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Westinghouse AP1000 GDA Step 3 Structural Integrity Assessment Report

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Westinghouse AP1000 GDA Step 3 Structural Integrity Assessment Report Health and Safety Executive NUCLEAR DIRECTORATE GENERIC DESIGN ASSESSMENT – NEW CIVIL REACTOR BUILD STEP 3 STRUCTURAL INTEGRITY ASSESSMENT OF THE WESTINGHOUSE AP1000 DIVISION 6 ASSESSMENT REPORT NO. AR 09/013-P HSE Nuclear Directorate Redgrave Court Merton Road Bootle Merseyside L20 7HS HSE Nuclear Directorate Division 6 Assessment Report No. AR 09/013-P EXECUTIVE SUMMARY This report records my assessment of the nuclear safety-related structural integrity aspects of the Westinghouse AP1000 design, for Step 3 of the Generic Design Assessment (GDA). The Nuclear Directorate (ND) usage of the term ‘structural integrity’ covers metal pressure boundary components, their supports and some of the associated internal support structures (e.g. for a PWR, the core barrel). In this GDA Step 3 assessment of the structural integrity aspects of the AP1000 design proposed for the UK, I have not identified any matters that would lead to a recommendation to raise a Regulatory Issue. During GDA Step 3 I have raised a number of matters with Westinghouse; I have done this mostly through thirteen Regulatory Observations (RO). Some matters raised are relatively more significant than others. I consider useful progress has been made across all of these Regulatory Observations. Several aspects of these Regulatory Observations remain to be resolved. I consider there is a reasonable prospect of achieving such resolution by carrying these remaining open aspects forward into GDA Step 4. For structural integrity aspects of the AP1000 and from an ND perspective I believe there has been a significant improvement in HSE understanding of the design. For components where ‘the likelihood of gross failure is claimed to be so low it can be discounted’, Westinghouse has indicated a willingness to implement a method of achieving and demonstrating integrity consistent with UK practice. I regard this as substantial progress within GDA Step 3 and a basis for going forward. Execution of the programme of work will extend into GDA Step 4. Full implementation might extend beyond GDA Step 4. If so, some interim work might be done within GDA Step 4 to give confidence for final implementation. I believe ND will want to take an interest in the detail of this work and how the programme of works progresses. The question of which components have the claim that the likelihood of gross failure is so low it can be discounted still remains to be completely resolved. Westinghouse has proposed a programme of work to address this matter. I regard this as substantial progress within GDA Step 3 and a basis for going forward. Execution of the programme of work will extend well into GDA Step 4. I believe ND will want to take an interest in the detail of this work and how the programme of works progresses. Aspects of the chemical composition of the low alloy ferritic steels for the main vessels (Reactor Pressure Vessel, Steam Generators and Pressuriser) remain to be resolved. This topic will also carry into GDA Step 4, but it is an item that needs to be resolved sooner rather than later. Largely based on authoritative advice received under a support contract, there may be detailed aspects to discuss with Westinghouse relating to several matters of material specification. However, I do not see these aspects as fundamental impediments to progress and resolution. For the purposes of this assessment, I have assumed the Reactor Pressure Vessel (RPV) will have set-in (also referred to as set-through) nozzles, rather than an integral nozzle shell course design. If an integral nozzle shell course was proposed for the RPV, this would require specific assessment. For neutron irradiation embrittlement of regions of the RPV, the AP1000 design takes account of what is now known regarding chemical composition of the base materials and welds. However, the end of life maximum neutron dose to the forgings is quite high. I recommend ND seeks an As Low As Reasonably Practicable (ALARP) review of practical options for meaningful reduction of neutron dose to the RPV. As a minimum this should consider the locations of peak neutron dose, which occur in the forging. On the face of it, there is the potential for adopting a ‘low leakage core’ fuel management arrangement in service. The basis of Reactor Coolant Pump Casing construction based on casting technology has been justified. However, there are still aspects to resolve in how to deal with large repairs to the castings Page (i) HSE Nuclear Directorate Division 6 Assessment Report No. AR 09/013-P made by welding. The areas still open relate to how to obtain confidence that crack-like defects of a size of concern for integrity can be detected. Useful progress has been made in understanding the approach to be used for an AP1000 in setting Pressure-Temperature limit curves for the RPV. However, there are aspects still to be resolved, for instance consideration of what is ALARP. The design of the steel containment shell complies with the American Society of Mechanical Engineers (ASME) code. There are some general aspects that need further consideration: plate thickness available for corrosion allowance for most of the shell, the toughness properties of the plates and welds to meet the requirements for no post weld heat treatment and tolerance on plate thickness relevant to both corrosion allowance and no post weld heat treatment. There are a number of matters to take forward for further assessment. The AP1000 Design Control Document (DCD) states that the Steam Generator tubing will be made using mill annealed Alloy 690 in the Thermally Treated (TT) condition. Based on my knowledge of UK experience of Thermally Treated Alloy 690 Steam Generator tubing and a general perception of international experience of this material, I had no particular concerns about its use. But, given the past interest in the UK of this aspect of Pressurised Water Reactor (PWR) structural integrity, I judged it prudent to give the matter some consideration. I decided to do this through a support contract to review PWR Steam Generator tube materials and manufacturing routes. Overall, I conclude from the review and my general knowledge of this area that Alloy 690 in the Thermally Treated condition is a sound choice of material for Steam Generator Tubing. When supported by detailed manufacturing practice and in-service water chemistry control, Alloy 690TT tubing exhibits good resistance to stress corrosion cracking. However, material choice, manufacturing practice and in-service water chemistry are not a panacea. The general design and construction aspects of the Steam Generator as they affect the tubing also have a role. Important factors are the minimisation of ‘crevice’ conditions, support for the tubing to avoid vibration induced wear and support materials that themselves do not corrode. Most of these general design and construction factors have been understood for many years, and the AP1000 Steam Generator design takes these into account. A number of matters are identified above for carrying forward in to GDA Step 4 and some will require significant effort and programmes of work on the part of Westinghouse (e.g. the work for Regulatory Observation RO-AP1000-19). In addition GDA Step 4 for structural integrity needs to move to the next level of detail and consider the content of documents such as: Design Specifications. Analyses for loading conditions (mainly thermal-hydraulics analyses - this will require involvement of other ND assessment functions). Design Reports. Equipment Specifications. for a range of components. From an ND perspective, I consider there has been a reduction in regulatory risk. Page (ii) HSE Nuclear Directorate Division 6 Assessment Report No. AR 09/013-P LIST OF ABBREVIATIONS ALARP As Low as Reasonably Practicable ASME American Society of Mechanical Engineers BMS (Nuclear Directorate) Business Management System BSL Basic Safety Level (in SAPs) BSO Basic Safety Objective (in SAPs) BWR Boiling Water Reactor CFR Code of Federal Regulations (USA) CVS Chemical and Volume System DCD Design Control Document dpa displacements per atom EA The Environment Agency EPRI Electric Power Research institute FSER Final Safety Evaluation Report (by USNRC) GDA Generic Design Assessment GDC General Design Criteria (or in singular Criterion). In Appendix A of 10 CFR 50 HSE The Health and Safety Executive IAEA The International Atomic Energy Agency IASCC Irradiation Assisted Stress Corrosion Cracking IGA Intergranular Attack IGSCC Intergranular Stress Corrosion Cracking ND Nuclear Directorate (of HSE) PCSR Pre-construction Safety Report PID Project initiation Document (HSE / ND) P-T Pressure-Temperature PWSCC Primary Water Stress Corrosion Cracking QA Quality Assurance RCC-M Design and Construction Rules for Mechanical Components of PWR Nuclear Islands (AFCEN, France) RCS Reactor Coolant System RI Regulatory Issue RIA Regulatory Issue Action RO Regulatory Observation ROA Regulatory Observation Action RP Requesting Party RPV Reactor Pressure Vessel Page (iii) HSE Nuclear Directorate Division 6 Assessment Report No. AR 09/013-P LIST OF ABBREVIATIONS SAP Safety Assessment Principle (plural SAPs) SSC System, Structure and Component TAG (Nuclear Directorate) Technical Assessment Guide TQ Technical Query TT Thermally Treated (referring to Steam Generator tubing) US NRC (or NRC) United States Nuclear Regulatory Commission WEC Westinghouse Electric Company LLC Page (iv) HSE Nuclear Directorate Division 6 Assessment Report No. AR 09/013-P TABLE OF CONTENTS 1 INTRODUCTION.....................................................................................................................
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