Index Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021

Page numbers followed by f and t indicate figures and tables, respectively.

Abandonment, Canadian oil and gas pipeline systems, 1-15–1-16 Applications Above-ground NUHOMS® system, advantages of, 20-6–20-11 FANC, 5-3–5-4 robust canister ESP for aging management, 20-9–20-10 of HVD, 9-15–9-26 shielding performance, 20-7–20-8 in seismic and vibration isolation systems for nuclear structures unloading operations, safe and simple, 20-8–20-9 and components, 9-20, 9-21, 9-22, 9-24, 9-27–9-29 Accidents to NB, 2-10 due to high-pressure gas, HPGSL, 13-4 pressure equipment, Belgium, 5-3–5-4 management guidelines, JSME codes, 13-29–13-30 RI-ISI, for piping (Spanish regulation), 7-9–7-12 NPPs in Finland at Spanish NPPs, 7-10–7-12 EOP, guidance for operators, 8-12 risk-informed, NPPs in Finland Act of Atomic Energy, in Hungary, 11-1 for operating license, 8-4, 8-12–8-13 Advanced fuel cycle initiatives, 1-10–1-11 during operation, 8-4–8-6 Advanced Fuel Cycle Reactor (AFCR), 1-11 TSO Bel-V, 5-3–5-4 Advanced Heavy Water Reactor (AHWR), 16-27–16-28 AREVA TN AEC. See Atomic Energy Council (AEC) high-capacity high-heat load NUHOMS® EOS system, 20-11–20-12 AECL (Atomic Energy of Canada Limited), 14-1, 14-9 licensing regulatory risk management, 20-2–20-3 AERB (Atomic Energy Regulatory Board), 16-7, 16-23, 16-27, 16-28 MP197HB transportation package, 20-12–20-14 AFCEN (French association of design, construction, and inservice description, 20-13–20-14 inspection rules for nuclear island components), 3-1, 3-3–3-8, robust canister ESP for aging management, 20-9–20-10 3-54 ASME BPV Code, China, 17-17–17-18 working group, 3-7–3-8 ASME China International Working Group (CIWG), 17-4–17-7 Aging Figure 17.10 (photograph of whole family of B&PVC III CIWG, management, robust canister ESP for, 20-9–20-10 April 2015), 17-7 AGMS (Annulus Gas monitoring System), 16-11 future of, 17-6–17-7 AHWR (Advanced Heavy Water Reactor), 16-27–16-28 ASME Code in China Air Cooled Diesel Generator (DG), 16-2, 16-7 history of applying, 17-3–17-5 Aligned flaws, combination rules for, 19-8–19-9 ASTS (automatic seismic trip system), 15-5–15-6 Alignment rules, for non-aligned flaws, 19-6–19-8 Atomic Energy Act, 11-3, 11-6, 12-1, 12-4 Allowable fracture toughness, WWER, 10-2–10-3 Atomic Energy Board (AEB), 12-1 Allowable stresses Atomic Energy Corporation (AEC), 12-4 Japanese codes Atomic Energy Council (AEC), 15-1, 15-2 maximum, 13-5–13-6, 13-8 Figure 15.2 (real-time nuclear power operational status on AEC pressure equipment website), 15-2 Japanese codes, 13-2, 13-3, 13-4, 13-5–13-6 Figure 15.3 (organization structure of the AEC in Taiwan), 15-3 Alternative grouping, quasi-laminar flaws, 19-10–19-11 Figure 15.4 (organizational structure of the department of nuclear American Society of Mechanical Engineers (ASME) regulation of the AEC in Taiwan), 15-4 Board on Nuclear Codes and Standards (BNCS) Atomic Energy of Canada Limited (AECL), 1-35, 14-1, 14-9 non-mandatory appendices, revising risk-informed, 3-46 Atomic Energy Regulatory Board (AERB), 16-7, 16-23, 16-27, 16-28 specifications, for steel, 2-21–2-22 Austenitic stainless steels ANI (Authorized Nuclear Inspector), 17-8 yield stress safety factor for, 9-4 Annulus Gas monitoring System (AGMS), 16-11 Authorized Inspectors (AI), 5-8 Annulus spacers, 1-6–1-7 Authorized Nuclear In-service Inspector, 5-8 I-2 • Index

Authorized Nuclear Inspector (ANI), 5-8, 17-8 Boron, reactor coolant, 1-8 Autoclave sterilizers, 5-2 Bottom-mounted instrumentation (BMI), 15-8 Autoclaving, coolant tube, 16-12 Brayton cycle, recuperative, 12-6 Automated welding system, NUHOMS® system, 20-5–20-6 British Standards (BS), specific types Automatic seismic trip system (ASTS), 15-5–15-6 BS 3915 (steel vessels for primary circuits of nuclear reactors), Automobiles, high-pressure cylinders for on-board natural gas fuel 4-8 storage (Canadian standards), 1-12, 1-14 BS EN 286 (simple unfired pressure vessels designed to contain Automotive propane vessels, Canadian standards, 1-12 air/nitrogen), 4-3, 4-7 Availability of materials, nuclear reactors in India and, 16-28 Brittle fracture AVN (Association Vinçotte Nuclear) toughness conformance of pressure equipment, 2-20 transfer of regulatory activities, 5-4 Buckling interstiffener, 4-5–4-6 BARC (Bhabha Atomic Research Centre), 16-20 strain, theoretical, 4-6–4-7

BDB (beyond design basis) events BWRs. See Boiling water reactors (BWRs) Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 dry storage system design for recovery from, 20-12 Belgium pressure equipment regulation, 5-1–5-13 Calandria, 1-5–1-6, 1-8 application, 5-3–5-4 assembly, CANDU® nuclear power plants, 1-21–1-22 derogation, 5-3–5-4 Calandria-End shield assembly, 16-7–16-8 rules and experience feedback, evolution, 5-12–5-13 Calandria tubes, 16-2, 16-5, 16-11–16-12 transposition of ASME Code, Section XI, 5-3, 5-4, 5-5–5-12 coolant channel assembly, 16-11–16-12 Appendix IX (Application Rules of ASME Code Section III reactor components, 16-13 or Other Regulations for Repair or Replacement of Canada Deuterium Uranium (CANDU®) reactor design and licensing Components in Operating Nuclear Units), 5-9–5-11 basis, 1-1 Appendix X (Independent Body Distinct from Mandated advanced fuel cycle initiatives, 1-10–1-11 Organization Performing AIA Role), 5-11–5-12 CANDU® Owners Group (COG), 1-35 derogations, 5-5–5-6 CANDU® 6 reactor, 1-5–1-8 Subsection IWA (General Requirements), 5-6–5-8 concrete containment structures for, 1-25 Subsection IWB (Requirements for Class 1 Components of design, 1-5–1-8 Light-Water Cooled Plants), 5-8–5-9 enhanced CANDU 6® (EC6®), 1-8–1-10 Bel-V, TSO fuel channel assembly, 1-6–1-7 application, 5-3–5-4 fuel channel end-fitting assemblies, 1-23–1-24 scope of work, 5-6–5-7 future developments, 1-8–1-11 Beyond design basis (BDB) events moderator, 1-10 dry storage system design for recovery from, 20-12 plant life extension in Canadian nuclear standards, 1-36–1-37 B factor, 3-58 power reactors, 1-18 Bhabha Atomic Research Centre (BARC), 16-20 Canadian Boiler and Pressure Vessel Standards, 1-1–1-37 BMI (bottom-mounted instrumentation), 15-8 class 2 components, 1-22–1-23, 1-25 Boiler and pressure vessel codes (BPVC) Figure 1.5 (Simplified Schematic of CANDU Fuel Channel ASME Assembly in Reactor Core), 1-7 motivation of, adaptation in Hungary, 11-1–11-3 overview, 1-1–1-2 Paks NPP, Hungary, 11-1–11-9 plant life extension in Canadian nuclear standards, 1-36–1-37 PED and, 2-20–2-22, 3-2 standards governing, 1-2–1-11 pressure equipment directive perspectives, 2-1–2-31 Canadian Boiler and Pressure Vessel Standards, specific types Russian nuclear standard PNAE vs., 9-5–9-13 B series (Tolerance Specifications and Pressure Boundary Section XI, transposition (Belgian environment), 5-3, 5-4, Standards), 1-5 5-5–5-12 CAN/CSA-B51 (Boilers, Pressure Vessels, and Pressure Piping), Boilers 1-1, 1-4, 1-11–1-14, 1-19 Canadian standards, 1-2–1-11, 1-12 comparison to ASME Code, 1-14 inservice inspection, 1-28–1-36 overview, 1-11–1-12 in Germany, 6-1–6-10 registration, 1-12–1-13 Japanese codes, 13-2–13-4 subcommittees, 1-11–1-12 maintenance, 3-41–3-45 CAN/CSA-B149.2 (Propane Storage and Handling Code), 1-4, 1-13 Boiling water reactors (BWRs), 15-1 CAN/CSA-N285.0 (General Requirements, Classification, CANDU® design, 1-5 Registration, and Reporting), 1-1, 1-18–1-23, 1-25, 1-26 Cofrentes BWR plant, 7-5, 7-11 calandria assembly, 1-21–1-22 JSME S NH1-2006 Code, 13-29 fabrication-related issues, 1-20–1-21 pressure boundary integrity, 15-8 nuclear classification, 1-18–1-19 reactor designs, 1-5 registration, 1-19–1-20 Santa Maria de Garona, 7-5, 7-9 CAN/CSA-N285.4 (Periodic Inspection of CANDU® Nuclear specification for defect manufacturing in, 7-7 Power Plant Components), 1-1, 1-21, 1-28, 1-29–1-31, 1-35 of US design, 7-1, 7-5 CAN/CSA-N285.5 (Periodic Inspection of CANDU® Nuclear Bolted flanged connections Power Plant Containment Components), 1-1, 1-21, 1-23, PD 5500 (U.K.), 4-8–4-9 1-28, 1-31–1-33, 1-34 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-3

CAN/CSA-N285.7 (Periodic Inspection of Pressure Retaining China, nuclear boiler and pressure vessel codes and standards, BOP Systems, Components and Supports), 1-1–1-2, 1-29, 17-1–17-9 1-33–1-34 ASME Code in China, 17-3–17-5 CAN/CSA-N285.8 (Flaw Evaluation of CANDU® Zirconium Figure 17.6 (hoisting reactor vessel of AP1000 in Sanmen NPP, Alloy Pressure Tubes), 1-1, 1-29, 1-35–1-36 September 2011), 17-5 CAN/CSA-N287.7 (Inservice Examination and Testing history of applying, 17-3–17-5 Requirements for Concrete Containment Structures for CIWG, 17-4–17-7 CANDU® Nuclear Power Plants), 1-1, 1-25, 1-29, 1-31, 1-32, Figure 17.10 (photograph of whole family of B&PVC III 1-34–1-35 CIWG, April 2015), 17-7 CAN/CSA-N393 (Fire Protection for Facilities that Process, future of, 17-6–17-7 Handle, or Store Nuclear Materials), 1-5 history of applying ASME code, 17-3–17-4 CAN/CSA-N290 (A) (Reactor Control Systems, Safety Systems improvements of Chinese codes and standards, 17-2–17-3 and Instrumentation), 1-5 international counterparts, 17-1–17-2

CAN/CSA-N290 (B) (Reactor Safety and Risk Management), 1-5 needs in nuclear codes and standards, 17-7–17-9 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 CAN/CSA-N285.6-Series (Material Standards for Reactor certification, 17-7–17-8 Components for CANDU® Nuclear Power Plants), 1-1, 1-21, domestic production of equipments, 17-7–17-8 1-22, 1-23–1-24 “Going Global” strategy, 17-8–17-9 CAN/CSA-N286 Series (Quality Assurance Program independent supervision, 17-7–17-8 Requirements), 1-1, 1-5, 1-18, 1-23, 1-24–1-25 Chloride-induced stress corrosion cracking (CISCC), 20-9–20-10 CAN/CSA-N289 Series (Seismic Qualification of CANDU® CISCC (chloride-induced stress corrosion cracking), 20-9–20-10 Nuclear Power Plant Structures and Systems), 1-1, 1-5, 1-18, CIWG (ASME China International Working Group), 17-4–17-7 1-26–1-28 Class 1 components CAN/CSA-Z305.3 (Pressure Regulators, Gauges, and Flow- RCC-M, 3-50–3-54 Metering Devices for Medical Gases), 1-13 Class 2 components CAN/CSA-Z662 (Oil and Gas Pipeline Systems), 1-4, 1-13, Canadian Boiler and Pressure Vessel Standards, 1-22–1-23, 1-25 1-14–1-18 Class 1 piping, RI-ISI application Canadian Standards Association (CSA) Spanish PWRs, 7-10–7-11 annexes, nonmandatory and mandatory, 1-4 Code Convergence Board format and structure of standards, 1-3–1-4 SDOs’ establishment of, 18-4–18-5 non-nuclear boiler, pressure vessel, and piping design and Code de construction des Appariels à Pression (CODAP), 2-12, 3-1, construction standards, 1-11–1-18 3-2, 3-3, 3-6, 3-8–3-23, 3-41, 3-61–3-62, 4-7, 4-8, 5-2 nuclear boiler and pressure vessel design and construction design, 3-13–3-16 standards, 1-18–1-28 material, 3-9–3-13 nuclear boiler and pressure vessel in-service inspection standards, permissible welded joints, 3-18–3-19 1-28–1-36 scope, 3-8–3-9 nuclear standards, 1-3–1-4 tolerances on branches, 3-17 plant life extension in Canadian nuclear standards, 1-36–1-37 Code de construction des générateurs de Vapeur (COVAP), 3-1, 3-2, standards developing process, 1-2–1-3 3-3, 3-6, 3-35–3-41, 3-61–3-62 Canistered high burnup licensed transportation cask, MP197HB, Code de construction de Tuyauteries Industrielles (CODETI), 2-12, 20-12–20-14 3-1–3-2, 3-3, 3-6, 3-23–3-34, 3-61–3-62 description, 20-13–20-14 design, 3-28–3-30 27 cases of Notices of the Nuclear Safety and Security Commission dimensional tolerances for prefabricated spools, 3-31, 3-34 for the reactor regulation, 14-1 flexibility characteristic, and flexibility and stress intensification Cask Code, 13-25–13-27 factors, 3-33 Cask Loader tool, 21-6. See also Entergy Corporation links with PED, 3-24–3-25 CBA (core barrel assembly), in PBMR, 12-7–12-9 NDE, extent and nature of, 3-35–3-36 Certification safety factors, 3-29 nuclear codes in China, 17-7–17-8 scope, 3-23–3-24 Challenges, IPHWR, 16-17–16-18 Combination rules, for aligned flaws, 19-8–19-9 Chemical plants, fired-heater pressure coils in, 1-12, 1-13–1-14 Commercial nuclear power plants in Korea (as of January 2015). Chernobyl NPP, unit 3, 9-18–9-19, 9-22, 10-1 See Korean regulatory system and codes of nuclear boiler Chi-Chi earthquake, 15-5–15-6 and pressure vessels China, non-nuclear boiler and pressure vessel codes and standards, Commissariat à l’Energie Atomique (CEA), 3-4, 3-5, 3-7 17-9–17-18 Committee of Enquiry into Pressure Vessel Industry (U.K.), 4-1 application of ASME BPV code, 17-17–17-18 Component designer, CANDU® nuclear power plants, 1-18, 1-19–1-20 development history, 17-9 Components, IPHWR design, 16-7–16-13 equipment, 17-9–17-10 Calandria-End shield assembly, 16-7–16-8 Figure 17.13 (hydrocracking reactor for petrochemical industry), coolant channel assembly, 16-11–16-13 17-10 Figure 16A.5 (hanging Calandria-end shield assembly of RAPS-1,2 Figure 17.14 (zirconium composite plate reactor), 17-11 and MAPS to the integral Calandria-end shield assembly regulatory mechanism and characteristics, 17-10–17-13 from NAPS onward), 16-8 technical safety regulations and standards, 17-13–17-17 main PHT system and components, 16-8–16-10 relevant reference standards, 17-14–17-15 philosophy, 16-10–16-11 I-4 • Index

Computer programs, quality assurance, for CANDU® nuclear power Czech & Slovakian codes, 10-1–10-15 plants, 1-24–1-25 COVERS continuation, 10-6–10-7 Conceptual designs, nuclear reactors in India, 16-26–16-28 Figure 10.1 (Comparison of Allowable Stres Intensities Advanced Heavy Water Reactor (AHWR), 16-27–16-28 (Allowable Fracture Toughnes Values)), 10-4 Fast Breeder Test Reactor (FBTR), 16-26–16-27 Figure 10.2 (Temperature Dependence of Static Fracture Prototype Fast breeder Reactor (PFBR), 16-27 Toughness Data), 10-5 Concrete Containment Vessels Code, 13-21–13-23 Figure 10.3 (Temperature Dependence of Static Fracture CVE-3000 Design, 13-22–13-23 Toughness Data Adjusted to 1 IN. Thickness of 15Kh2MFA CVE-1000 General, 13-21–13-22 Type Steel), 10-6 Conformity assessment modules history, 10-1 with/without QA, 2-8–2-10 IAEA guidelines, 10-7–10-11 Conformity assessment procedures, 2-1, 2-2, 2-4, 2-6, 2-8–2-11, 3-9 Appendix A (Leak Before Break (LBB) Concept Application manufacturer’s responsibility, 2-17–2-18 to Selected Piping Systems of WWER Type NPPs), 10-7

Conical shells, 4-4–4-5 Appendix C (Integrity and Lifetime Assessment Procedure Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 Conservative deterministic failure margin (CDFM), 9-12–9-13 of RPV Internals in WWER NPP´s during Operation), Consolidated Interim Storage (CIS), 20-1 10-7–10-11 Construction Appendix XVIII (Recommendations on Calculation of nuclear boiler and pressure vessels, Canadian standards, 1-11–1-18 Cyclic Strength, taking into Account Effect of Water pressure equipment, Japanese codes, 13-13–13-14, 13-23–13-25 Environment), 10-7 Containment design, IPHWR, 16-18–16-20 NTD ASI code for WWER reactor components, 10-1–10-3 design against radial stress in IC dome, 16-19–16-20 allowable fracture toughness, 10-2–10-3 monitoring, 16-20 facts and reasons, 10-2 Containment safety systems, EC6® reactor, 1-9–1-10 procedures and criteria, 10-2 Contents Section I (Welding and brazing of components and piping of KEPIC, 14-25–14-26 WWER type NPPs), 10-2, 10-11–10-12 of Nuclear Safety Act, 14-4–14-5 Section II (Characteristics of materials for components and Coolant ­piping of WWER type NPPs), 10-2, 10-12–10-13 for CANDU® reactor, 1-5 Section III (Strength assessment of components and piping of reactor, boron, 1-8 WWER type NPPs), 10-2, 10-13 Coolant channel assembly, IPHWR, 16-11–16-13 Section IV (Evaluation of residual lifetime of compo- Calandria tubes, 16-11–16-12 nents and piping of WWER type NPPs), 10-2, 10-3, coolant tubes, 16-12–16-13 10-13–10-14 Coolant tubes Section V (Material testing procedures and evaluation), 10-2, leak-before-break methodology, 16-12–16-13 10-3, 10-14–10-15 Cooperation in Reactor Design Evaluation and Licensing Section VI (Air condition systems for WWER type NPPs), (CORDEL), 18-3, 18-4 10-2 Copper Table 10.1 (Comparison of Safety Factors of Different Codes for pressure equipment Based on Yield and Ultimate Tensile Strength), 10-10 Japanese codes, 13-5–13-6 Table 10.2 (Allowable Stress Intensity Limits for WWER Reactor Copper alloys Pressure Vessels (RPVs) and Bolting Joints), 10-10 for pressure equipment VERLIFE procedure, 10-3–10-6, 10-7 Japanese codes, 13-5–13-6 CORDEL (Cooperation in Reactor Design Evaluation and Licensing), DAE (Department of Atomic Energy), 16-20 18-3, 18-4 Daily inspection by resident inspectors, 14-7 Core barrel assembly (CBA), in PBMR, 12-7–12-9 Deactivation, Canadian oil and gas pipeline systems, 1-15–1-16 COVERS, WWER Safety Research, 10-6–10-7 DEIR (Designated Equipment Inspection Regulations), 13-3, 13-4–13-5, Czech Association of Mechanical Engineers (ASI) 13-8 NTD, code for WWER reactor components, 10-1–10-3 Delayed Hydride Cracking (DHC), 16-11, 16-12–16-13 allowable fracture toughness, 10-2–10-3 Department of Atomic Energy (DAE), 16-20 facts and reasons, 10-2 Department of Nuclear Regulation, Taiwan, 15-2–15-3, 15-4 procedures and criteria, 10-2 Derogation, Belgian pressure equipment regulation, 5-4, 5-5–5-6, 5-8 Section I (Welding and brazing of components and piping of Design(s) WWER type NPPs), 10-2, 10-11–10-12 CANDU® nuclear power plants, 1-18–1-28 Section II (Characteristics of materials for components and EN 13445, 4-16–4-22 piping of WWER type NPPs), 10-2, 10-12–10-13 French codes Section III (Strength assessment of components and piping of FBRs, 3-57–3-60 WWER type NPPs), 10-2, 10-13 pressure equipment Section IV (Evaluation of residual lifetime of components and Japanese codes, 13-13–13-14, 13-23–13-25 piping of WWER type NPPs), 10-2, 10-3, 10-13–10-14 PD 5500 (U.K.), 4-2, 4-4–4-16 Section V (Material testing procedures and evaluation), 10-2, RCC-M, 3-49–3-54 10-3, 10-14–10-15 Design and Construction Rules for Mechanical Components of Section VI (Air condition systems for WWER type NPPs), 10-2 Nuclear Installations (RCC-MRx), 3-1, 3-4, 3-5, 3-57–3-59 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-5

Design and Construction Rules for Mechanical Components of PWR Figure 20.13 (UNF Unloading Operation With NUHOMS® Nuclear Islands (RCC-M), 3-1, 3-3, 3-4, 3-45–3-60, 3-62, 5-7, System), 20-8 5-11, 6-2, 6-4, 6-9 Figure 20.14 (NUHOMS® 24P DSC), 20-9 class 1 pressure components, design, 3-50–3-54 Figure 20.15 (Advanced NUHOMS® 24PT1 DSC), 20-9 design and construction rules, 3-4–3-8, 3-49–3-54 Figure 20.16 (Slice of Cross Section of Temperature Distribution fabrication, 3-54–3-56 in DSC), 20-10 FBRs, 3-57–3-60 Figure 20.17 (Cosmetic Damage to HSM After North Anna general requirements, 3-45–3-46 Earthquake), 20-10 links with PED, 3-46–3-47 Figure 20.19 (NUHOMS® EOS 37PTH DSC Basket), 20-11 material and procurement, 3-48–3-49 Figure 20.20 (NUHOMS® EOS 89BTH Basket), 20-11 reactor vessel steel, 3-50, 3-51 Figure 20.21 (NUHOMS® HSM Inside Storage Building), 20-12 structure of, 3-6–3-7 Figure 20.22 (Integrating Transportation and Storage of UNF), testing/inspection, 3-56–3-57 20-13 ®

Designated equipment (DE), 13-3, 13-4–13-5, 13-8 Figure 20.23 (General Arrangement of NUHOMS MP197HB Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 Designated Equipment Inspection Regulations (DEIR), 13-3, 13-4–13-5, Packaging), 20-13 13-8 MP197HB transportation package, 20-12–20-14 Design by analysis description, 20-13–20-14 nuclear pressure vessels PD 5500 (U.K.), 4-14–4-16 design approaches for high burnup fuel, 20-14 Design phase PRA, for OL3 NPP, 8-11–8-12 payload description, 20-14 plant design changes, examples of, 8-11–8-12 NUHOMS® system Development procedure, KEPIC, 14-22–14-24, 14-26 above-ground, advantages of, 20-6–20-11 DG (Air Cooled Diesel Generator), 16-2, 16-7 auxiliary equipment, 20-5–20-6 DHC (Delayed Hydride Cracking), 16-11, 16-12–16-13 benefits of, 20-3–20-5 Direct use of spent pressurized water reactor fuel in CANDU® designed for safety during BDB events, 20-10 (DUPIC), 1-5 EOS 37 PWR system capabilities, 20-11 Dished ends EOS 89 PWR system capabilities, 20-11–20-12 EN 13445, 4-18, 4-19–4-20 EOS system, advanced high-capacity high-heat load, Dry cask storage, 21-4–21-7. See also Spent fuel storage 20-11–20-12 “Dry Cask Storage Characterization Project; CASTOR V/21 Cask heat transfer performance of, 20-8 Opening and Examination,” 21-5 robust canister ESP for aging management, 20-9–20-10 Dry shielded canister (DSC), NUHOMS® design, 20-3–20-5, 20-6, safe against floods, 20-11 20-7–20-8, 20-9–20-10, 20-11–20-12, 20-13, 20-14 shielding performance, 20-7–20-8 Dry storage, license types, 21-4–21-5 simple ISFSI pad/basemat requirements, 20-7 Dry storage methods, 13-25 unloading operations, safe and simple, 20-8–20-9 Dry storage system, for UNF (case study), 20-1–20-15 AREVA TN EA-12-051, Order To Modify Licenses With Regard To Reliable Spent designed metal casks, 20-11 Fuel Pool (SFP) Instrumentation, 21-3 licensing regulatory risk management, 20-2–20-3 Earthquake, design for, 13-32–13-33 MP197HB transportation package, 20-12–20-14 Earthquake experiences, Taiwan, 15-5–15-6 overview, 20-2 Elastoplastic analysis, French pressure equipment, 3-52 robust canister ESP for aging management, 20-9–20-10 Electricity production, nuclear energy. See Korean regulatory system background, 20-1–20-2 and codes of nuclear boiler and pressure vessels Figure 20.18 (TN-24 Storage Casks Stored at Fukushima in Electric Power Generation Station (ITC-EP-2), 7-2, 7-4–7-5 Japan), 20-11 Electric Utility Industry Law, 13-2–13-3 Figure 20.1 (UNF Management After On-Site Storage), 20-2 End fittings, 1-6–1-7, 1-21, 1-23–1-24 Figure 20.2 (AREVA’s Complete Fuel Cycle Management Expertise), Enforcement, licensing system, 14-8–14-9 20-2 Enforcement and Regulatory Authorities, Belgian, 5-8 Figure 20.3 (UNF in Transit Using AREVA TN Transportation Enhanced CANDU 6® (EC6®), 1-8–1-10 Package), 20-2 containment safety systems and reactor building, 1-9–1-10 Figure 20.4 (AREVA TN’s Worldwide UNF Storage and fuel handling and storage system, 1-8 Transportation Systems), 20-3 reactor design, 1-8 Figure 20.5 (Typical DSC Construction With Tube and Disc Basket), Entergy Corporation 20-4 Dry Fuel Storage Systems, 21-6t Figure 20.6 (Typical DSC Construction With Fuel Compartments spent fuel wet storage systems, 21-2t and Basket Rails), 20-4 Environmental fatigue evaluation code, 13-14–13-16 Figure 20.7 (NUHOMS® Transfer Cask), 20-5 Environmental Survey Laboratories (ESLs), 16-21 Figure 20.8 (Examples of NUHOMS® HSM Arrays), 20-5 EPR (Evolutionary Pressurized Reactor), 18-2 Figure 20.9 (NUHOMS® System Auxiliary Equipment), 20-6 EPRI-TR-112657, rev. B, 7-10 Figure 20.10 (Loaded DSC/TC Transfer to ISFSI Pad), 20-7 Flammanville 3 Project, 3-4, 3-46, 3-53, 3-54 Figure 20.11 (Inserting NUHOMS® DSC Into HSM), 20-7 safety features, OL3 NPP, 8-10 Figure 20.12 (Convective Air Flow Between NUHOMS® HSM Equivalent flat-bottom hole criteria, 3-56 and DSC), 20-8 Eskom, 12-1, 12-4–12-5 I-6 • Index

ESLs (Environmental Survey Laboratories), 16-21 plant design changes, examples of, 8-11–8-12 Essential safety requirements (ESRs) risk informed pre- and in-service inspection program, 8-12–8-13 of Pressure Equipment Directive, 2-2–2-2, 2-4, 2-6, 2-7, 2-11–2-15, PRA 2-18, 2-20 licensing process of new designs, 8-3–8-6 basic principles, 2-11–2-12 operation, risk-informed applications during, 8-4–8-6 design, 2-12–2-13 personnel training, 8-5–8-6 manufacturing, 2-13–2-15 regulatory process, 8-2–8-6 for material, 2-16–2-17 RI-ISI applications in, 8-6–8-10 Euro Norm (EN) Standards, specific types Loviisa NPP RI-ISI projects, experiences of, 8-7–8-10 EN 13445 Olkiluoto NPP RI-ISI projects, experiences of, 8-10 design, 4-16–4-22 selection process, piping required by Reg. guide YVL E.5, EN ISO 9712, 2-13–2-14 8-6–8-7 for steel, 2-21–2-22 risk-informed activities, extension of

European Approval of Materials (EAM), 2-12, 2-15–2-16, 2-21, 3-26, grading of regulatory authority activities, 8-13 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 3-38, 4-2 Finland Olkiluoto 3 project, 3-3, 8-1, 8-2–8-3, 8-10 European Fast Reactor (EFR) studies, 3-4 Fired-heater pressure coils, 1-12, 1-13–1-14 European Nuclear Regulators, 7-6 Fired or otherwise heated equipment, categorization, 2-6–2-7 European Pressurized Water Reactor (EPR), 6-4, 8-1, 18-2 Fitness-for-service (FFS) codes European Union (EU), 2-1–2-2, 2-4, 2-6, 2-11, 2-21, 4-2, 5-2, 6-1, JSME codes for nuclear power planar components, 13-16–13-19 7-1 JSME S NA1-2012, seismic stresses for FFS code, 13-33–13-35 Evolved safety program (ESP), robust canister Fixed boilers, steam generators, 5-2 for aging management, 20-9–20-10 Flanges Export, nuclear regulations for pressure vessels, 6-2–6-10 bolted joints, PD 5500 (U.K.), 4-8–4-9, 4-21 Extended Optimized Storage (EOS) system Flat Bottom Holes (FBHs), 11-4, 11-7 NUHOMS®, advanced high-capacity high-heat load, Flaw (crack) 20-11–20-12 evaluation, Japanese codes, 13-6–13-7, 13-18, 13-19 EOS 37 PWR system capabilities, 20-11 Flaws, modeling/characterization, 19-1–19-11 EOS 89 PWR system capabilities, 20-11–20-12 Figure 19.4 (Re-Characterization from Surface to Through-Wall Flaw), 19-3 Fabrication Figure 19.5 (Transformation from Subsurface to Surface Flaw), CANDU® nuclear power plants, 1-20–1-21 19-4 IPHWR, 16-20–16-21 Figure 19.6 (Proximity factor Y for Subsurface Flaw Obtained), 19-6 Fabricator, CANDU® nuclear power plants, 1-19, 1-20 grouping of multiple laminar flaws, 19-9–19-11 Factor multiplication method, 13-14, 13-15 multiple flaws, 19-6–19-9 Fast breeder reactors (FBRs), 16-20 alignment rules for non-aligned flaws, 19-6–19-8 French codes, 3-4, 3-45, 3-57–3-60 combination rules for aligned flaws, 19-8–19-9 Fast Breeder Test Reactor (FBTR), 16-26–16-27 single flaw, 19-1–19-6 Fatigue proximity rule for subsurface flaw, 19-4–19-6 analysis, French codes, 3-16, 3-52–3-53, 3-58 Table 19.2 (Flaw to Surface Proximity Rules for Subsurface evaluation code, environmental, 13-14–13-16 Flaws), 19-5 of pressure equipment, PD 5500 (U.K.), 4-12–4-14 Forgings reduction factor, pressure vessels (PD 5500 (U.K.)), 4-14 Chinese pressure equipment, 17-9–17-10 strain correction factors, 3-52 Fracture toughness strength/life-property curves (S/N curves), pressure vessels allowable, WWER, 10-2–10-3 (PD 5500 (U.K.)), 4-12–4-14 data strength reduction factor, 3-52 temperature dependence of static, 10-5, 10-6 FBRs (fast breeder reactors), 16-20 French codes, restrictions, 3-47 French codes, 3-4, 3-45, 3-57–3-60 JSME vs. ASME Codes, 13-28 FBTR (Fast Breeder Test Reactor), 16-26–16-27 requirements, difference in, 13-28 FCMA (Fuel Cycle and Materials Administration), 15-1, 15-3–15-4 Framatome, 3-4, 12-1, 12-4, 14-9 Federal Agency for Nuclear Control (FANC) Framatome-ANP, 14-26 application, 5-3–5-4 French codes, dealing with pressure equipment, 3-1–3-62 Bel-V, 5-3–5-4 Figure 3.11 (Dimensional Tolerances for Prefabricated Spools), safety cases, evaluation reports on, 5-12–5-13 3-31 scope of work, 5-6–5-7 Figure 3.16 (RCC-M—Structure of Class 1 Design Rules and Federal Atomic Energy Agency, 9-4 Relation with Nonmandatory Appendices), 3-50 Feeders, PHT, 16-9–16-10 non-nuclear sector, 3-1–3-3 Final safety analysis report (FSAR), 5-5–5-6, 8-1 general, 3-1 Finite element analysis (FEA), 13-13–13-14 organization, 3-3 Finland, NPPs in, 8-1–8-14 PED and European Harmonized Standards, 3-2 OL3 NPP risk-informed licensing, experiences of, 8-10–8-13 publications, updates, interpretations, and inquiries, 3-3 applications for operating license, 8-12–8-13 scope, 3-1–3-3 design phase PRA, 8-11–8-12 structure, 3-2–3-3 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-7

nuclear sector, 3-3–3-8 Heat recovery steam generator (HRSG), 17-9 general, 3-3–3-4 Helium organization, 3-4–3-6 recuperative Brayton cycle with, 12-6 publications, updates, interpretations, and inquiries, 3-7–3-8 Helium pressure boundary (HPB), 12-6, 12-7–12-8 scope, 3-4 High burnup canistered fuel, MP197HB, 20-12–20-14 structure of RCC-M, 3-6–3-7 description, 20-13–20-14 pressure equipment, maintenance, 3-41–3-45 High-capacity high-heat load NUHOMS® EOS system, 20-11–20-12 Table 3.9 (Permitted Type of Acceptance, Design Stress, and EOS 89 PWR system capabilities, 20-11–20-12 Weld Joint Efficiency According to Construction Category of High confidence low probability failure (HCLPF), 9-12–9-13 Vessel), 3-13 High-pressure cylinders, for on-board storage of natural gas fuel for Table 3.13 (Tolerances on Branches (Extract)), 3-17 automobiles, 1-12, 1-14 Table 3.14 (Permissible Welded Joints (Extract-Division 1)), High Pressure Gas Safety Law (HPGSL), 13-3, 13-4–13-5, 13-8 3-18–3-19 High Pressure Institute Standard (HPIS), 13-5–13-7

Table 3.25 (Steel Grades and Maximum Permissible Thicknesses flaw evaluation procedures in pressure equipment, 13-6–13-7 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 in Relation to Construction Categories), 3-29 HPIS Z 101 level 2, 13-6–13-7 Table 3.26 (Safety Factors), 3-29 MAS for B&PV, 13-5–13-6 Table 3.28 (Flexibility Characteristic, and Flexibility and Stress High-pressure polyethylene units, reactors for, 13-5 Intensification Factors), 3-33 High-temperature gas-cooled reactor (HTGR), 12-1, 12-6, 12-7, Table 3.29 (Dimensional Tolerances for Prefabricated Spools), 12-9 3-34 High viscous dampers (HVD), European Table 3.30 (Extent and Nature of Nondestructive Testing), application of, 9-15–9-26 3-35–3-36 in seismic and vibration isolation systems for nuclear structures Table 3.40 (Main Requirements Applying to Material Properties in and components, 9-20, 9-21, 9-22, 9-24, 9-27–9-29 ESPN Order), 3-46 approach in protecting NPP primary and secondary systems, 9-13, Table 3.42A (RCC-M—Reactor Vessel Steel), 3-50 9-14–9-29 Table 3.43A (RCC-M—Reactor Vessel Steel), 3-51 general operational characteristics, 9-14–9-15, 9-16, 9-17 Fuel channels, CANDU® nuclear power plants, 1-21, 1-23–1-24 Historical background, nuclear power in Taiwan, 15-1 Fuel Cycle and Materials Administration (FCMA), 15-1, 15-3–15-4 Holtec HI-STORM dual purpose system, 21-5 Fuel handling and storage system, of EC6® reactor, 1-8 Hook up provision, 700 MWe design, 16-6–16-7 Fukushima Dai-ichi (reactor accident), 5-12, 6-1, 8-3, 13-1, 13-2, Horizontal storage module (HSM), NUHOMS® design, 20-3, 20-5, 13-25, 13-29, 13-31, 16-6, 16-23–16-24, 21-3, 21-4 20-7–20-8, 20-9, 20-10, 20-12 Fusion Experimental Reactor (FER), 13-31 HPGSL (High Pressure Gas Safety Law), 13-3, 13-4–13-5, 13-8 Fusion powerplant component ITER, JSME code, 13-30–13-31 HPIS (High Pressure Institute Standard), 13-5–13-7 Fusion Reactors (FR-ITER), 3-4 HRSG (heat recovery steam generator), 17-9 Hungarian Atomic Energy Authority (HAEA), 11-1, 11-6, 11-8 German Atomic Energy Act (13th Amendment), 6-1, 6-7 Hungary, Paks NPP in, 11-1–11-9 Germany, boiler and pressure vessels in, 6-1–6-10 analysis and evaluation of individual documents, 11-6–11-8 legal adaptations, PED and, 6-1–6-2, 6-3, 6-4, 6-5 background studies, 11-6–11-7 nuclear regulations, applied to NPP, 6-2, 6-6, 6-7 ISI system, development of, 11-6–11-7 for potential export, 6-2, 6-4, 6-5, 6-8, 6-9, 6-10 Section III, 11-6–11-7 Table 6.4 (13th Amendment to German Atomic Energy Act - July Section XI, 11-6–11-7, 11-8 31, 2011), 6-7 structural examinations, 11-7–11-8 Global harmonization, nuclear construction codes and standards, ASME BPVC ISI rules, 11-8–11-9 18-1–18-5 ISI of VVER units, characteristics of, 11-8–11-9 evolution of, 18-1–18-2 RPV UT from OD, 11-8–11-9 perspectives on achieving, 18-4–18-5 international experiences, 11-4–11-6 technical harmonization/convergence, 18-2–18-3 ISI system, main features, 11-3–11-4 “Going Global” strategy, China, 17-8–17-9 ISI/NDE program, 11-3–11-4 Gosatomenergonadzor (GAEN), 10-1 legislative/regulatory framework, US vs., 11-6 Grading of regulatory authority activities, risk informed, 8-13 operational life extension, 11-1–11-2 GROIV, working group, 7-8–7-9 regulatory aspects, 11-2–11-3 Ground storage vessels, compressed natural gas (Canadian regulatory framework, 11-1 standards), 1-12, 1-14 Section XI, 11-3–11-4, 11-5 Grouping, of multiple laminar flaws, 19-9–19-11 Hydrogen control system, EC6® reactor, 1-10 alternative, quasi-laminar flaws, 19-10–19-11 GRUVAL, working group, 7-8–7-9 IAEA (International Atomic Energy Agency), 14-1, 16-7, 16-21 Specific Safety Requirements SSR-2/1, 18-5 Handling, Canadian oil and gas pipeline systems, 1-14–1-15 IAEA guidelines, Appendix C (Integrity and Lifetime Assessment Harmonization, nuclear construction codes and standards, Procedure of RPV Internals in WWER NPP´s during 18-1–18-5 Operation), 10-7–10-11 evolution of, 18-1–18-2 analysis of strength of investigated component, 10-8–10-10 perspectives on achieving, 18-4–18-5 non-mandatory annexes, 10-10, 10-11 technical harmonization/convergence, 18-2–18-3 structure of, 10-7–10-8 I-8 • Index

IGCAR (Indira Gandhi Center for Atomic Research), Kalpakkam, Industrial piping 16-26 design, French codes, 3-28–3-30 Independent Spent Fuel Storage Installation (ISFSI), 21-1, 21-5, French codes (CODETI), 3-23–3-34 21-6–21-7 maintenance, 3-41–3-45 Independent spent fuel storage installation (ISFSI), 20-1, 20-5, 20-6, In-service Inspection (ISI) 20-7, 20-8, 20-9 Rules for Mechanical Components of PWR nuclear islands (RSE-M), Independent supervision, nuclear codes in China, 17-7–17-8 3-1, 3-4, 3-5, 3-46, 3-59–3-60, 3-62, 19-4, 19-8, 19-9 India In-service inspection (ISI) programs, 15-7, 16-2, 16-15–16-17, 16-22 IPHWR, design, 16-1–16-24 In-service inspections (ISI) nuclear power and generation IV nuclear reactors in, French pressure equipment, 3-59–3-60 16-26–16-28 Hungarian ISI system availability of materials, 16-28 ASME BPVC ISI ISI rules, not applicable, 11-8–11-9 conceptual designs, 16-26–16-28 development of, 11-6–11-7

Figure 16B.1 (results for a scenario for India with current main features, 11-3–11-4 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 domestic uranium resources and assuming deployment of RPV UT from OD, 11-8–11-9 FBRs from 2021), 16-26 of VVER units, characteristics of, 11-8–11-9 Figure 16B.2 (heat transport system of Indian prototype FBR), qualification of NDT for, 7-6–7-9 16-27 background, 7-6–7-7 Figure 16B.3 (simplified schematic arrangement of AHWR), description, 7-7–7-8 16-28 implementation of methodology, 7-8–7-9 historical perspective, 16-26 RI-ISI international collaborations, 16-28 applications for piping, 7-9–7-12 technical know how, 16-28 at Spanish NPPs, 7-10–7-12 Indian PHWR (IPHWR), design, 16-1–16-24 Inspection(s) challenges, 16-17–16-18 Canadian standards codes, guides, and standards, development of, 16-22–16-23 nuclear boilers and pressure vessels, 1-29–1-31 components, 16-7–16-13 requirements, 1-14–1-15 Calandria-End shield assembly, 16-7–16-8 French codes coolant channel assembly, 16-11–16-13 pressure equipment, 3-56–3-57 design philosophy, 16-10–16-11 PD 5500 (U.K.), 4-10–4-11 Figure 16A.5 (hanging Calandria-end shield assembly of for pressure equipment, Spain RAPS-1,2 and MAPS to the integral Calandria-end shield ITC-EP-2 (Electric Power Generation Station), 7-2, 7-4–7-5 assembly from NAPS onward), 16-8 non-nuclear industry, 7-1, 7-2–7-5 main PHT system and components, 16-8–16-10 Installation containment design, 16-18–16-20 Canadian oil and gas pipeline systems, 1-14–1-15, 1-16 description, 16-2–16-6 Spanish regulation in non-nuclear industry, 7-1, 7-2–7-5 fabrication, 16-20–16-21 ITC-EP-2 (Electric Power Generation Station), 7-2, 7-4–7-5 in-service inspection, 16-15–16-17, 16-22 Interim Storage Facility (ISF), 13-25, 20-1 introduction to, 16-1–16-2 International Atomic Energy Agency (IAEA), 1-26, 9-2, 14-1, 16-7, 700 MWe PHWR design 16-21 PDHRS, 16-2, 16-6–16-7 guidelines, 10-7–10-11 safety feature in, 16-2, 16-6–16-7 Appendix A (LBB Concept Application to Selected Piping operation, 16-21–16-22 Systems of WWER Type NPPs), 10-7 Passive Decay Heat Removal System (PDHRS), 16-6–16-7 Appendix B (No-Break-Area Assessment for WWER Piping), post Fukushima (Japan) Accident, 16-23–16-24 10-7 reactor components, design philosophy, 16-13 Appendix C (Integrity and Lifetime Assessment Procedure of RPV Calandria tube, 16-13 Internals in WWER NPP´s during Operation), 10-7–10-11 coolant channel components, 16-13 Appendix D (Risk Informed In-Service Inspection), 10-7 regulatory structure, 16-7 Appendix E (NDE System of Qualification), 10-7 Table 16A.2 (partial list of regulatory documents related to Appendix F (Component and Piping Supports), 10-7 nuclear power plants), 16-7 Appendix XVIII (Recommendations on Calculation of seismic qualification, 16-13–16-15 Cyclic Strength, taking into Account Effect of Water Table 16A.1 (design features of IPHWR—700, 540, and 220 Environment), 10-7 MWe), 16-3–16-4 IAEA GS-R-3:2006, 3-8 Indira Gandhi Center for Atomic Research (IGCAR), Kalpakkam, Specific Safety Requirements SSR-2/1, 18-5 16-26 VERLIFE procedure, 10-7 Individual documents, analysis, and evaluation, Hungary, International collaborations, nuclear reactors in India, 16-28 11-6–11-8 International counterparts, Chinese codes and standards, 17-1–17-2 ISI system, development of, 11-6–11-7 International Organization for Standardization (ISO) standards background studies, 11-6–11-7 ­specific types Section III, 11-6–11-7 ISO 11439:2000 (Gas Cylinders-High-Pressure Cylinders for the Section XI, 11-6–11-7, 11-8 On-Board Storage of Natural Gas and Hydrogen as a Fuel for structural examinations, 11-7–11-8 Automotive Vehicles), 1-14 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-9

International Thermonuclear Experimental Reactor (ITER) pipe wall thinning management code, 13-27–13-29 JSME code fusion powerplant component, 13-30–13-31 severe accident management guidelines, 13-29–13-30 vacuum vessel development, 5-58 thermal power plant components, codes for, 13-10–13-12 Interstiffener buckling, 4-5–4-6 organization of code, 13-11–13-12 Investigated component, analysis of strength, 10-8–10-10 Joint efficiency factor, 4-4 ISI (in-service inspection) programs, 15-7, 16-2, 16-15–16-17, Joints 16-22 bolted flanged, PD 5500 (U.K.), 4-8–4-9 ISO 16528 (Boilers and Pressure Vessels—Registration of Codes lattice tube-to-calandria tubesheet, 1-22 and Standards to Promote International Recognition by ISO/TC 11), 17-18 KAERI (Korea Atomic Energy Research Institute), 14-3 ITC-EP-2 (Electric Power Generation Station), 7-2, 7-4–7-5 Kaiga Atomic Power Station, 16-2 Kakrapar Atomic Power Station, 16-2 Jacketed vessels, as pressure equipment, 4-9–4-10 KAMINI (Kalpakkam Mini) reactor, 16-27

Japan, BPVC code and standards, 13-1–13-35 Koeberg power station, 12-1, 12-4–12-5 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 DEIR, 13-3, 13-4–13-5, 13-8 historic information and description, 12-4 Figure 13.4 (Relation Between Laws and JIS (Voluntary risk-informed decision making, approach to, 12-5 Standards) under Mandatory Laws), 13-8 Korea Atomic Energy Research Institute (KAERI), 14-3 Figure 13.8 (Strain Rate and Environmental Fatigue Life Korean Electric Power Industry Codes (KEPIC) Correction Factor Calculated Using Detailed Method), 13-15 background and status of development, 14-14, 14-24–14-25 Figure 13.12 (Organizaton and Contents of Concrete Containment contents of, 14-25–14-26 Code), 13-22 development procedure, 14-22–14-24, 14-26 Figure 13.17 (Comparison of Previous and New Regulatory Table 14.4 (KEPIC to be applied as technical standards of nuclear Requirements), 13-32 reactor facilities), 14-12–14-14 HPGSL, 13-3, 13-4–13-5, 13-8 Table 14.5 (limitations for application of KEPIC 2000 edition HPIS, 13-5–13-7 with addenda and 2005 edition with 2006 addenda I and II), JIS, 13-1, 13-7–13-9 14-15–14-21 JSME codes Table 14.6 (comparison between KEPIC codes and referenced fusion powerplant component ITER, 13-30–13-31 foreign or international codes and standards for the mechani- for nuclear power planar components, 13-13–13-30 cal components and structures), 14-22–14-24 for thermal power plant components, 13-10–13-12 Korean regulatory system and codes of nuclear boiler and pressure for petroleum and petrochemical plants, 13-4–13-9 vessels, 14-1–14-26 seismic rules, 13-31–13-35 guidelines for the application of the Korea electric power industry design for earthquake and tsunami, 13-32–13-33 codes, 14-11–14-14 enforcement of new regulation requirements, 13-31–13-32 Addenda, Article 1 (Enforcement Date), 14-14 JSME S NA1-2012, stresses for FFS code, 13-33–13-35 Addenda, Article 2 (Transitional Measures), 14-14 Table 13.14 (Comparison of Major Items Between JSME and Article 1 (Purpose), 14-11 ASME Codes on Transport/Storage Components), 13-27 Article 2 (Definitions of Terms), 14-11 Table 13.16 (Difference in Fracture Toughness Requirements), Article 3 (Scope of Application), 14-11 13-28 Article 4 (Application Method), 14-11 Table 13.21 (Mandatory Appendices of Fusion Code), 13-31 Article 5 (Interpretation of Standard), 14-11 Japanese Industrial Standards (JIS), 13-1, 13-7–13-9 Article 6 (Application of Verified Technical Contents), 14-11 JIS B 8265 (Construction of pressure vessel: General principles), Article 7 (Application of Unverified Technical Contents), 13-2, 13-3, 13-4, 13-5, 13-7–13-8, 13-9 14-14 MAS for, 13-5–13-6 Article 8 (Report), 14-14 Japan Society of Mechanical Engineers (JSME) Article 9 (Due Date of Reconsideration), 14-14 Concrete Containment Vessels Code, 13-21–13-23 Table 14.4 (KEPIC to be applied as technical standards of CVE-3000 Design, 13-22–13-23 nuclear reactor facilities), 14-12–14-14 CVE-1000 General, 13-21–13-22 KEPIC. See Korean Electric Power Industry Codes (KEPIC) design, construction and maintenance, codes for, 13-2–13-3 legislation system, 14-3–14-6 design and construction code for liquid metal reactors, Figure 14.2 (legislation system), 14-3 13-23–13-25 Table 14.1 (contents of the Nuclear Safety Act), 14-4–14-5 elevated temperature material strength standard, 13-24 Table 14.2 (NSSC notices applicable to reactor facilities), elevated temperature structural design, 13-24 14-5 organization, 13-23–13-24 licensing system and safety assessment, 14-6–14-9 fusion powerplant component ITER, code, 13-30–13-31 CP, 14-6, 14-7 JSME S CA1-2005 Code, 13-27–13-28 daily inspection by resident inspectors, 14-7 JSME S NA1-2012, seismic stresses for FFS code, 13-33–13-35 decommissioning of nuclear installation, 14-7 JSME S NE1-2011, 13-21–13-22 enforcement, 14-8–14-9 for nuclear power planar components, 13-13–13-30 ESA, 14-6, 14-7 Cask Code, 13-25–13-27 Figure 14.3 (reactor licensing and regulation system), 14-6 environmental fatigue evaluation code, 13-14–13-16 Figure 14.4 (reactor inspection process), 14-6–14-9, 14-8 fitness-for-service code, 13-16–13-19 licensing system, 14-6–14-7 LWRs, design and construction code for, 13-13–13-14 OL, 14-6, 14-7 I-10 • Index

periodic inspection, 14-7 Life extension preoperational inspection, 14-7 of Paks NPP, 11-1–11-2 quality assurance audit, 14-7 plant, 1-36–1-37 regulatory inspections, 14-7 Lifting Yoke, NUHOMS® system, 20-5–20-6 SDA, 14-6, 14-7 Ligament efficiency factor, 4-8 special inspection, 14-7 Light water reactors (LWRs), 16-26 locations of nuclear power plants, 14-9–14-10 design and construction code for, 13-13–13-14 Figure 14.5 (locations of commercial nuclear power plants in fuel, 1-10, 1-11 Korea, as of January 2015), 14-9 Liquid Metal Fast Breeder (LMFBR) reactors, 3-4 Table 14.3 (commercial nuclear power plants in Korea, as of Liquid metal reactors, design and construction code for, January 2015), 14-10 13-23–13-25 notices of nuclear safety and security commission, 14-10–14-11 Liquid zone control units, EC6® reactor, 1-8 Notice No. 2012-8, 14-10-14-11 LMFBR (Liquid Metal Fast Breeder) reactors, 3-4

Notice No. 2012-9, 14-10 Local government rules, pressure equipment, 17-14–17-15 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 Notice No. 2012-10, 14-11 Locations of nuclear power plants Notice No. 2012-14, 14-10 Korea, 14-9–14-10 Notice No. 2012-23, 14-11 Loss of Coolant accident (LOCA), 1-18, 1-26–1-27 nuclear regulatory organizations, 14-2–14-3 Loviisa nuclear power plant Figure 14.1 (working mechanism of nuclear safety regulation), RI-ISI projects, experiences of, 8-7–8-10 14-2 new program for piping, old program vs., 8-9–8-10 KINS, 14-3 Low pressure generators, 5-2 NSSC, 14-2 LWRs (light water reactors), 16-26 overview, 14-1 design and construction code for, 13-13–13-14 fuel, 1-10, 1-11 Labeling, of pressure equipment, 2-14–2-15 Laminar flaws MACSTOR® facility, 1-10 multiple, grouping of, 19-9–19-11 Madras Atomic Power Station, 16-1. See also Indian PHWR LBB (leak-before-break) (IPHWR), design application to selected piping systems of WWER type NPPs, 10-7 Maintenance, Canadian oil and gas pipeline systems, 1-15–1-16 Leak-before-break (LBB) Mandated Organization (MO) application to selected piping systems of WWER type NPPs, Inspector, 5-6 10-7 Mandatory and Recommended Standards, 17-1 Leak-before-break (LBB) approach Manufacturer coolant tube, 16-12–16-13 conformity assessment modules for PHT piping, 16-18–16-13 without QA, 2-8–2-10 Leckie/Penny formulation, 4-8–4-9 drawing up technical documentation, 2-10–2-11 Legislation system, nuclear facilities in Korea, 14-3–14-6 responsibility for design, and conformity assessment and, Table 14.1 (contents of the Nuclear Safety Act), 14-4–14-5 2-17–2-18 Table 14.2 (NSSC notices applicable to reactor facilities), 14-5 self certification, 2-8–2-9 Licensed transportation cask, high burnup Material(s) MP197HB, 20-12–20-14 for construction, PED vs. ASME code, 2-20–2-21 description, 20-13–20-14 PD 5500 (U.K.), 4-3–4-4 Licensing pressure equipment regulatory risk management, 20-2–20-3 PD 5500 (U.K.), 4-3–4-4 Licensing, risk-informed RCC-M, 3-48–3-49 OL3 NPP, experiences of, 8-10–8-13 specifications, of pressure equipment, 2-15–2-17 applications for operating license, 8-12–8-13 Maximum allowable stress (MAS) design phase PRA, 8-11–8-12 for B&PV, Japanese codes, 13-5–13-6, 13-8 plant design changes, examples of, 8-11–8-12 Ministerial Decrees (MD), Belgium, 5-1, 5-2, 5-3–5-4, 5-10, 5-11 risk informed pre- and in-service inspection program, Ministry of Economy, Trade and Industry (METI) 8-12–8-13 advisory bodies for, 13-2–13-3 Licensing process of new designs, PRA in, 8-3–8-6 Notification, 13-2–13-3 risk-informed applications Ordinance, 13-2–13-3 during operation, 8-4–8-6 Mixed uranium/plutonium oxide fuel (MOX), 1-11 personnel training, 8-5–8-6 Mobile boilers, steam generators, 5-2 Licensing system and safety assessment, nuclear facilities in Korea, Moderator, CANDU®, 1-10 14-6–14-9 Moment loading, 4-11–4-12 daily inspection by resident inspectors, 14-7 MP197HB transportation package, 20-12–20-14 enforcement, 14-8–14-9 description, 20-13–20-14 Figure 14.3 (reactor licensing and regulation system), 14-6 Multiple flaws, characterization, 19-6–19-9 Figure 14.4 (reactor inspection process), 14-6–14-9 alignment rules for non-aligned flaws, 19-6–19-8 licensing system, 14-6–14-7 combination rules for aligned flaws, 19-8–19-9 regulatory inspections, 14-7 laminar flaws, grouping of, 19-9–19-11 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-11

700 MWe PHWR design Notices of nuclear safety and security commission, Korea, hook up provision, 16-6–16-7 14-10–14-11 PDHRS, 16-2, 16-6–16-7 Notified Bodies (NB), 2-1, 2-2, 2-9–2-11, 2-20, 2-21, 2-22 safety feature in, 16-2, 16-6–16-7 application to, 2-10 experimental design approval of pressure equipment, 2-12, 2-15–2-16 NAPS. See Narora Atomic Power Station (NAPS) monitoring by, 2-9 Narora Atomic Power Station (NAPS), 16-1–16-2. See also Indian Novatome, 3-4 PHWR (IPHWR), design Nozzles Calandria, 16-8 reinforcing, PD 5500 (U.K.), 4-7–4-8 National Board of Boiler and Pressure Vessel Inspectors NPCIL (Nuclear Power Corporation of India Limited), 16-20–16-21 Belgium, 5-11 NSSC notices applicable to reactor facilities, 14-5 Canada, 1-4 NSTMP (National Science and Technology Major Project), United States, 1-12 17-2–17-3

National Nuclear Regulator Act (NNRA), 12-1–12-2, 12-4 Nuclear boilers and pressure vessels Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 National Science and Technology Major Project (NSTMP), codes and standards 17-2–17-3 China, 17-1–17-18 Natural gas fuel, high-pressure storage cylinders, automotive, Korea, 14-1–14-26. See also Korean regulatory system and Canadian storage, 1-12 codes of nuclear boiler and pressure vessels New Approach to Technical Harmonization and Standards, 2-1, 6-1 design and construction standards, CSA, 1-18–1-28 Nickel-based alloy(s) inservice inspection, Canadian, 1-28–1-36 fuel channel spacers, 1-23–1-24 in Taiwan, 15-1–15-9 Non-aligned flaws, alignment rules for, 19-6–19-8 Nuclear codes and standards Non-coplanar flaws, 19-6–19-8 China Nondestructive examination/testing (NDE/NDT) improvements, 17-2–17-3 PED, manufacturing, 2-13–2-14 global harmonization, 18-1–18-5 program, Hungarian ISI system, 11-3–11-4 Korea, 14-1–14-26 qualification, for ISI, 7-6–7-9 Nuclear energy. See Korean regulatory system and codes of nuclear background, 7-6–7-7 boiler and pressure vessels description, 7-7–7-8 Nuclear Horizontal Waste Storage (NUHOWS®), 20-12 implementation of methodology, 7-8–7-9 Nuclear industry, Spanish regulation, 7-5–7-12 Non-nuclear boiler, pressure vessel, and piping design and qualification of NDT for ISI, 7-6–7-9 ­construction standards, CSA, 1-11–1-18 background, 7-6–7-7 Non-nuclear industry, Spanish regulation, 7-1–7-5 description, 7-7–7-8 installation, inspections, and tests, 7-1, 7-2–7-5 implementation of methodology, 7-8–7-9 ITC-EP-2 (Electric Power Generation Station), 7-2, 7-4–7-5 RI-ISI applications for piping, 7-9–7-12 Non-nuclear sector, French codes, 3-1–3-3 at Spanish NPPs, 7-10–7-12 general, 3-1 Nuclear Power Corporation of India Limited (NPCIL), 16-20–16-21 organization, 3-3 Nuclear power planar components, JSME codes for, 13-13–13-30 PED and European Harmonized Standards, 3-2 Cask Code, 13-25–13-27 publications, updates, interpretations, and inquiries, 3-3 Concrete Containment Vessels Code, 13-21–13-23 scope, 3-1–3-3 CVE-3000 Design, 13-22–13-23 structure, 3-2–3-3 CVE-1000 General, 13-21–13-22 Normative technical documentation (NTD) design and construction code for liquid metal reactors, 13-23–13-25 ASI code for WWER reactor components, 10-1–10-3 elevated temperature material strength standard, 13-24 allowable fracture toughness, 10-2–10-3 elevated temperature structural design, 13-24 facts and reasons, 10-2 organization, 13-23–13-24 procedures and criteria, 10-2 environmental fatigue evaluation code, 13-14–13-16 Section I (Welding and brazing of components and piping of fitness-for-service code, 13-16–13-19 WWER type NPPs), 10-2, 10-11–10-12 LWRs, design and construction code for, 13-13–13-14 Section II (Characteristics of materials for components and pipe wall thinning management code, 13-27–13-29 ­piping of WWER type NPPs), 10-2, 10-12–10-13 severe accident management guidelines, 13-29–13-30 Section III (Strength assessment of components and piping of Nuclear power plants (NPPs) WWER type NPPs), 10-2, 10-13 in Finland, 8-1–8-14 Section IV (Evaluation of residual lifetime of components Loviisa NPP, 8-7–8-10 and piping of WWER type NPPs), 10-2, 10-3, Olkiluoto NPP. See Olkiluoto nuclear power plants 10-13–10-14 risk informed requirements, 8-2–8-3 Section V (Material testing procedures and evaluation), 10-2, in Hungary, Paks, 11-1–11-9 10-3, 10-14–10-15 in Korea (as of January 2015). See Korean regulatory system and Section VI (Air condition systems for WWER type NPPs), codes of nuclear boiler and pressure vessels 10-2 Spain North Anna Earthquake, cosmetic damage to HSM after, 20-10 codes and standards, 7-5–7-12 Notice No. 2012-8 (nuclear safety and security commission in qualification of NDT for ISI, 7-6–7-9 Korea), 14-10–14-11 RI-ISI applications for piping, 7-9–7-12 I-12 • Index

Nuclear reactor cores, 1-6–1-8 Paks nuclear power plant, Hungary, 11-1–11-9 Nuclear regulations of pressure vessels, German NPP, 6-2–6-10 operational life extension, 11-1–11-2 Nuclear Regulators Working Group (NRWG-TF-NDTQ), 7-6 regulatory aspects, 11-2–11-3 Nuclear Regulatory Agency, 13-1, 13-13 regulatory framework, 11-1 Nuclear Regulatory Authorities (NRA), 10-4, 13-1, 13-25, Parallel flaws, 19-6–19-8 13-29 Passive autocatalytic recombiners (PARs), 1-10 Nuclear Regulatory Commission (NRC), 9-5, 13-1, 21-5 Passive Decay Heat Removal System (PDHRS), 16-2, 16-6–16-7 Nuclear regulatory organizations, 14-2–14-3 hook up provision, 16-6–16-7 Figure 14.1 (working mechanism of nuclear safety regulation), Pebble bed modular reactor (PBMR), 12-1, 12-5–12-9 14-2 ASME B&PV Code involvement, 12-9 KINS, 14-3 challenges, 12-9 NSSC, 14-2 design approach in terms of codes and standards, 12-7 Nuclear Safety Act design characteristics, 12-6

contents of, 14-4–14-5 HPB, 12-6, 12-7–12-8 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 Nuclear Safety Monitoring Center, 15-3 origin and description, 12-5–12-6 Nuclear sector, French codes, 3-3–3-8 RIM methodology, 12-9 general, 3-3–3-4 Personnel organization, 3-4–3-6 training, planning of, 8-5–8-6 publications, updates, interpretations, and inquiries, 3-7–3-8 Petroleum and petrochemical plants scope, 3-4 B&PV codes and standards, in Japan, 13-4–13-9 structure of RCC-M, 3-6–3-7 DEIR, 13-3, 13-4–13-5, 13-8 NUHOMS® system HPGSL, 13-3, 13–4-13-5, 13-8 above-ground, advantages of, 20-6–20-11 HPIS, 13-5–13-7 robust canister ESP for aging management, 20-9–20-10 JIS, 13-7–13-9 shielding performance, 20-7–20-8 Petroleum plant, fired-heater pressure coils in, 1-12, 1-13–1-14 unloading operations, safe and simple, 20-8–20-9 PFBR (Prototype Fast breeder Reactor), 16-27 auxiliary equipment, 20-5–20-6 PHWR (pressurized heavy water reactor), 14-1 benefits of, 20-3–20-5 Piping EOS system, advanced high-capacity high-heat load, Canadian standards, 1-4, 1-14–1-15 20-11–20-12 fittings, 1-11–1-18, 1-13–1-14 EOS 89 PWR system capabilities, 20-11–20-12 NPP, nuclear codes for design and analysis, 9-3–9-5 MP197HB transportation package, 20-12–20-14 RI-ISI applications for, 7-9–7-12 at Spanish NPPs, 7-10–7-12 Off-set flaws, 19-6–19-8 RI-ISI program Oil pipeline systems, 1-14–1-18 new vs. old, 8-9–8-10 OJSC Concern Rosenergoatom, 9-2 R.G. YVL E.5, 8-6–8-7 Olkiluoto nuclear power plants stress analysis, formulas for, 9-6, 9-7–9-8 RI-ISI projects, experiences of, 8-10 wall thinning management code, 13-27–13-29 unit 3 (OL3), 3-3, 8-1, 8-2–8-3, 8-5, 8-6 Plant life extension, in Canadian nuclear standards, 1-36–1-37 applications for operating license, 8-12–8-13 Plastic pipelines, Canadian standards, 1-16–1-17 design phase PRA, 8-11–8-12 Plastic strain correction factor (Ke), 3-52, 3-54 EPR safety features, 8-10 Plutonium-239 (by-product), 16-26 plant design changes, examples of, 8-11–8-12 Plutonium stockpiles, 1-11 PRA review findings, 8-11 PNAE, Russian nuclear standard risk-informed licensing, experiences of, 8-10–8-13 ASME BPVC vs., application to seismic analysis of primary loop risk informed pre- and in-service inspection program, of PWR (WWER) reactor, 9-5–9-13 8-12–8-13 bend’s parameter of curved pipe, 9-7–9-8 risk-informing construction and operating license, 8-10 comparative analysis of PCLS, results, 9-8, 9-9–9-13. See also OPB-73, Soviet regulatory document OPB-73, 11-2 Primary Coolant Loop System (PCLS) Operating license (OL) documentation of guidelines for seismic analysis, 9-5 risk-informed applications for, 8-4, 8-12–8-13 equipment classification, 9-5–9-6 Operation formulas for piping stress analysis, 9-6, 9-7–9-8 Canadian oil and gas pipeline systems, 1-15–1-16 level D service limits, 9-6 risk-informed applications during, 8-4–8-6 N-1210 (Earthquake description), 9-8 personnel training, 8-5–8-6 N-1220 (Methods of dynamic analysis), 9-8 Operational life extension, of Paks NPP, 11-1–11-2 N-1230 (Damping), 9-8 Operational vibration piping components strength analysis, 9-6 European HVD approach in protecting NPP primary and seismic analysis, 9-11 secondary systems, 9-13, 9-14–9-29 seismic loads, definition of, 9-8 Outer diameter (OD), RPV UT from, 11-8–11-9 Postulated defect case, 7-7–7-8 Owner, pressure equipment (Belgium) Power uprate, Taiwan, 15-8–15-9 report Pressure boundary integrity, 15-7–15-8 transposition of “data reports” and, 5-15–5-18 Pressure coils, fired-heater, 1-12, 1-1–1-14 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-13

Pressure equipment Nuclear Islands maintenance, 3-41–3-45 RCC-CW, 3-1 regulations RCC-F, 3-1 Belgium, 5-1–5-13 RCC-M. See Design and Construction Rules for Mechanical Spain, 7-1–7-12 Components of PWR Nuclear Islands (RCC-M) Pressure Equipment Directive (PED) (97/23/EC), 2-1–2-31 RSE-M, 3-1, 3-4, 3-5, 3-58, 3-60 Annex I (Essential Safety Requirements) plants, class 1 piping segments classification for, 7-10 basic principles, 2-11–2-12 plants, inspection areas resulting from ASME Section XI and design, 2-12–2-13 RI-ISI Programs for Class 1 piping, 7-10, 7-11 manufacturing, 2-13–2-15 PNAE vs. ASME BPVC, application to seismic analysis of Annex III (Conformity Assessment Procedures), 2-2, 2-3, ­primary loop of PWR (WWER) reactor, 9-5–9-13 2-8–2-11, 3-9, 3-24, 3-36 bend’s parameter of curved pipe, 9-7–9-8 Article 9 (Obligations of Distributors), 2-3, 3-9, 3-24, 3-36 comparative analysis of PCLS, results, 9-8, 9-9–9-13. See also

Article 10 (Obligations Apply to Importers and Distributors), 2-3, Primary Coolant Loop System (PCLS) Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 3-9, 3-24, 3-36 documentation of guidelines for seismic analysis, 9-5 Article 11 (Identification of Economic Operators), 2-3 equipment classification, 9-5–9-6 Article 14 (Conformity Assessment Procedures), 2-3, 2-4, formulas for piping stress analysis, 9-6, 9-7–9-8 2-8–2-11 level D service limits, 9-6 Article 16 (User inspectorates), 2-3, 2-4 N-1210 (Earthquake description), 9-8 ASME Code vs., 2-20–2-21 N-1220 (Methods of dynamic analysis), 9-8 classification of pressure equipment, 2-4–2-8 N-1230 (Damping), 9-8 conformity assessment categories (I to IV), 4-2 piping components strength analysis, 9-6 conformity assessment modules, 2-4, 2-8–2-11, 4-2 PNAE Code G-7-002-86, 9-7 conformity assessment procedures, 2-1, 2-2, 2-4, 2-6, 2-8–2-11, seismic loads, definition of, 9-8 3-9, 3-24, 3-26, 3-31, 3-36, 3-38, 4-2 pressure boundary integrity, 15-8 development, 2-2–2-4 Sizewell B PWR (UK), 6-2 ESRs, 2-1, 2-2–2-3, 2-4, 2-6, 2-7, 2-11–2-15, 2-18, 2-20 specification for defect manufacturing in, 7-7 basic principles, 2-11–2-12 Primary Coolant Loop System (PCLS) design, 2-12–2-13 comparative analysis, by PNAE and ASME BPVC, 9-8, 9-9–9-13 manufacturing, 2-13–2-15 results, 9-11, 9-12–9-13 for material, 2-16–2-17 Primary heat transport system (PHTS), 1-5–1-8, 1-21, 1-22, 1-23, 1-31 essential safety requirements, 2-24–2-31 feeders, 16-9–16-10 German legal adaptations and, 6-1–6-2, 6-3, 6-4, 6-5 Reactor Headers, 16-9 historical background, 2-2 Steam Generator, 16-8–16-9 marking and labeling, 2-14–2-15 Primary Reference Response (PRR), 11-3–11-4 material specifications, 2-15–2-17, 2-21 Primary Shutdown System (PSS), seismic testing, 16-14–16-15 NDT, 2-13–2-14 Primary Shutdown System (SDS-1), IPHWR, 16-14–16-15 Notified Bodies (NB), 2-1, 2-2, 2-9–2-11, 2-12, 2-14, 2-15, 2-16, Probabilistic risk assessment (PRA) studies, Finnish 2-18, 2-20, 2-21, 4-2 design phase, for OL3 NPP, 8-11–8-12 PD 5500 (U.K.) vs., 4-1–4-3 applications for operating license, 8-12–8-13 pressure equipment, RGPT (Belgium), 5-2–5-3 plant design changes, examples of, 8-11–8-12 RCC-M, links with, 3-46–3-47 risk informed pre- and in-service inspection program, 8-12–8-13 Pressure equipments, China, 17-9–17-10 regulatory process, 8-2–8-6 technical safety regulations and standards, 17-13–17-18 condition of systems, structures, and components, 8-5 relevant reference standards, 17-14–17-15 construction license, risk-informed applications for, 8-4 Pressure pipes, regulations and standards, 17-16–17-17 disturbance and emergency operation procedures, 8-5 Pressure regulators, in scope of PED, 2-4 exemption from technical specifications, 8-5 Pressure vessels licensing process of new designs, 8-3–8-6 Canadian standards, 1-2–1-11, 1-12 operating license, risk-informed applications for, 8-4 in Germany, 6-1–6-10 operation, risk-informed applications during, 8-4–8-6 inservice inspection, Canadian, 1-28–1-36 outage risks, 8-5 Japanese codes, 13-2–13-4 personnel training, 8-5–8-6 maintenance, 3-41–3-45 plant modifications, 8-5 regulations and standards, China, 17-15–17-16 reporting of operational events, 8-5 Pressurized heavy water reactor (PHWR), 1-5, 14-1 risk informed requirements for NPPs, development of, 8-2–8-3 ALARA, 16-4 technical specifications, 8-5 IC, 16-4 Prototype Fast breeder Reactor (PFBR), 16-27 India and, 16-1–16-24. See also Indian PHWR (IPHWR), design Proximity rule, for subsurface flaw, 19-4–19-6 RCC, 16-4 PSS (Primary Shutdown System), seismic testing, 16-14–16-15 Pressurized water reactors (PWRs), 14-1 Published Document (PD) in CANDU® design, 1-5 PD 6439 (A review of methods of calculating stresses due to local French codes, 3-45 loads and local attachments of pressure vessels), 4-3 JSME S NG1-2006 Code, 13-28, 13-29 PD 6550 (Explanatory supplement to BS 5500:1988), 4-3, 4-5, 4-16 I-14 • Index

Published Document (PD) 5500 (United Kingdom), 2-12, 4-1–4-22 Reactor protection system (RPS), 15-6 bolted flanged joints, 4-8–4-9 Reactor Regulation Act, 13-25 design, 4-4–4-16 Reactor vessel, component of NSSS, 5-3 design for fatigue, 4-12–4-14 Reactor vessel internals (RVI) Figure 4.11B (Dished End Thicknesses for 10% Torispherical components, 10-8 Form), 4-20 lifetime assessment, 10-7–10-8 Figure 51.11C (Dished End Thicknesses for 6% Torispherical lifetime on strength criterion, logic diagram for calculation of, Form), 4-20 10-9 inspection, 4-10–4-11 materials, basic stressors acting on, 10-8 jacketed vessels, 4-9–4-10 strength assessment, 10-8 loads, local, 4-11–4-12 Recognized third-party organization (RTPO) materials, 4-3–4-4 approval of joining procedure qualifications, 2-13–2-14 nozzle reinforcing, 4-7–4-8 Recyclable-Fuel Storage Center, 13-25

PED vs., 4-1–4-3 Recyclable-Fuel Storage Company, 13-25 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 shells under external pressure, 4-5–4-7 Registration shells under internal pressure, 4-4–4-5 Canadian standards, 1-12–1-13 testing, 4-10–4-11 CANDU® nuclear power plants, 1-19–1-20 PWRs. See Pressurized water reactors (PWRs) Registration numbers Canadian, 1-12–1-13 Qinshan CANDU® 6 reactor, 1–10 Regulation on Technical Standards for Reactor Facilities, etc. (101), Qualification, of NDT for ISI, 7-6–7-9 14-1 background, 7-6–7-7 Regulatory authority description, 7-7–7-8 non-nuclear boiler and pressure vessels in China, 17-10–17-13 implementation of methodology, 7-8–7-9 nuclear boiler and pressure vessels in Taiwan, 15-1–15-4, 15-3 Quality assurance (QA) Regulatory Guide 1.13, 21-2–21-3 conformity assessment modules without, 2-8–2-10 Regulatory Guides Quasi-laminar flaws NRC document, 9-5 alternative grouping of, 19-10–19-11 RD-03-42-97, 9-2 RD EO 0186-00, 9-2 RAB (Reactor Auxiliary Building), 16-6 R.G. 1.26, 5-3, 5-4 Radial stress in IC dome, 16-19–16-20 R.G. 1.143, 5-3, 5-4 Radiation Monitoring Center (RMC), 15-1, 15-2 RTM 108.020.37-81, 9-3 Radioactive waste management YVL 2.8, 8-2, 8-3, 8-6 EC6® reactor, 1-10 YVL 3.8, 8-2, 8-6, 11-5 Radwaste canister (RWC), NUHOWS®, 20-12–20-14 YVL A.7, 8-3, 8-4, 8-5, 8-6, 8-11 Rajasthan Atomic Power Station (RAPS), 16-1, 16-2. See also Indian YVL A.10, 8-5 PHWR (IPHWR), design YVL E.5, 8-2, 8-6–8-10 RAPS (Rajasthan Atomic Power Station), 16-1, 16-2 Regulatory inspections, nuclear facilities in Korea, 14-7 Reactivation, Canadian oil and gas pipeline systems, 1-15–1-16 Regulatory structure, IPHWR, 16-7 Reactor Auxiliary Building (RAB), 16–6 Table 16A.2 (partial list of regulatory documents related to nuclear Reactor building, 1-8 power plants), 16-7 EC6®, 1-9–1-10 Reinforced composite pipelines, Canadian standards, 1-16–1-17 Reactor cavity conditioning system (RCCS), function of, 12-8 Relevant reference standards, pressure equipment, 17-14–17-15 Reactor components, IPHWR design, 16-13 Repair and replacement Calandria tube, 16-13 methods, Canadian oil and gas pipeline systems, 1-15–1-16 coolant channel components, 16-13 “Research on China’s Advanced Nuclear Power Standard System,” Reactor Coolant Pump (RCP) design, 8-11 17-2–17-3 Reactor design Research Reactors (RR), 3-4 CANDU®, 1-5–1-8 Resident inspectors, daily inspection by, 14-7 EC6®, 1-8 Result(s) Reactor Harmonization Working Group (RHWG), 11-1 comparative seismic analysis, 9-11, 9-12–9-13 Reactor Header, 16-9 RGPT (Règlement Général pour la Protection du Travail), 5-1–5-4 Reactor Inlet Headers (RIH), 16-9 derogation, 5-3–5-4 Reactor inspection process, 14-6–14-9 PED, 5-2–5-3 Reactor licensing and regulation system, 14-6 Title 3, 5-1–5-2 Reactor Outlet Headers (ROH), 16-9 RIH (Reacto–Inlet Headers), 16-9 Reactor pressure vessels (RPVs) Risk-informed activities, NPPs in Finland internals, 11-5 grading of regulatory authority activities, 8-13 French codes, 3-54 Risk-informed applications, NPPs in Finland in PBMR, 12-7, 12-8 for operating license, 8-4, 8-12–8-13 UT from OD, 11-8–11-9 during operation, 8-4–8-6 WWER, allowable stress intensity limits for, 10-10 personnel training, 8-5–8-6 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-15

Risk Informed In-Service Inspection (RI-ISI) formulas for piping stress analysis, 9-6, 9–7–9-8 applications, in Finland, 8-6–8-10 level D service limits, 9-6 Loviisa NPP RI-ISI projects, experiences of, 8-7–8-10 N-1210 (Earthquake description), 9-8 Olkiluoto NPP RI-ISI projects, experiences of, 8-10 N-1220 (Methods of dynamic analysis), 9-8 selection process, piping required by Reg. guide YVL E.5, N-1230 (Damping), 9-8 8-6–8-7 piping components strength analysis, 9-6 applications for piping, 7-9–7-12 seismic loads, definition of, 9-8 at Spanish NPPs, 7-10–7-12 system and list of standards, state safety regulation, 9-2–9-3 new program for piping, old program vs., 8-9–8-10 Table 9.8 (Damping Values for Pipes According to ASME BPVC), NPPs in Finland, 8-1, 8-2–8-3, 8-4 9-9 Risk informed requirements, for NPPs, 8-2–8-3 Risk management, licensing regulatory, 20, 2-20-3 SAFARI-l research reactor, 12-1, 12-4 RMC (Radiation Monitoring Center), 15-1, 15-2 Safety assessment, nuclear facilities in Korea, 14-6–14-9

ROH (Reactor Outlet Headers), 16-9 Safety factors, French codes, 3-29 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 Rosatom, 9-2 Safety features Royal Decrees (RD) in 700 MWe PHWR design, 16-2, 16-6–16-7 Belgium, 5-1, 5-2, 5-3–5-4, 5-5, 5-8, 5-10, 5-11, 5-12 pressure equipment, China, 17-11–17-12 RPS (reactor protection system), 15-6 Safety systems, EC6® reactor, 1-8, 1-9–1-10 RSE-M (In-service Inspection Rules for Mechanical Components of Safety valves PWR nuclear islands), 3-1, 3-4, 3-5, 3-46, 3-59–3-60, 3-62, FSAR mandatory inspections, 5-5 19-4, 19-8, 19-9 SDS-1 (Primary Shutdown System), IPHWR, 16-14–16-15 RTNDT brittle to ductile transition temperature determination, Section I (Power Boilers), 2-20, 3-6, 13-2–13-4, 13-11, 13-12 French codes, 3-56 Section III, Division 1 (Rules for Construction of Nuclear Power Russia, regulation and codes in nuclear power, 9-1–9-29 Plant Components), 1-18–1-20, 3-6, 5-3, 13-13, 13-26, 13-27, brief history, 9-1 13-31 design and analysis of NPPs equipment and piping, 9-3–9-5 Section III, Division 2 (Code for Concrete Reactor Vessels and European HVD approach in protecting NPP primary and Containments), 13-22 secondary systems, 9-13, 9-14–9-29 Class MC, components, provisions for, 1-18 general operational characteristics, 9-14–9-15, 9-16, 9-17 Subsection NCA, 1-19–1-20, 13-13 HVD, application of, 9-15–9-26 NCA-1271, 1-22 in seismic and vibration isolation systems for nuclear structures NCA-2142, 1-20 and components, 9-20, 9-21, 9-22, 9-24, 9-27–9-29 NCA-2143, 1-20 Fig 9.14 (Verification Tests of HVD at IHI 35 Tons Shaking NCA-3250, 1-20 Table), 9-17 NCA-3550, 1-20 Figure 9.32 (Installation of HVDs at Feed-Water Piping with Section VIII, Division 1 (Rules for Construction of Pressure Connection to Floor), 9-26 Vessels), 2-20, 5-3, 13-2–13-4, 13-5, 13-7, 13-9, 13-11, Figure 9.11 (Stress Distribution in PCLS Subjected to Seismic 13-12 Loads), 9-14 compared to CODAP rules, French codes, 3-21 Figure 9.15 (Experimental Results of Verification Tests of HVD Section XI, (In-Service inspection of NPP) at IHI 35 Tons Shaking Table with Two Dampers Installed at Nonmandatory Appendix R, 8-6, 8-12 Model of BWR Piping), 9-17 failure potential categorization, 8-9 Figure 9.19 (Wear of Piping Rod Hanger and Collapse of Piping transposition, in Belgian environment, 5-3, 5-4, 5-5–5-12 Support Due to Operational Vibration), 9-20 Appendix IX (Application Rules of ASME Code Section Figure 9.33 (FE Structural Model of 1200 MWT NPP Reactor III or Other Regulations for Repair or Replacement of Building), 9-27 Components in Operating Nuclear Units), 5-9–5-11 Figure 9.35 (Floor Responce Spectra at Reactor’s Supports Appendix X (Independent Body Distinct from Mandated Elevation), 9-27 Organization Performing AIA Role), 5-11–5-12 Figure 9.36 (Riera Load Function and Response Spectra for Non- derogations, 5-5–5-6 Isolated and Isolated NPP Reactor Building), 9-28 Subsection IWA (General Requirements), 5-6–5-8 Figure 9.37 (Isolation of Spent Fuel Storage Tank Against Seismic Subsection IWB (Requirements for Class 1 Components of Event), 9-28 Light-Water Cooled Plants), 5-8–5-9 Figure 9.38 (Seismic Isolation of NPP’s Steam Generator by HVD Section XI, Appendix IX (Application Rules of ASME Code Section Application), 9-28 III or Other Regulations for Repair or Replacement of Figure 9.40 (Acceleration at Top of Non-Isolated and BCS Isolated Components in Operating Nuclear Units), 5-9–5-11 Structure), 9-29 Article IX-2000 (Application of ASME Code Section III), PNAE vs. ASME BPVC, application to seismic analysis of 5-9–5-10 primary loop of PWR (WWER) reactor, 9-5–9-13 Article IX-3000 (Application of Other Rules Than Those of ASME bend’s parameter of curved pipe, 9-7–9-8 Code) comparative analysis of PCLS, results, 9-8, 9-9–9-13. See also IX-3200 (Additional Rules Applicable to Components under Primary Coolant Loop System (PCLS) IX-2000), 5-10 documentation of guidelines for seismic analysis, 9-5 Article IX-5000 (Transposition of ASME Code Section III), equipment classification, 9-5–9-6 5-10–5-11 I-16 • Index

Section XI, Subsection IWA (General Requirements), 5-6–5-8 SFSB (Spent Fuel Storage Bay), 16-10 Article IWA-1000 (Scope and Responsibility) SG (Steam Generators), 16-8–16-9 IWA-1310 (Components Subject to Inspection and Testing), 5-6 Shakedown factors, 4-8, 4-12 Article IWA-2000 (Examination and Inspection) Shielding performance, NUHOMS® system, 20-7–20-8 IWA-2110 (Duties of Mandated Organization Inspector), 5-5, Shutdown system, CANDU® 6 reactor, 1-7 5-6 Single flaw, characterization, 19-1–19-6 IWA-2111 (Scope of Work of FANC and Its TSO Bel-V), 5-5, proximity rule for subsurface flaw, 19-4–19-6 5-6–5-7 SNPTC (State Nuclear Power Technology Company), 17-3–17-4 IWA-2120 (Qualification of Authorized Inspection Agencies, Snubbers Inspectors, and Supervisors), 5-7 FSAR mandatory inspections, 5-5 Article IWA-6000 (Records and Reports), 5-7–5-8 Sodium Fast Reactors (SFR), 3-4 Section XI, Subsection IWB (Requirements for Class 1 Components Sour service pipelines, Canadian standards, 1-17–1-18 of Light-Water Cooled Plants), 5-8–5-9 South Africa, nuclear industry in, 12-1–12-9

Article IWB 3000 (Acceptance Standards), 5-8–5-9 Figure 12.5 (Passive Heat Transport Path from Core to Heat Sink), Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 IWB-3630 (Acceptance Criteria for Steam Generator Tubing), 12-8 5-9 historical perspective, 12-1 IWB-3730 (Fracture Toughness Criteria for Protection Against nuclear code and standards usage, 12-3–12-9 Failure), 5-9 Koeberg power station, 12-4–12-5 Seismic analysis of primary loop of PWR (WWER) reactor, PNAE PBMR, 12-5–12-9 vs. ASME BPVC in application, 9-5–9-13 regulatory control of pressurized equipment, 12-1–12-3 comparative analysis of PCLS, results, 9-8, 9-9–9-13 Department of Labour, 12-2 brief description, 9-8, 9-9 NNRA, 12-1–12-2, 12-4 dynamic analysis model, 9-9 SABS, 12-3 input seismic excitation, 9-9, 9-10 SANAS, 12-3 results, 9-11, 9-12–9-13 South African National Nuclear Regulator (NNR), 12-2, 12-4, 12-5 documentation of guidelines, 9-5 Spain, pressure equipment regulations, 7-1–7-12 equipment classification, 9-5–9-6 nuclear industry, codes and standards in, 7-5–7-12 formulas for piping stress analysis, 9-6, 9-7–9-8 qualification of NDT for ISI, 7-6–7-9 bend’s parameter of curved pipe, 9-7–9-8 RI-ISI applications for piping, 7-9–7-12 N-1210 (Earthquake description), 9-8 regulation in non-nuclear industry, 7-1–7-5 N-1220 (Methods of dynamic analysis), 9-8 installation, inspections, and tests, 7-1, 7-2–7-5 N-1230 (Damping), 9-8 ITC-EP-2 (Electric Power Generation Station), 7-2, 7-4–7-5 PNAE Code G-7-002-86, 9-7 Table 7.1 (Classification of Pressure Equipment According to seismic loads, definition of, 9-8 R.D. 769/1999), 7-2 level D service limits, 9-6 Table 7.3 (Equipment Subjected to Action of Flame/Heat Input), piping components strength analysis, 9-6 7-3 Seismic design, nuclear boiler and pressure vessels in Taiwan, Spent fuel storage, 21-1–21-7 15-5–15-7 dry cask storage, 21-4–21-7 automatic seismic trip system, 15-5–15-6 spent fuel pools, 21-1–21-4 commercial-grade items for replacement items, 15-6–15-7 design requirements, 21-2–21-3 earthquake experiences, 15-5–15-6 Spent Fuel Storage Bay (SFSB), 16-10 Figure 15.5 (tectonic plates around Taiwan island, after Angelier Stainless steels 1986), 15-4 for pressure equipment Figure 15.7 (recorded vs. design response spectrum, 20061226 Japanese codes, 13-24–13-25, 13-31 hengchun earthquake, in Maanshan NPP at reactor basemat PD 5500 (U.K.), 4-3–4-4 floor, NS direction), 15-6 specific types Figure 15.8 (seismic hazard curve in Taiwan’s NPP sites, earth- Mod.9Cr-1Mo steel, 13-24–13-25 quake data up to February 2007), 15-7 Standards Development Organizations (SDOs) Seismic forces, 13-32–13-33 Code-Convergence Board, establishment of, 18-4–18-5 Seismic isolation systems, HVDs in, 9-20, 9-21, 9-22, 9-24, State Mining Regulatory Body, 9-1 9-27–9-29 State Nuclear Power Technology Company (SNPTC), 17-3–17-4 Seismic loads Steam generators (SG), 16-8–16-9 European HVD approach in protecting NPP primary and low pressure, 5-2 secondary systems from, 9-13, 9-14–9-29 NSSS, components, 5-3 Seismic qualification, IPHWR, 16-13–16-15 pressure components, 5-2 Seismic rules, Japan, BPVC code and standards, 13-31–13-35 tubing, 1-31, 5-5 design for earthquake and tsunami, 13-32–13-33 Steam piston engines, 5-2 enforcement of new regulation requirements, 13-31–13-32 Steam storage vessels, 5-2 JSME S NA1-2012, stresses for FFS code, 13-33–13-35 Stoomwezen, 4-21 Seismic testing of PSS, 16-14–16-15 Storage tanks, French codes, 3-54 Separation distance, vertical/horizontal/direct, 19-6–19-8, 19-11 Stress(es) Service Building, EC6® reactor, 1-10 French codes, 3-51–3-54 Severe accident management guidelines, JSME codes, 13-29–13-30 local loads on cylindrical shells, PD 5500 (U.K.), 4-11–4-12 GLOBAL APPLICATIONS OF THE ASME BOILER & PRESSURE VESSEL CODE • I-17

PD 5500 (U.K.), 4-4, 4-11–4-12 Tensile strength, PED limit, 4-2–4-3 pressure equipment, EN 13445, 4-18–4-19 Testing primary and secondary PD 5500 (U.K.), 4-10–4-11 PD 5500 (U.K.), 4-15–4-16 pressure equipment, French codes, 3-56–3-57 seismic Spanish regulation in non-nuclear industry, 7-1, 7-2–7-5 for FFS code, JSME S NA1-2012, 13-33–13-35 ITC-EP-2 (Electric Power Generation Station), 7-2, 7-4–7-5 Stress reduction factor, 4-9 Test pressure, of pressure equipment, PD 5500 (U.K.), 4-10–4-11 Structural factor (SF) (safety factor) “The Guidelines for the Application of the Korea Electric Power pressure vessel, PD 5500 (U.K.), 4-11 Industry Code as the Technical Standards for the Nuclear Subsurface flaw Power Reactor and its Related Facilities,” 14-2 combination rules for aligned flaws, 19-8–19-9 Thermal power plant components, JSME codes for, 13-10–13-12 proximity rule for, 19-4–19-6 organization of code, 13-11–13-12 transformation procedure, 19-4–19-6 Time-of-flight diffraction (TOFD) method, 17-10

Supervision, nuclear codes in China, 17-7–17-8 TIRM (Taipower Integrated Risk Monitor), 15-6 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 Surface flaw TOFD (time-of-flight diffraction) method, 17-10 combination rules for aligned flaws, 19-8–19-9 Tolerance specifications and pressure boundary standards, Canadian transformation from subsurface to, 19-4–19-6 standards, 1-5 System, structure or component (SSC), 1-26–1-27 Toroidal bellows, 3-16 Toroidal field (TF) conductors, 13-30 Taipower Integrated Risk Monitor (TIRM), 15-6 Toughness Taiwan, development of nuclear boiler and pressure vessels fracture. See Fracture Toughness historical background, 15-1 pressure equipment conformance, 2-20 power uprate, 15-8–15-9 Transfer Cask (TC), NUHOMS® design, 20-3–20-5, 20-6, 20-7, pressure boundary integrity, 15-7–15-8 20-10, 20-11 pressure vessel codes of CNS Transfer Trailer, NUHOMS® system, 20-5–20-6 Table 15.5 (CNS codes related to non-nuclear pressure vessels), Transformation, of subsurface to surface flaw, 19-4–19-6 15-9 Transportation, Canadian oil and gas pipeline systems, 1-14–1-15 role of regulatory authority, 15-1–15-4, 15-3 Transportation package, MP197HB, 20-12–20-14 Department of Nuclear Regulation, 15-2–15-3, 15-4 description, 20-13–20-14 Department of Nuclear Technology, 15-3 Transposition, ASME Code Section XI (Belgium), 5-4, 5-5–5-12 Department of Radiation Protection, 15-3 Appendix IX (Application Rules of ASME Code Section III or FCMA, 15-3–15-4 Other Regulations for Repair or Replacement of Components Figure 15.3 (organization structure of the AEC in Taiwan), 15-3 in Operating Nuclear Units), 5-9–5-11 Figure 15.4 (organizational structure of the department of Article IX-2000 (Application of ASME Code Section III), nuclear regulation of the AEC in Taiwan), 15-4 5-9–5-10 seismic design, 15-5–15-7 Appendix X (Independent Body Distinct from Mandated automatic seismic trip system, 15-5–15-6 Organization Performing AIA Role), 5-11–5-12 commercial-grade items for replacement items, 15-6–15-7 derogations, 5-5–5-6 earthquake experiences, 15-5–15-6 Subsection IWA (General Requirements), 5-6–5-8 Figure 15.5 (tectonic plates around Taiwan island, after Angelier Article IWA-6000 (Records and Reports), 5-7–5-8 1986), 15-4 Subsection IWB (Requirements for Class 1 Components of Figure 15.7 (recorded vs. design response spectrum, 20061226 Light-Water Cooled Plants), 5-8–5-9 hengchun earthquake, in Maanshan NPP at reactor Article IWB 3000 (Acceptance Standards), 5-8–5-9 basemat floor, NS direction), 15-6 Tritium, total package limit, 1-19 Figure 15.8 (seismic hazard curve in Taiwan’s NPP sites, TRMS (Taiwan Radiation Monitoring Station), 15-1 earthquake data up to February 2007), 15-7 TSG Z0004-2007 (Basic Requirements for Special Equipment Taiwan Radiation Monitoring Station (TRMS), 15-1 Quality Assurance System on Manufacture, Installation, Task Groups, VERLIFE procedure, 10-3–10-4 Alteration, and Repair), 17-16–17-17 Technical Barriers to Trade (TBT), 14-2, 14-24 Tsunami, design for, 13-32–13-33 Technical committees (TCs), 1-2, 1-4, 1-5, 1-11–1-12 Turbine Building, EC6® reactor, 1-10 Technical documentation of pressure equipment, 2-10–2-11 Ultimate tensile strength (UTS) Technical harmonization/convergence, nuclear codes and standards, of pressure equipment, PD 5500 (U.K.), 4-3–4-4 18-2–18-3 Ultrasonic test (UT) (examination) Technical know how, nuclear reactors in India, 16-28 RPV UT from OD, 11-8–11-9 Technical Safety Organisation (TSO), Bel-V, 5-3–5-4, 5-6–5-7, 5-10 Underwriters’ Laboratories of Canada (ULC), 1-2, 1-4 Technical safety regulations and standards, China, 17-13–17-17 UNF (used nuclear fuel), dry storage system for. See Dry storage relevant reference standards, 17-14–17-15 system Technical specifications, risk-informed applications, 8-4, 8-5, Unified procedure for Lifetime Assessment of Components and 8-12–8-13 Piping in WWER Type Nuclear Power Plants (VERLIFE) Technical standards of nuclear reactor facilities, KEPIC and, procedure, 10-3–10-6, 10-7 14-12–14-14 United Kingdom (U.K.) Technical Standards of Reactor Facilities, 14-3 unfired pressure vessel rules, 4-1–4-22 I-18 • Index

United States (US), Hungarian legislative/regulatory framework vs., Section III (Strength assessment of components and piping of 11-6 WWER type NPPs), 10-2, 10-13 United States Nuclear Regulatory Commission (USNRC), 5-1, 5-3, Section IV (Evaluation of residual lifetime of components and 5-12, 6-2 piping of WWER type NPPs), 10-2, 10-3, 10-13-10-14 Unloading operations, with NUHOMS® system, 20-8–20-9 Section V (Material testing procedures and evaluation), 10-2, Upgrading, Canadian oil and gas pipeline systems, 1-15–1-16 10-3, 10-14–10-15 U.S. Nuclear Regulatory Commission (USNRC), 18-3 Section VI (Air condition systems for WWER type NPPs), 10-2 Usage factor, French codes, 3-52–3-53 PNAE vs. ASME BPVC, application to seismic analysis of Used nuclear fuel (UNF), dry storage system for. See Dry storage primary loop of PWR (WWER) reactor, 9-5–9-13. See also system PNAE, Russian nuclear standard USNRC (U.S. Nuclear Regulatory Commission), 18-3 RPVs and bolting joints, allowable stress intensity limits for, 10-10 US Nuclear Regulatory Commission (US NRC), 1-26 Unified Procedure for, 10-3 WWER-440 reactor pressure vessels, 10-5, 10-6, 10-7 ®

Vacuum drying system, NUHOMS system, 20-5–20-6 WWER-1000 reactor pressure vessels, 10-5, 10-6 Downloaded from http://asmedigitalcollection.asme.org/ebooks/book/chapter-pdf/2802667/861073_bm.pdf by guest on 27 September 2021 Valve(s) Water–Water Energetic Reactor, 17-3 design rules, French codes, 3-53–3-54 Weld efficiency factor, French codes, 3-54 Ventilated Storage Cask VSC-24 system, 21-5 Welding VERLIFE (Unified procedure for Lifetime Assessment of oil and gas pipeline systems, Canadian, 1-14–1-15, 1-16 Components and Piping in WWER Type Nuclear Power Welds(s) Plants) procedure, 10-3–10-6, 10-7 inspection in pressure coils exposed to direct radiant heat, Vibration isolation systems, HVDs in, 9-20, 9-21, 9-22, 9-24, Canadian standards, 1-12 9-27–9-29 joint factors, U.K. rules, 4-1 VVERs, Russian designed PWRs permissible joints in pressure vessels, French codes, 3-18–3-19 ISI of VVER units, characteristics of, 11-8–11-9 Western European Nuclear Regulators’ Association (WENRA), 11-1 RPV UT from OD, 11-8–11-9 Working Group Pressure (WGP) type NPPs, 11-5–11-6 Guideline 7/24, 2-16–2-17 World Association of Nuclear Operators (WANO), 16-21 Wall thinning management code, in piping systems, 13-27–13-29 World Trade Organization (WTO), 9-4 WANO (World Association of Nuclear Operators), 16-21 WTO/TBT Agreement, 13-2 Water water energetical reactor (WWER) WWER. See Water water energetical reactor (WWER) components, NTD ASI code for, 10-1–10-3 allowable fracture toughness, 10-2–10-3 Yield Strength (YS) facts and reasons, 10-2 of pressure equipment, PD 5500 (U.K.), 4-3–4-4 procedures and criteria, 10-2 YVL guides Section I (Welding and brazing of components and piping of YVL E.5, 8-2, 8-6–8-10 WWER type NPPs), 10-2, 10-11–10-12 Section II (Characteristics of materials for components and Zirconium alloys piping of WWER type NPPs), 10-2, 10-12–10-13 for fuel channel pressure tubes, 1-21, 1-23, 1-35–1-36, 31