey ONRs expectations, state that design basis and beyond design basis analysis should apply an appropriate combination of engineering, deterministic and probabilistic methods.

ONR expects that PSA will contribute to the identification of suitable and sufficient safety mechanisms, devices and circuits, as required by LC27 and provide a significant input for LC28 in identifying plant that may affect safety for which regular, systematic examination, inspection, maintenance and testing will be required.

When preparing modifications to the existing safety cases in areas relevant to fire hazards, licensees have looked at the effect on nuclear risks - in some cases by using their PSA as a tool to show the effect on risks.

The methods and details of analysis that would be acceptable to ONR may be different at various life-cycle stages, as an existing facility may be able to offer operating experience and feedback as an alternative to detailed analysis which is not possible for a new-build facility.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

The ONR carries out an assessment of the Fire PRA submitted to determine whether it meets national and international standards and guidance as identified in Appendix 1 - Assessment Expectations for Review of PSAs for Nuclear Power Plants of ONR TAG-030 [ONR-16] (this 115

includes IAEA and NUREG standards). Additional guidance is provided by ONR TAG-014 -

Internal Hazards [ONR-16a], and TAG-013 - External Hazards [ONR-18a].

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

Methodological guidance is provided in ONR TAG-030 - Probabilistic Safety Analysis [ONR-16],

which includes reference to International Atomic Energy standards and other international guidance. Additional guidance is provided by ONR TAG-014 - Internal Hazards[ONR-16a], and TAG-013 - External Hazards [ONR-18a].

Due to the option of using alternate methods, one challenge is the determination of the equivalency of an alternate method to the listed accepted methods within Appendix 1 -

Assessment Expectations for Review of PSAs for Nuclear Power Plants of TAG-030 [ONR-16].

One issue with some methods is determining initiating event frequencies for different types of fire events. Evidence of incorporation of learning from previous PSAs can be a challenge when determining the adequacy of a Fire PSA.

4.3.14 United States of America Q.1 Are the NPP fire protection regulations in your country based on prescriptive (deterministic) principles or performance-based risk-informed methods?

The NPPs have the option of using either type of fire protection regulations in the United States:

• Prescriptive (deterministic) - The NRC uses regulation to ensure U.S. nuclear power plants are safe. Since the deterministic fire protection requirements were established, no fires have challenged safe shutdown. Every plant must have a fire protection plan that satisfies 10 CFR Part 50, Appendix A [NRC-01], Criterion 3 and 10 CFR 50.48(a) [NRC-09]. The fire protection plan must outline the overall fire protection program, installed fire protection systems, and the means to ensure that the reactor can be safely shutdown in the event of a fire.

Plants that were licensed before January 1, 1979 were also subject to the prescriptive requirements of 10 CFR 50.48(b) and Appendix R [NRC-02]. Plants that were licensed after January 1, 1979 typically followed the same prescriptive requirements to demonstrate conformance with the fire protection regulations. These requirements were based on an assumed serious fire, where generic criteria were established for all plants.

In cases where these requirements were not practical, or there was a more favourable approach to achieving an equivalent level of safety, applicants and licensees sought alternatives, typically through specific exemptions and amendments.

For example, one of the prescriptive requirements related to the fire protection requirements for safe shutdown capability. This regulation requires that one train of systems necessary to achieve and maintain hot shutdown is free of fire damage. The regulation prescribes that the trains will have:

- a three-hour barrier between them,

- 20' of separation, automatic fire suppression, and fire detection, or

- a one-hour barrier between them, automatic fire suppression, and fire detection.

• Risk Informed - In 2004, the NRC amended its fire protection requirements in 10 CFR 50.48 to add 10 CFR 50.48(c) [NRC-09] to allow licensees to adopt, on a voluntary basis, the 2001 edition of the National Fire Protection Association Standard 805, "Performance-Based Standard for Fire Protection for Light-Water Reactor Electric 116

Generating Plants" (NFPA 805 [NFPA-01]), in lieu of their existing fire protection licensing basis. This approach offers plants the opportunity to use a new and scientifically sound way of reducing fire risks further.

NFPA 805 is part of an NRC effort to incorporate risk information into the agency's regulations and enhance safety. The risk-informed performance-based approach considers risk insights as well as other factors to better focus attention and resources on design and operational issues according to their importance to safety. This NFPA 805 risk-informed approach relies on a range of methods, tools, and data that are acceptable to the NRC to support the corresponding risk analysis. It allows licensees to focus their fire protection activities on the areas of greatest risk.

NFPA 805 enables leveraging the state-of-the-art in fire protection evaluation techniques to maintain and enhance safety. It establishes a fundamental fire protection program and design requirements for fire protection systems and features, including prevention, fire detection and suppression, and safe shutdown. It allows nuclear safety performance criteria to be satisfied by using both fire modelling and quantitative fire risk evaluation. By using this approach, resources can be focused on higher risk areas.

Q.2 Are Fire PRAs used to support license applications?

Yes, Fire PRAs are used to support license applications.

Q.3 How does the regulator inspect and ensure the quality of the Fire PRA supporting the license amendment?

The NRC reviews the transition plan and schedules. If the plans and schedules are approved, the licensee then submits a license amendment request (LAR) that requests use of a NFPA 805 licensing basis. The NRC staff reviews the LAR, conducts regulatory audits, requests additional information from licensees, and may hold public meetings to discuss details of the review, and writes a safety evaluation report approving or disapproving the license amendment. Throughout the transition period, the NRC has the ability to grant the licensee enforcement discretion, which means that the licensee will not receive violations for those potential fire protection non-compliances found during the transition that are not of high safety-significance.

The NRC typically allows a licensee up to three years to submit the LAR to transition to NFPA 805 [NFPA-01]. During the transition and subsequent LAR review period, the NRC continues to monitor individual licensee actions to address plant-specific fire protection technical issues through its Reactor Oversight Process (ROP). As of May 2020, 43 of the 96 operating reactor units have transitioned to NFPA 805, and safety evaluations are being finalized for an additional two operating reactor units. (https://www.nrc.gov/reactors/operating/oversight.html).

Q.4 If Fire PRAs are being developed and applied, what methodological guidance is being used and what issues/limitations have been identified, if any, with these methods?

One of the primary tools used to support Fire PRAs in the United States was developed by the NRC Office of Nuclear Regulatory Research (RES) and the Electrical Power Research Institute (EPRI). NUREG/CR-6850/EPRI 1011989, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities published in 2005 [NRC-08] and Supplement 1 to NUREG/CR6850 and EPRI 1011989 published in 2009 [NRC-13] document Fire PSA methods, tools, and data to support risk assessments and discusses methods to perform fire risk analyses. The methodologies presented in these two documents have been enhanced and updated by recent NUREG publications including but not limited to:

- NUREG/CR-7150 (JACQUE-FIRE), EPRI 1026424 - Joint Assessment of Cable Damage and Quantification of Effects from Fire; Volumes 1-3 [NRC-16];

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- NUREG/CR-7010 (CHRISTI-FIRE) - Cable Heat Release, Ignition, and Spread in Tray Installations During Fire [NRC-23];

- NUREG/CR-7100 (DESIREE-Fire) - Direct Current Electrical Shorting in Response to Exposure Fire [NRC-17];

- NUREG/CR-6931 (CAROLFIRE) - Cable Response to Live Fire [NRC-10];

- NUREG-1824 and NUREG-1824 Supplement 1; EPRI 3002002182 - Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications [NRC-24];

- NUREG/CR-7114 - A Framework for Low Power/Shutdown Fire PRA [NRC-19];

- NUREG-1921, EPRI 1023001 - Fire Human Reliability Analysis Guidelines [NRC-18];

- NUREG-1921, EPRI 3002009215, Supplement 1, EPRI-NRC RES Fire Human Reliability Analysis Guidelines;Qualitative Analysis Guidance for Main Control Room Abandonment Scenarios, January 2020;

- NUREG-2169; EPRI 3002002936 - Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database [NRC-22];

- NUREG/CR-7197 (HELEN-FIRE) - Heat Release Rates of Electrical Enclosure Fires, Final Report [NRC-25];

- NUREG-2178, EPRI 3002005578 (RACHELLE-FIRE) - Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume [NRC-30];

- NUREG-2180 (DELORES-VEWFIRE); - Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities

[NRC-27];

- NUREG-1934; EPRI 1023259 - Nuclear Power Plant Fire Modelling Analysis Guidelines (NPP FIRE MAG) [NRC-29];

- NUREG-2178; EPRI 3002016052 (RACHELLE-FIRE) - Refining and Characterizing Heat Release Rates From Electrical Enclosures During Fire Volume 2: Fire Modelling Guidance for Electrical Cabinets, Electric Motors, Indoor Dry Transformers, and the Main Control Board [NRC-30];

- NUREG-2230; EPRI 3002016051 - Methodology for Modelling Fire Growth and Suppression Response for Electrical Cabinet Fires in Nuclear Power Plants [NRC-31];

- NUREG-2232; EPRI 3002015997 - Heat Release Rate and Fire Characteristics of Fuels Representative of Typical Transient Fire Events in Nuclear Power Plants [NRC-32];

- NUREG-2233; EPRI 3002016054 - Methodology for Modelling Transient Fires in Nuclear Power Plant Fire Probabilistic Risk Assessment [NRC-33].

These documents provide methods, tools and data for use in a PRA model however do not become part of a licensing basis until they are used and approved by the AHJ during a license amendment request (LAR) and subsequent reviews prior to the issuance of a safety evaluation report.

Additional guidance on PSA quality is provided in Regulatory Guide 1.174 [NRC-15] and Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated February 2004, [NRC-12]. In 118

addition, The American Nuclear Society (ANS) issued a standard [ANS-09] for evaluating the technical adequacy of each plants fire risk assessment for regulatory applications. The ANS standard is intended to provide the necessary information for determining the technical adequacy of the licensees fire risk analyses for regulatory applications.

NFPA 805 [NFPA-01] requires that the PSA approach, methods, and data must be acceptable to the AHJ). In the case of the United States, the AHJ is the U.S. NRCs Office of Nuclear Reactor Regulation (NRR). This regulatory position provides guidance with respect to acceptability of the approaches, methods and data used for the PSA approach. Additional guidance for the PSA approach is provided by NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), [NEI 05a],

including Sections 5.1.3, 5.3.4, J.4, and J.5.

Licensees should justify that the methods that the NRC finds acceptable for use in meeting NFPA 805 [NFPA-01] requirements are appropriate for each specific application. These analyses may use screening methods or more complex quantitative PSA methods, depending on the specific conditions of the scenario being evaluated. When licensees choose to rely on information in an internal events-based PSA model to quantify risk associated with fires, they should review the analysis to ensure that the model addresses applicable NFPA 805 requirements, including the engineering analysis requirements in Section 2.4.2, Nuclear Safety Capability Assessment, of NFPA 805. Based on the review, the licensee should modify its internal events-based PSA model, as necessary, to meet applicable NFPA 805 requirements.

119

5 CONCLUSIONS AND RECOMMENDATIONS Based on the contributions by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Fire Incidents Records Exchange (FIRE)

Database Project, member countries significant efforts and resources have been allocated to the regulatory programs associated with fire protection regulations. Many countries ensure that the requirements of the country-specific regulation are met and the fire protection codes and standards are followed by implementing a fire protection program for all operational stages of the facilities. The fire protection program demonstrates that fire protection measures are implemented in a controlled, coordinated, and effective manner to protect the health and safety of people and the environment from fire events.

In recent years, a significant amount of work has been initiated into regulatory activities based on a risk-informed approach and the development of new regulatory structures to take advantage of this advancement.

5.1 General Conclusions This Topical Report presents at a high level the processes and methods each country undertakes to meet specific regulatory goals related to fire safety. It highlights that the 14 FIRE member countries employ various defense-in-depth approaches for ensuring nuclear safety and that the plant can be safely shut down in the event of a fire.

This Topical Report provides a valuable tool to the members of the OECD/NEA FIRE Database Project to aid in the development of future Topical Reports. It can, for instance, be used in supporting the currently ongoing investigations with respect to fire brigade effectiveness and implementation strategies including country-specific approaches to fire brigade and fire suppression techniques. Details include to investigate if and in how far according to the national fire safety regulations in place professional onsite versus offsite fire brigades are used and evaluate challenges and impact on non-suppression probabilities for the two approaches. A unified source for information on regulatory requirements allows for effective searching for the regulatory source materials from each member country and compiling relevant information to inform future Topical Reports.

One general conclusion from the survey questionnaire is that a wide range of countries are using both some form of prescriptive as well as risk-informed methodological approaches to carry out their regulatory framework related to fire safety. The general trend among member countries is moving towards using Fire PSA as a tool for gaining valuable risk insights regardless of regulatory framework guidelines. This trend supports the conclusions from several investigations of the level of maturity for Fire PSA that more recently (cf. [NEA 19]) has amongst others been an area of interest of the OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) Working Group on Risk Assessment (WGRISK).

5.2 Recommendations to the FIRE Database Project High-quality PSA reviews in member countries largely rely on online plant inspections, sensitivity and uncertainty studies, and various international standards to ensure the assumptions and quality meet corresponding member country standards. These inspections were generally conforming to various international standards such as the International Atomic Energy Agency (IAEA) guidelines and further guidance documents such as the American Society of Mechanical Engineers /American Nuclear Society PRA Standards, NUREG publications, and various methodologies for peer reviews. The complexity of Fire PSA methodologies is clearly highlighted through the level of effort described in the quality review 121

responses. Review of plant-specific Fire PSA methodology assumptions and model choices has been shown to require substantial resources.

In the future, the FIRE Database may be extended as a mechanism for efficient feedback on fire event experience including the development of additional measures for prevention and oversight activities. In addition, the Database may be used to focus on events that document failures of fire protection features to be used as a data source reliability. It may also be used for reviewing observations and findings from inspections of nuclear power plant fire protection programs.

Updates of the Topical Report are to be foreseen in case of changes in the national standards and regulations of FIRE Database Project member countries.

5.3 Recommendations to CSNI and CNRA This Topical Report provides an overview on the range of fire-related regulations in a broad sample of OECD/NEA members countries representative for different types of reactors and fire safety programs in place. Many of the results and insights of this activities reflect the actual progress in the expert community dealing with fire hazards considering from various fire incidents in OECD/NEA member countries. This also underlines the value of operating experience information to decisionmakers.

Based on the aforementioned insights, it is recommended to the CSNI and the Committee for Nuclear Regulatory Activities to (1) Consider supporting further joint Working Group or Database Project activities concerning lessons from major operational incidents aiming on enhancing the regulations by operational feedback.

(2) More generally, continue to support efforts to increase interactions between the FIRE Database Project, in particular with the Working Group on Operating Experience and WGRISK, but also with other relevant OECD/NEA Working Groups.

(3) Encourage and facilitate cooperation with the IAEA on related activities to address the challenges of fire-related guidance and decision-making.

122

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[JPN-04] NRA Ordinance on Standards for the Location, Structures and Equipment of Commercial Power Reactors, http://nsr-portal/kyoyu0101/SARIS_Attachment/L03The%20NRA%20Ordinance%2 0on%20Standards%20for%20the_.pdf.

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[JPN-05] NRA Ordinance on Technical Standards for Commercial Power Reactors Facilities, http://nsr-portal/kyoyu0101/SARIS_Attachment/L02The%20NRA%20Ordinance%2 0on%20Technical%20Standards.pdf.

[JPN-06] Regulatory Guide of the NRA Ordinance on Standards for the Location, Structure, and Equipment of Commercial Power Reactors, http://nsr-portal/kyoyu0101/SARIS_Attachment/L04The%20Regulatory%20Guide%

20of%20the%20NRA%20Ordinance_.pdf.

[JPN-07] Regulatory Guide of the NRA Ordinance on Technical Standards for Commercial Power Reactors Facilities, http://nsr-portal/kyoyu0101/SARIS_Attachment/L05(For%20SARIS%20module%20 Module9-5%20QID63)Article%2035%20of%20the_.pdf.

[JPN-08] Fire Protection Review Standard for Power Reactor Facilities, in Japanese only, http://warp.da.ndl.go.jp/info:ndljp/pid/8729504/www.nsr.go.jp/nra/kettei/da ta/20130628_jitsuyounaiki03.pdf.

[JPN-09] Guide for Evaluating the Effects of Internal Fires at Nuclear Power Stations, in Japanese only, http://warp.da.ndl.go.jp/info:ndljp/pid/8729504/www.nsr.go.jp/nra/kettei/da ta/20130628_jitsuyounaikasai.pdf.

[KOR-01] Nuclear Laws of the Republic of Korea: Enforcement Regulation of the Nuclear Safety Act, Article 20: Details of Periodic Safety Review, Korea Institute of Nuclear Safety (KINS), Enacted by Regulation of the Prime Minister No. 2, Nov. 11, 2011 (Entered into force, Nov. 11, 2011),

Amended by Presidential Decree No. 24689, Aug. 16, 2013, http://www.nssc.go.kr/nssc/en/nci/elif/Enforcement_Regulation_of_the_N uclear_Safety_Act.pdf.

[KOR-02] Nuclear Laws of the Republic of Korea: Regulation on Technical Standards for Nuclear Reactor Facilities, Etc., Article 14: Protection against Fire Protection, Etc., Korea Institute of Nuclear Safety (KINS),

Enacted by Ordinance of the Ministry of Education, Science and Technology No. 16, Apr. 18, 2000, Amended by Ordinance of the Ministry of Education, Science and Technology No. 31, Jul. 28, 2001, Ordinance of the Ministry of Education, Science and Technology No. 92, Jul. 19, 2006, Ordinance of the Ministry of Education, Science and Technology No. 1, Mar. 4, 2008Regulation of the Nuclear Safety and Security Commission No. 3, Nov. 11, 2011, http://www.nssc.go.kr/nssc/en/nci/elif/Regulations_on_Technical_Standar ds_for_Nuclear_Reactor_Facilities,ETC.pdf.

[KOR-03] Nuclear Laws of the Republic of Korea: Regulation on Technical Standards for Nuclear Reactor Facilities, Etc., Article 59: Fire Protection Program, Korea Institute of Nuclear Safety (KINS), 2008, Regulation of the Nuclear Safety and Security Commission No. 3, Nov. 11, 2011, http://www.nssc.go.kr/nssc/en/nci/elif/Regulations_on_Technical_Standar ds_for_Nuclear_Reactor_Facilities,ETC.pdf.

[KOR-04] Nuclear Safety and Security Commission (NSSC): Notice 2015-11:

Technical Standards for Fire Hazard Analysis, in Korean only, 130

http://www.law.go.kr/admRulLsInfoP.do?chrClsCd=010202&admRulSeq=

2100000033523.

[KOR-05] Nuclear Safety and Security Commission (NSSC) Notice 2015-12:

Regulation on Establishment and Implementation of Fire Protection Program, 2015, in Korean only, http://www.law.go.kr/admRulLsInfoP.do?chrClsCd=010202&admRulSeq=

2100000033524.

[KOR-06] Korea Institute of Nuclear Safety (KINS): Fire Protection for Nuclear Power Plants, Regulatory Guide 10.6, in Korean only, http://www.kins.re.kr/nsic.do?menu_item=technologyStatus.

[KOR-07] Korea Institute of Nuclear Safety (KINS): KINS/GE-07: Safety Review Guides for PSR.

[KTA-00] KTA 2101, Part 1: Fire Protection in Nuclear Power Plants, Basic Requirements, Version 12/2000 (superseded by [KTA-15]),

http://www.kta-gs.de/e/standards/2100/2101_1_engl_2000_12.pdf.

[KTA-00a] KTA 2101, Part 2: Fire Protection in Nuclear Power Plants, Fire Protection of Structural Plant Components, Version 12/2000 (superseded by [KTA-15a]),

http://www.kta-gs.de/e/standards/2100/2101_2_engl_2000_12.pdf.

[KTA-00b] KTA 2101, Part 3: Fire Protection in Nuclear Power Plants, Fire Protection of Mechanical and Electrical Plant Components, Version 12/2000 (superseded by [KTA-15c]),

http://www.kta-gs.de/e/standards/2100/2101_3_engl_2000_12.pdf.

[KTA-09] Nuclear Safety Standards Commission (KTA, German for:

Kerntechnischer Ausschuss): Design of Nuclear Power Plants against Damaging Effects from Lightning; KTA 2206, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2009-11, http://www.kta-gs.de/e/standards/2200/2206_engl_2009_11.pdf

[KTA-11] Nuclear Safety Standards Commission (KTA, German for:

Kerntechnischer Ausschuss): Design of Nuclear Power Plants against Seismic Events; Part 1: Principles, KTA 2201.1, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2011-11, http://www.kta-gs.de/e/standards/2200/2201_1_engl_2011_11.pdf.

[KTA-15] Nuclear Safety Standards Commission (KTA, German for Kerntechnischer Ausschuss): Fire Protection in Nuclear Power Plants, Part 1: Basic Requirements, KTA 2101.1, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2015-11, December 2015, http://www.kta-gs.de/e/standards/2100/2101_1_engl_2015_11.pdf.

[KTA-15a] Nuclear Safety Standards Commission (KTA, German for Kerntechnischer Ausschuss): Fire Protection in Nuclear Power Plants, Part 2: Fire Protection of Civil Structures, KTA 2101.2, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2015-11, December 2015, http://www.kta-gs.de/e/standards/2100/2101_2_engl_2015_11.pdf.

[KTA-15b] Nuclear Safety Standards Commission (KTA, German for Kerntechnischer Ausschuss): Fire Protection in Nuclear Power Plants, Part 3: Fire Protection of Mechanical and Electrical Plant Components, KTA 2101.3, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2015-11, December 2015, http://www.kta-gs.de/e/standards/2100/2101_3_engl_2015_11.pdf.

131

[KTA-15c] Nuclear Safety Standards Commission (KTA, German for:

Kerntechnischer Ausschuss): Explosion Protection in Nuclear Power Plants with Light Water Reactors (General and Case-specific Requirements), KTA 2103, Safety Standards of the Nuclear Safety Standards Commission (KTA), 2015-11, December 2015, http://www.kta-gs.de/e/standards/2100/2103_engl_2015_11.pdf.

[NEA-13] Organisation for Economic Co-operation and Development (OECD)

Nuclear Energy Agency (NEA), Committee on the Safety of Nuclear Installations (CSNI): OECD FIRE Project - Topical Report No. 1, Analysis of High Energy Arcing Fault (HEAF) Fire Events, NEA/CSNI/R(2013)6, Paris, France, June 2013, http://www.oecd-nea.org/documents/2013/sin/csni-r2013-6.pdf.

[NEA-16] Organisation for Economic Co-operation and Development (OECD)

Nuclear Energy Agency (NEA), Committee on the Safety of Nuclear Installations (CSNI): OECD FIRE Project - Topical Report No. 3, Combinations of Fires and Other Events, NEA/CSNI/R(2016)7, Paris, France, January 2016, http://www.oecd-nea.org/documents/2016/sin/csni-r2016-7.pdf.

[NEA-17] Organisation for Economic Co-operation and Development (OECD)

Nuclear Energy Agency (NEA), Committee on the Safety of Nuclear Installations (CSNI): Experimental Results from the International High Energy Arcing Fault (HEAF) Research Program Testing Phase 2014 to 2016, NEA/CSNI/R(2017)7, Paris, France, 2017, http://www.oecd-nea.org/documents/2016/sin/csni-r2017-7.pdf.

[NEA-19] Organisation for Economic Co-operation and Development (OECD)

Nuclear Energy Agency (NEA): CSNI Technical Opinion Paper No. 17, Fire Probabilistic Safety Assessments for Nuclear Power Plants: 2019 Update, OECD 2019, , NEA No. 7417, Paris, France, 2019, http://www.oecd-nea.org/nsd/pubs/2019/7417-csni-top17.pdf.

[NED-01] Nuclear Energy Act (Kernenergiewet or Kew), 1963, https://wetten.overheid.nl/BWBR0002402/2018-10-16.

[NED-02] NVR NS-G-2.1: Brandveiligheid in de bedrijfsvoering van kernenergiecentrales.

[NED-03] NVR NS-G-1.7: Bescherming tegen interne branden en explosies in het ontwerp van kernenergiecentrales.

[NED-04] NVR NS-G-1.11: Bescherming tegen interne gevaren anders dan branden en explosies in het ontwerp van kernenergiecentrales.

[NED-05] NVR-NS-R1 (Safety Requirements for Nuclear Power Plant Design),

https://www.iaea.org/publications/6002/safety-of-nuclear-power-plants-design.

[NED-06] NVR-SSG-2 (Deterministic Safety Analysis),

https://www-pub.iaea.org/MTCD/publications/PDF/Pub1428_web.pdf.

[NED-07] The Safety Regions Act (Wet Veiligheidsregio's, WVR), October 1, 2010, http://wetten.overheid.nl/BWBR0027466/geldigheidsdatum_11-10-2014.

[NED-08] The Decree Safety Regions (Besluit veiligheidsregios, June 24, 2010, http://wetten.overheid.nl/BWBR0027844/geldigheidsdatum_11-10-2014.

[NED-09] The Building Decree (Bouwbesluit, 2012), April 1, 2012, http://wetten.overheid.nl/BWBR0030461/geldigheidsdatum_11-10-2014.

[NED-10] Dutch Safety Requirements for Nuclear Reactors: Fundamental Safety Requirements, 19.03.2015, https://www.oecd-nea.org/nsd/docs/2015/csni-r2015-15.pdf.

132

[NED-11] Vienna Declaration on Nuclear Safety (VNDS),

https://english.autoriteitnvs.nl/topics/guidelines-on-the-safe-design-and-operation-of-nuclear-reactors.

[NED-12] IPSART mission report to Borssele Nuclear Power Plant. IAEA-REM-2011, Vienna, Austria, 2011.

[NED-13] Probabilistic Safety Assessment for the Borssele Nuclear Power Plant.

IPSART-12, 2013 intermediate model, 2013.

[NEI-05] Nuclear Energy Institute(NEI): Guidance for Post-Fire Safe Shutdown Circuit Analysis, NEI 00-01, Washington, DC, USA, January 2005, https://www.nrc.gov/docs/ML0503/ML050310295.pdf.

[NEI-05a] Nuclear Energy Institute(NEI): Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), NEI-04-02, Rev. 1, Washington, DC, USA, September 2005, https://www.nrc.gov/docs/ML0525/ML052590476.pdf.

[NFPA-98] National Fire Protection Association (NFPA): NFPA 803, Standard for Fire Protection for Light Water Nuclear Power Plants, 1998 Edition, Quincy, MA, USA, 1998.

[NFPA-01] National Fire Protection Association (NFPA): NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, Quincy, MA, USA, 2001.

[NFPA-15] National Fire Protection Association (NFPA): NFPA 804, Standard for Fire Protection for Advanced Light Water Reactor Electric Generating Plants, 2015 Edition, Quincy, MA, USA, 2015.

[NFPA-18] National Fire Protection Association (NFPA): NFPA 5000, Building Construction and Safety Code, 2018 Edition, Quincy, MA, USA, 2018.

[NFPA-18a] National Fire Protection Association (NFPA): NFPA 101, Life Safety Code, 2018 Edition, Quincy, MA, USA, 2018.

[NFPA-18b] National Fire Protection Association (NFPA): NFPA 1, Fire Code, 2018 Edition, Quincy, MA, USA, 2018.

[NRA-13] Nuclear Regulation Authority (NRA): Standard for the Examination of Practical Power Generation Nuclear Reactors and Associated Facilities Regarding their Fire Protection, NUCREGTEC1306195, June 2013 (only Japanese).

[NRA-13a] Nuclear Regulation Authority (NRA): Guide for Evaluating the Effects of External Fires at Nuclear Power Stations, NUCREGTEC13061912, June 2013 (only Japanese).

[NRC-01] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 Fire Protection, August 2007, Appendix A to Part 50 - General Design Criteria for Nuclear Power Plants, (updated) January 1, 2011, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appa.html.

[NRC-02] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 Fire Protection, Appendix R to 10 CFR Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979", updated 2017, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appr.html.

[NRC-04] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 133

Fire Protection, August 2007, http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0048.html.

[NRC-05] United States Nuclear Regulatory Commission (NRC): Appendix A to Branch Technical Position (BTP) APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants, 1976, https://www.nrc.gov/docs/ML0706/ML070660458.pdf.

[NRC-07] United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research: Fire Protection for Nuclear Power Plants, Regulatory Guide 1.189, Rev. 2, Washington, DC, USA, October 2009, https://www.nrc.gov/docs/ML0925/ML092580550.pdf.

[NRC-08] United States Nuclear Regulatory Commission (NRC): EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Final Report, NUREG/CR-6850, EPRI 1011989, Washington, DC, USA, 2005, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6850/.

[NRC-09] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 Fire Protection, Alternate Fire Protection Rule [10 CFR 50.48(c), NFPA 805] January 16, 2004, https://www.nrc.gov/reactors/operating/ops-experience/fire-protection/protection-rule.html.

[NRC-10] United States Nuclear Regulatory Commission (NRC): Cable Response to Live Fire (CAROLFIRE), NUREG/CR-6931, Washington, DC, USA, April 2008, https://www.nrc.gov/docs/ML0811/ML081190230.pdf.

[NRC-11] United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research: Fire Protection for Nuclear Power Plants, Regulatory Guide 1.189, Rev. 2, Washington, DC, USA, October 2009, https://www.nrc.gov/docs/ML0925/ML092580550.pdf.

[NRC-12] United States Nuclear Regulatory Commission (NRC): Regulatory Guide 1.200: An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Rev.

2, March 2009, https://www.nrc.gov/docs/ML0904/ML090410014.pdf.

[NRC-13] United States Nuclear Regulatory Commission (NRC): Fire Probabilistic Risk Assessment Methods Enhancements, NUREG/CR-6850, EPRI 1019259, Supplement 1, Washington, DC, September 2010, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6850/s1/cr6850s1.pdf.

[NRC-15] United States Nuclear Regulatory Commission (NRC): Regulatory Guide 1.174: An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Rev. 2, May 2011, https://www.nrc.gov/docs/ML1009/ML100910006.pdf.

[NRC-16] United States Nuclear Regulatory Commission (NRC): Joint Assessment of Cable Damage and Quantification of Effects from Fire; Volumes 1-3, (JACQUE-FIRE), NUREG/CR-7150, EPRI 1026424, Washington, DC, USA, 2012/2014/2017, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7150/.

[NRC-17] United States Nuclear Regulatory Commission (NRC): Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire),

134

NUREG/CR-7100 Washington, DC, USA, April 2012, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7100/final/index.html.

[NRC-18] United States Nuclear Regulatory Commission (NRC): EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, NUREG-1921, EPRI 1023001, Washington, DC, USA, May 2012, https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1921/.

[NRC-19] United States Nuclear Regulatory Commission (NRC): A Framework for Low Power/Shutdown Fire PRA, Final Report, NUREG/CR-7114, Washington, DC, USA, May 2013, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7114/.

[NRC-20] United States Nuclear Regulatory Commission (NRC): Fire Protection and Fire Research Knowledge Management Digest, 2013, NUREG/KM-003, Washington, DC, USA, January 2014, https://www.nrc.gov/reading-rm/doc-collections/nuregs/knowledge/km0003/.

[NRC-21] United States Nuclear Regulatory Commission (NRC): B.5.b - Mitigation Strategies Requirements from Order EA-02-026, Section B.5.b, the Subsequent License Conditions, and 10 CFR 50.54 (hh), August 2015, http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0054.html.

[NRC-22] United States Nuclear Regulatory Commission (NRC): Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009, NUREG-2169, EPRI 3002002936, Washington, DC, USA, January 2015, https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr2169/.

[NRC-23] United States Nuclear Regulatory Commission (NRC): Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTI-FIRE), NUREG/CR-7010, Washington, DC, USA, last update 2016, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7010/.

[NRC-24] United States Nuclear Regulatory Commission (NRC): Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Main Report, Volumes 1-7, and Supplement, NUREG-1824, EPRI 3002002182, Washington, DC, USA, last updated November 2016, https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1824/.

[NRC-25] United States Nuclear Regulatory Commission (NRC Heat Release Rates of Electrical Enclosure Fires (HELEN-FIRE), Final Report, NUREG/CR-7197, Washington, DC, USA, February 2016, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7197/.

[NRC-26] United States Nuclear Regulatory Commission (NRC): Refining and Characterizing Heat Release Rates From Electrical Enclosures During Fire Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume (RACHELLE-FIRE), Final Report, NUREG-2178, EPRI 3002005578, Washington, DC, USA, April 2016, https://www.nrc.gov/docs/ML1611/ML16110A140.pdf.

[NRC-27] United States Nuclear Regulatory Commission (NRC): Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE), NUREG-2180, Washington, DC, USA, December 2016, https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr2180/.

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[NRC-28] United States Nuclear Regulatory Commission (NRC): 10 CFR 50 Domestic Licensing of Production and Utilization Facilities; 10 CFR 50.48 Fire Protection, Appendix R to 10 CFR Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979", updated 2017, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appr.html.

[NRC-29] United States Nuclear Regulatory Commission (NRC): Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), Final Report, NUREG-1934; EPRI 1023259, Washington, DC, USA, November 2012, https://www.nrc.gov/docs/ML1231/ML12314A165.pdf

[NRC-30] United States Nuclear Regulatory Commission (NRC): Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric Motors, Indoor Dry Transformers, and the Main Control Board, Final Report, NUREG-2178; EPRI 3002016052, Washington, DC, USA, April 2016, https://www.nrc.gov/docs/ML1908/ML19087A210.pdf

[NRC-31] United States Nuclear Regulatory Commission (NRC): Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinet Fires in Nuclear Power Plants, Final Report, NUREG-2230; EPRI 300201605, Washington, DC, USA, April 2019, https://www.nrc.gov/docs/ML1916/ML19163A293.pdf

[NRC-32] United States Nuclear Regulatory Commission (NRC): Heat Release Rate and Fire Characteristics of Fuels Representative of Typical Transient Fire Events in Nuclear Power Plants, Final Report, NUREG-2232; EPRI 3002015997, August 2019, https://www.nrc.gov/docs/ML2009/ML20091L481.pdf

[NRC-33] United States Nuclear Regulatory Commission (NRC): Methodology for Modeling Transient Fires in Nuclear Power Plant Fire Probabilistic Risk Assessment, Final Report, NUREG-2233; EPRI 3002016054, December 2019, https://www.nrc.gov/docs/ML1934/ML19345F683.pdf

[NRC-34] United States Nuclear Regulatory Commission (NRC): Fire Protection for NPPs, NUREG-0800 BTP CMEB 9.5-1 Standard Review Plan, Rev. 3, Washington, DC, USA, July 1981, https://www.nrc.gov/docs/ML0706/ML070660454.pdf.

[NRC-35] United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research: Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire, NUREG-1852, Washington, DC, USA, October 2007, https://www.nrc.gov/docs/ML0730/ML073020676.pdf.

[NRCC-15] National Research Council Canada: National Building Code of Canada (NBCC) 2015, Toronto, ONT, Canada, 2015, https://nrc.canada.ca/en/certifications-evaluations-standards/codes-canada/codes-canada-publications.

[NRCC-15a] National Research Council Canada: National Fire Code of Canada (NFCC) 2015, Toronto, ONT, Canada, 2015, https://nrc.canada.ca/en/certifications-evaluations-standards/codes-canada/codes-canada-publications.

136

[ONR-14] Office for Nuclear Regulation (ONR): Safety Assessment Principles (SAPs) for Nuclear Facilities, 2014 Edition, Revision 0, November 2014, http://www.onr.org.uk/saps/index.htm.

[ONR-16] Office for Nuclear Regulation (ONR): Probabilistic Safety Analysis, Nuclear Safety Technical Assessment Guide NS-TAST-GD-030, Revision 5, June 2016, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-030.pdf.

[ONR-16a] Office for Nuclear Regulation (ONR): Internal Hazards, Nuclear Safety Technical Assessment Guide NS-TAST-GD-014, Revision 4, September 2016, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-014.htm.

[ONR-18] Office for Nuclear Regulation (ONR): Guidance on the Demonstration of ALARP (As Low As Reasonably Practicable), Nuclear Safety Technical Assessment Guide NS-TAST-GD-005, Revision 9, March 2018, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-005.pdf.

[ONR-18a] Office for Nuclear Regulation (ONR): External Hazards, Nuclear Safety Technical Assessment Guide NS-TAST-GD-013, Revision 7, October 2018, http://www.onr.org.uk/operational/tech_asst_guides/ns-tast-gd-013.htm

[SCI-06] Scientech Inc.: WinNUPRA' for Quantitative Risk Assessment, Extended Summary Ver. 31, USA, February 2006.

[SIS-01] Swedish Standards Institute (SIS): Fixed firefighting systems -

Components for automatic sprinkler and water spray systems, Series of Standards SS-EN 12259, 2001.

[SIS-04] Swedish Standards Institute (SIS): Fixed firefighting systems - Automatic sprinkler systems - Design, installation and maintenance, Standard SS-EN 12845:2004, October 15, 2004.

[SIS-06] Swedish Standards Institute (SIS): Fire detection and fire alarm systems, Series of Standards SS-EN 54, 2006, https://www.notifier.se/filer/EN%2054%20STANDARDER.pdf.

[SIS-09] Swedish Standards Institute (SIS): Fixed firefighting systems - Water spray systems - Design, installation and maintenance, Technical Specification SIS-CEN TS 14816, February 26, 2009.

[SONS-10] State Office for Nuclear Safety (SONS): Protection against Internal Fires -

Safety Instructions JB-3.1, Prague, Czech Republic, 2010, https://www.sujb.cz/fileadmin/sujb/docs/dokumenty/publikace/Ochrana_pr oti_vnitrnim_pozarum_BN_JB_3.1.pdf.

[SPN-01] The Nuclear Energy Act, Law 25/1964, of April 29th, 1964, https://www.boe.es/buscar/pdf/1964/BOE-A-1964-7544-consolidado.pdf, https://www.csn.es/documents/10182/1369702/Law+251964%2C+of+29t h+of+April%2C+on+Nuclear+Energy (unofficial English version).

[SPN-02] The Law15/1980, of April 22nd, 1980, creating the Nuclear Safety Council, https://www.boe.es/buscar/pdf/1964/BOE-A-1964-7544-consolidado.pdf, https://www.csn.es/documents/10182/1369702/Law+Creating+the+Nucle ar+Safety+Council (unofficial translation into English).

[SPN-03] The Royal Decree 1836/1999, of December 3rd, 1999, approving Regulation on Nuclear and Radioactive Facilities, https://www.boe.es/buscar/pdf/1999/BOE-A-1999-24924-consolidado.pdf, https://www.csn.es/documents/10182/1369702/Royal+Decree+1836-137

1999%2C+of+December+3rd%2C+approving+the+Regulation+on+Nucle ar+and+Radioactive+Facilities (unofficial translation into English).

[SPN-04] Royal Decree 1400/2018, of November 23rd, 2018, Regulation on nuclear safety in nuclear facilities, https://www.boe.es/buscar/pdf/2018/BOE-A-2018-16041-consolidado.pdf.

[SPN-05] Instrucción de Seguridad IS-30 del CSN, sobre requisitos del programa de protección contra incendios en centrales nucleares, https://boe.es/boe/dias/2016/11/30/pdfs/BOE-A-2016-11342.pdf, https://www.csn.es/documents/10182/1348817/Instruction%20IS-30,%20Revision%202,%20of%20November%2016th%202016,%20on%2 0the%20requirements%20of%20the%20fire%20protection%20programm e%20at%20nuclear%20power%20plants (unofficial English version).

[SPN-06] The Guia de Seguridad GS 1.19, Programa de Protección contra Incendios en Centrales Nucleares, https://www.csn.es/documents/10182/896572/GS+01-19+Requisitos+del+programa+de+protecci%C3%B3n+contra+incendios+

en+centrales+nucleares.

[SSM-09] Swedish Radiation Safety Authority (SSM): The Swedish Radiation Safety Authoritys Regulations and General Advice concerning Safety in Nuclear Facilities, SSMFS 2008:1, ISSN 2000-0987, January 30, 2009, (latest version in Swedish only),

https://www.stralsakerhetsmyndigheten.se/contentassets/e8ba282f8ef046 1ca392137c9495f466/ssmfs-20081-the-swedish-radiation-safety-authoritys-regulations-concerning-safety-in-nuclear-facilities.

[SSM-09a] Swedish Radiation Safety Authority (SSM): The Swedish Radiation Safety Authoritys Regulations concerning the Design and Construction of Nuclear Power Plants, SSMFS 2008:17, ISSN 2000-0987, January 30, 2009, https://www.stralsakerhetsmyndigheten.se/contentassets/ace5827a9cfc4 3fc9576fca14443da32/ssmfs-200817-the-swedish-radiation-safety-authoritys-regulations-concerning-the-design-and-construction-of-nuclear-power-reactors.

[SWD-01] Lag (2003:778) om skydd mot olyckor (Civil Protection Act), 20 November 2003, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/lag-2003778-om-skydd-mot-olyckor_sfs-2003-778.

[SWD-02] Frordning (2003:789) om skydd mot olyckor (Civil Protection Ordinance),

20 November 2003, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/forordning-2003789-om-skydd-mot-olyckor_sfs-2003-789.

[SWD-03] Freskrifter om ndring i Strlskerhetsmyndighetens freskrifter och allmnna rd (SSMFS 2008:1) om skerhet i krntekniska anlggningar, SSMFS 2011:3, 1 November 2011, https://www.stralsakerhetsmyndigheten.se/contentassets/6b6ce39b86b84 5c998cdc0062c07353e/ssmfs-20113-foreskrifter-om-andring-i-stralsakerhetsmyndighetens-foreskrifter-ssmfs-20081-om-sakerhet-i-karntekniska-anlaggningar.pdf.

[SWD-04] Miljbalk (1998:808) (Swedish Environmental Code), 11 June 1998, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/miljobalk-1998808_sfs-1998-808.

138

[SWD-05] Lag (2010:1011) om brandfarliga och explosiva varor (Flammable and Explosive Goods Act), 1 July 2010, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/lag-20101011-om-brandfarliga-och-explosiva_sfs-2010-1011.

[SWD-06] Arbetsmiljlag (1977:1160) (Work Environment Act), 19 December 1977, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/arbetsmiljolag-19771160_sfs-1977-1160.

[SWD-07] Arbetsmiljfrordning (1977:1166) (Work Environment Ordinance), 19 December 1977, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/arbetsmiljoforordning-19771166_sfs-1977-1166.

[SWD-08] Plan- och bygglag (2010:900) (Planning and Building Act), 1 July 2010, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/plan--och-bygglag-2010900_sfs-2010-900.

[SWD-09] Plan- och byggfrordning (2011:338) (Planning and Building Ordinance),

31 March 2011, https://www.riksdagen.se/sv/dokument-lagar/dokument/svensk-forfattningssamling/plan--och-byggforordning-2011338_sfs-2011-338.

[SWD-10] The Swedish National Board of Housing, Building and Plannings general recommendations (BFS 2011:27) on the analytical design of a buildings fire protection BBRAD, 18 June 2013, https://www.boverket.se/globalassets/publikationer/dokument/2013/bbrad

-bfs-2011-27-tom-2013-12-english.pdf.

[SWD-11] The Swedish National Board of Housing, Building and Plannings building regulations - mandatory provisions and general recommendations, BBR, BFS 2011:6 with amendments up to BFS 2011:26, 10 October 2011.

[SWD-12] Swedish Fire Protection Association: Regler fr automatisk brandlarmsanlggning (automatic fire alarm installations), SBF 110:6, 1 January 2001.

[SWD-13] Statens rddningsverk, Statens rddningsverks allmnna rd och kommentarer om systematiskt brandskyddsarbete, SRVFS 2004:3, ISSN 0283-6165, February 6, 2004, https://www.msb.se/contentassets/244bc5f1e438414eb6b58339cc58939 e/statens-raddningsverks-allmanna-rad-och-kommentarer-om-systematiskt-brandskyddsarbete.pdf.

[SWT-01] Swiss Confederation: Federal Constitution, 101, 18 April 1999, https://www.admin.ch/opc/en/classified-compilation/19995395.

[SWT-02] Federal Assembly of the Swiss Confederation, Nuclear Energy Act, 732.1, 21 March 2003, https://www.admin.ch/opc/en/classified-compilation/20010233.

[SWT-03] Federal Council, Nuclear Energy Ordinance, 732.11, 10 December 2004, https://www.admin.ch/opc/en/classified-compilation/20042217.

[SWT-04] Federal Department of the Environment, Transport, Energy and Communications (DETEC), Verordnung des UVEK über die Gefhrdungsannahmen und die Bewertung des Schutzes gegen Strflle in Kernanlagen, 732.112.2, 17 June 2009, https://www.admin.ch/opc/de/classified-compilation/20090231.

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139

[SWT-06] Swiss Federal Council, Radiological Protection Ordinance, 814.501, 22 June 1994, https://www.admin.ch/opc/en/classified-compilation/19940157.

[UK-01] Health and Safety at Work Act 1974 (HSWA74),

https://www.legislation.gov.uk/ukpga/1974/37.

[UK-02] The Energy Act 2013 (TEA13),

http://www.legislation.gov.uk/ukpga/2013/32/contents/enacted.

[UK-03] Nuclear Installations Act 1965 (NIA65),

https://www.legislation.gov.uk/ukpga/1965/57.

[UK-04] Ionising Radiations Regulations 2017 (IRR17),

http://www.legislation.gov.uk/uksi/2017/1075/contents/made.

[UK-05] Radiation (Emergency Preparedness and Public Information) Regulations 2001 (REPPIR),

https://www.legislation.gov.uk/uksi/2001/2975/contents/made.

[UK-06] Regulatory Reform (Fire Safety) Order 2005, http://www.legislation.gov.uk/uksi/2005/1541/contents/made.

[UK-07] Fire Safety (Scotland) Regulations 2006, http://www.legislation.gov.uk/ssi/2006/456/contents/made.

[UK-08] Fire Safety Regulations (Northern Ireland) 2010, https://www.legislation.gov.uk/nisr/2010/325/contents/made.

[UK-09] BS 9999:2017: Code of practice for fire safety in the design, management and use of buildings, 2017.

[UK-10] Nuclear Site Licence Conditions, http://www.onr.org.uk/licensing.htm.

[VKF-15] Vereinigung Kantonaler Feuerversicherungen (VKF),

Brandschutzvorschriften 2015, https://www.bsvonline.ch/de/vorschriften/?req=Norm_&anchor.

[WENRA-08] Western European Nuclear Regulators Reactor Association (WENRA)

Reactor Harmonization Working Group (RHWG): WENRA Reactor Safety Reference Levels - Issue S: Protection against Internal Fires, January 2008, http://www.wenra.org/media/filer_public/2012/11/05/list_of_reference_lev els_january_2008.pdf.

[WENRA-14] Western European Nuclear Regulators' Reactor Association (WENRA):

Waste and Spent Fuel Safety Reference Levels Report, Report of Working Group on Waste and Decommissioning (WGWD) Version 2.2, April 2014, http://www.wenra.org/media/filer_public/2014/05/08/wgwd_storage_report

_final.pdf.

[WENRA-14a] Western European Nuclear Regulators' Reactor Association (WENRA)

Reactor Harmonization Working Group (RHWG): WENRA Reactor Safety Reference Levels - Issue S: Protection against Internal Fires, September 2014.

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APPENDIX A GERMAN NUCLEAR FIRE AND EXPLOSION STANDARDS KTA 2101.1-3 AND KTA 2103 A-1

A-2 A-3 A-4 A-5 A-6 A-7 A-8 A-9 A-10 A-11 A-12 A-13 A-14 A-15 A-16 A-17 A-18 A-19 A-20 A-21 A-22 A-23 A-24 A-25 A-26 A-27 A-28 A-29 A-30 A-31 A-32 A-33 A-34 A-35 A-36 A-37 A-38 A-39 A-40 A-41 A-42 A-43 A-44 A-45 A-46 A-47 A-48 A-49 A-50 A-51 A-52 A-53 A-54 A-55 A-56 A-57 A-58 A-59 A-60 A-61 A-62 A-63 A-64 A-65 A-66 A-67 A-68 A-69 A-70 A-71 A-72 NUREG/IA-0519, Rev. 1 Survey of Member Countries Nuclear Power Plant Fire Protection Regulations by the OECD Nuclear Energy Agency (NEA) Fire Incidents September 2020 Records Exchange (FIRE) Database Project - Topical Report No. 2 Nicholas B. Melly, Thomas H. Aird, Mark Henry Salley Technical Organisation for Economic Co-operation and Development (OECD)

Nuclear Energy Agency (NEA)

Committee on the Safety of Nuclear Installations (CSNI)

Paris, France Division of Risk Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Nations establish and authorize agencies with the responsibility to protect the health and safety of the public and the environment by licensing and regulating nuclear power plants (NPP). Fires have been shown to be a major risk to NPP safety. Fire protection regulations built on defense-in-depth principles have been established in each country to minimize this risk. The purpose of this report is to collect and share the fire protection regulations and strategies used in different countries to ensure reactor safety. The report is built by each member country assembling their major fire protection regulations with all the regulations translated in the common language of English. In addition, an international trend exists for regulations to evolve from prescriptive requirements to risk-informed performance-based requirements. This report includes that information where applicable. The completed report now provides a single reference where countries can review and contrast their NPP fire protection regulations with other member countries. Through international cooperation in projects such as this research effort, each member country may discover new insights and ideas for their NPP fire protection regulations.

Fire Risk Fire safety Regulations Fire protection Nuclear Energy Agency Committee on the Safety of Nuclear Installations Organisation for Economic Co-operation and Development OECD

NUREG/IA-0519, Rev. 1 Survey of Member Countries Nuclear Power Plant Fire Protection September 2020 Regulations by the OECD Nuclear Energy Agency (NEA) Fire Incidents Records Exchange (FIRE) Database Project - Topical Report No. 2