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INDEX

A1 and A2 values, radioactive material packaging, 348, 350, 351Q-system for calculation of, 345, 350, 353, 356

Abrasion, of pressure equipment, 153AC. See Alternating current.Accelerated intergranular corrosion test, French codes, 249, 253Acceptance criteria, of age management program, 58Access door, in pressure equipment, 152Accident sequence analysis, 93Accreditation, of Canadian organizations developing standards, 160Accredited standards-developing organizations (SDOs), 160Acetylene gas, compressed, 260ACI. See American Concrete Institute.Acoustic circuit analyses, 6Acoustic emission, 254

CODAP future specifications, 208ACR®. See Advanced CANDU® Reactor.ACRS. See Advisory Committee on Reactor Safeguards.Active component failure rate, 96Active power plant structures/components, surveillance and

maintenance programs, 31ACVG. See Alternating current voltage gradient method.Addenda to the Code

1972 Addenda to Section III, Appendix G, 441983 Addenda to Section XI, 1161988 Addenda to Section XI, 7, 1181994 Addenda (2004 Edition as revision), 296, 298, 2991999 Addenda, 271, 307, 666, 6672001 Addenda, 298, 668, 6732002 Addenda to Section XI, 118, 119, 121,

Section XI, Appendix C, 19–20, 21, 22, 118, 1262003 Addenda, 668

Adjustment factor (Ke factor), 273. See also Ke factor.AD Merkblätter code, 316, 329, 330Administrative Procedure Act of 1946 (APA), 338, 594AD 2000, 139, 553, 554, 555, 557, 561,Advanced CANDU® Reactor (ACR®), 188Advisory Committee for Energy, Nuclear and Industrial Safety

Subcommittee, 259Advisory Committee on Reactor Safeguards (ACRS), 505Advanced Notice of Proposed Rulemaking, 350Advantica (formerly BG Technology), 400

AE. See Aging effect. AEA. See Atomic Energy Act.AEC. See U.S. Atomic Energy Agency.Aerospace Material Specifications, materials standards, 163A0 factor, 275AFCEN. See French association for design, construction and

inservice inspection rules for nuclear island components.AFCEN Quality Manual, 197AFIAP. See Association Française de Ingenieurs en Appareils à

Pression.AFNOR. See French Standardization Organization.AGA. See American Gas Association.Aging

fitness-for-service rules (Japan), 276indicators, 58managing the effects of, 35pressure equipment conformance, 142preventative action, 35

Aging degradation, 58–59Aging effect (AE), 58–59

environmental, 38Aging management, of pressurized water reactor (PWR) vessel

internals, 57–60Aging management program (AMP), 21, 30–31, 33–35, 39, 41, 58–59

audits, 33–34, 36–37during extended operation, 38elements, 35, 57environmental aging effect and, 41GALL Report and, 33–34license renewal and, 32plant-specific, 37–38, 57

Aging management review (AMR), 30–35, 38, 41, 57Aging management strategies, 59–60AI. See Authorized Inspector. AIA. See Authorized Inspection Agencies.Air environments

austenitic stainless steels fatigue crack growth rate, 21ferritic steels fatigue crack growth rate, 21

Air conditioning, Japanese codes, 261Air Conditioning and Refrigeration Institute, cooling equipment

standards, 163

Additional page numbers with information about individual ASME Specifications (SA and SB numbers) can be found under the headings“American Society of Mechanical Engineers Ferrous Material Specifications” and “American Society of Mechanical EngineersNonferrous Material Specifications.” Referrals to Code Paragraphs and Sections can be located by their alphabetical code (NA, NB, etc.).

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688 • Index

Air transport, of radioactive materials, 347, 352, 353AISC. See American Institute for Steel Construction. ALARA. See As Low as Reasonably Achievable.ALARP. See As Low As is Reasonably Practical region.Allowable flaw depth, 9, 118Allowable pressure

brittle fracture and, 55equation for, 49

Allowable pressure operating curve, 48–49Allowable pressure temperature (P-T) limits, 49, 51Allowable stresses, 149

austenitic stainless steel, PED and U.K., 310–311French codes, 191, 193, 196, 253, 653ferritic materials, U.K., 310in pressure equipment, 139, 143, 151–152, 157pressure equipment, Japanese codes, 44, 257, 258, 259pressure equipment, PED and European codes, 148, 314pressure vessels, EN 13445, 326–327pressure vessel, French codes, 191, 193, 196, 253, 653seismic design, Japanese codes, 44, 257, 258, 259transport tanks, 365, 366, 368welded joints, French codes, 246

Alloy ductile iron castings, for pressure equipment, French codes, 242Alloy steels

for pressure equipment, French codes, 252for pressure equipment, Japanese codes, 286

Alternating current (AC), loss of power, 33, 42, 395Alternating current voltage gradient (ACVG) method, for pipeline

system assessment, 395Alternating stress intensity

of containment vessels for radioactive materials, 346–347of pressure vessels, PD 5500 (U.K.), 321–324

Aluminumallowable stresses, pressure equipment, 139non-alloyed, allowable stresses in pressure equipment, 157in pressure equipment, 157for pressure equipment, PD 5500 (U.K.), 311for pressure vessels, French codes, 201, 202, 205, 208for pressure vessels, Japanese codes, 263–264, 286

Aluminum alloysallowable stresses, pressure equipment, 139in pressure equipment, 157for pressure equipment, PD 5500 (U.K.), 311–312for pressure equipment, PED codes, 311for pressure vessels, French codes, 201, 202, 205, 208for pressure vessels, Japanese codes, 263–264, 286

Aluminum/nitrogen2 ratios, minimum values for pressure equipment,143

Aluminum piping, Canadian standards, 172American Gas Association (AGA), 395American Institute for Steel Construction (AISC), AISC N-690,

Subsection NF, 247, 255, 667, 675, 676American National Standards Institute (ANSI), 162–163. See also

American Society of Mechanical Engineers Codes andStandards, specific types.

/ASQCZ1.4, 188B16.34, 246K61.1/CGAG-2.1, 188,N14.1, 351NB-23, 366NGV2-2000 (Basic Requirements for Compressed Natural Gas

Vehicle (NGV) Fuel Containers), 170

American National Standards Institute/American Society ofMechanical Engineers B31 G Manual, 376, 397–398, 401

B31 G assessment criterion, 398–401American National Standards Institute Committee N14, 61, 351American Nuclear Society (ANS)

ANSI/ANS-56.8-2002 (Containment System Leakage TestingRequirements), 186, 189

Nuclear Risk Management Coordinating Committee (NRMCC)with ASME and NRC, 108, 110

Probabilistic Risk Assessment (PRA) Standards, 108, 109, 110risk-informed safety classification efforts, 108Risk-Informed Standards committee (RISC), 103, 109

PRA standards development, 109, 110RISC-2, 103

Seismic and External Events Standard, 104Subcommittee 28, 107

American Nuclear Society (ANS) Standards, specific types 53.1 (Nuclear Safety Criteria for the Design of Modular Helium

Cooled Reactor Plants), 10958.21, 110, 112

American Petroleum Institute (API), 162American Petroleum Institute (API) Pressure Vessel Inspection Code

standards, specific typesAPI 530, 162, 170, 188API 579, 121API 1104 (Acceptance Standards of Production Welds), 400API 1160 (Managing System Integrity for Hazardous Liquid

Pipelines), 377, 380American Society for Nondestructive Testing (ASNT), 148, 264,

ASNT TC 1A (Personnel Approval), 148Master Curve test method, 43materials for pressure equipment construction, 147steels, toughness conformance, 147

American Society for Testing and Materials (ASTM) SpecialTechnical Publication (STP)

STP514, 127, 128STP 536, 127STP 668, 127STP 803 (Deformation Plasticity Failure Assessment Diagram

Approach to Flaw Evaluation), 27, 127, 127STP 803, Vol. 2, 27, 127, 128STP 896 (Deformation Plasticity Failure Assessment Diagram),

114, 119, 123, 128, 648STP 1046, 25

American Society for Testing and Materials (ASTM) Subcommittee,E 10.02, 54

American Society for Testing and Materials (ASTM) test methods,specific types

A 240, 311B 350, 177B 353, 177C 597-02 (Pulse Velocity through Concrete), 186, 189C 805-02 (Impact/Rebound Hammer Tests), 186, 189E-208 (Drop Weight Test), 360E 208-87a (Drop Weight Test), 50E 370-88a (Charpy V-Notch Test), 50E 399 (Cleavage Fracture Toughness), 52E 813-81 (Standard Test Method for Fracture Toughness), 114, 127E 900-02 (Guide for Predicting Radiation-Induced Transition

Temperature Shift in Reactor Vessel Materials), 54–55, 61E 900-87 (Standard Guide for Predicting Neutron Radiation

Damage to Reactor Vessel Materials), 54, 61



COMPANION GUIDE TO THE ASME BOILER & PRESSURE VESSEL CODE • 689

E 1921, 53E 1921-97, 61

American Society of Mechanical Engineers (ASME), 162, 163Class 1 ferritic piping, flaw evaluation procedures, 118Code cases, 103, 104, 106, 107, 108nameplate, removal of, 366safety factor, 149specifications, for steel, 149Subcommittee VIII, 208Subcommittee XII, Transport Tanks, 357

American Society of Mechanical Engineers Board on Nuclear Codesand Standards (BNCS), 103, 107, 109, 189

Code cases, 109–110, 112Committee on Nuclear Risk Management (CNRM), 90, 108environmental fatigue effects and, 21Independent Decision-Making Panels, 110non-mandatory appendices, revising risk-informed, 228, 358Nuclear Air and Gas Treatment Equipment Committee, 107Nuclear Codes and Standards (NS&S) Task Team, 107, 108, 109,

110, 112Nuclear Cranes Committee, 107Nuclear Quality Assurance Committee, 107Nuclear Risk Management Coordinating Committee (NRMCC),

108, 110Qualifications of Mechanical Equipment Committee, 107RIP-50 TG (Risk-Informed Part 50 Task Group), 107risk-informed ISI and IST implementation, 90, 95risk initiatives, 110, 112Risk Management Strategic Plan, 107, 108, 109Standards Committees and, 107Subcommittee, Nuclear Accreditation, 107Subcommittee, Section III (Nuclear Power), 108Subcommittee Section III, Division 1, 108Subcommittee Section III, Division 2, 108Subcommittee Section III, Division 3, 108Subcommittee XI (Inservice Inspection), 107, 109, 111, 112Working Group on Implementation of Risk-Based Examination

(IRBE), 94, 97Working Group on Optimization, 97Working Group on Risk, 97

American Society of Mechanical Engineers Board on PressureTechnology Codes and Standards (BPTCS), 107

American Society of Mechanical Engineers (ASME) Boiler andPressure Vessel Code (Code), 25, 143, 168, 171, 181,

adoption by USNRC, 353–354allowable crack depth, 24alternate inspection method for nozzle inner radii, 10canister design requirements for radioactive materials, 349comparison of code structure with French codes, 196comparison with Pressure Equipment Directive, 144, 192design fatigue curves, 34, 42environmental fatigue effects, 21identification for Canadian pressure equipment, 169initiatives, 108, 110Master Curve test method, 43pressure equipment directive perspectives, 129–157vs. RCC-M French code, 228–230, 233, 236–243, 246reactor vessel inspection requirements, 71–72requirements quantifying U.S. Type B transportation requirements,

334risk-informed code cases, 107stresses permitted in radioactive material packaging, 340, 341

weld-overlay-type repairs, 18weld repair criteria, 13

American Society of Mechanical Engineers (ASME) Boiler andPressure Vessel Committee, Risk-Informed Code Cases, 90

American Society of Mechanical Engineers (ASME) Boiler andPressure Vessel Standards Committee, 159, 324

American Society of Mechanical Engineers (ASME) Boiler andPressure Vessel Code (BPVC) Working Group on FlawEvaluation, 22, 127

American Society of Mechanical Engineers (ASME) Codes andStandards, 107

B 16.5, 316B 31G, 376, 397–400B 31.1 (Power Piping), 34, 42, 103, 170, 172, 188, 191, 209, 246

vs. CODETI, 216vs. Japanese codes, 270–271

B 31.3 (Process Piping Code), 170, 188,vs. CODETI, 216vs. EN 13445, 208

Table 326.1, 169B 31.4 (Pipeline System Repairs), 170, 188, 395, 405–406 B 31.5 (Refrigeration Piping and Heat Transfer Components), 170,

188B 31.8 (Pipeline Repairs), 377, 395, 403, 405–406, 422B 31.8S (Managing System Integrity of Gas Pipelines), 376–377,

403, 422B 31.9 (Building Services Piping), 170, 188NQA, 101, 102, 108, 109NQA-1, Appendix (Risk-Inform), 101, 102, 108RA-S, 110RA-S-2002, 90, 106–107, 110, 111

Addenda-2003, 106RA-S-2003, 106, 110RA-Sa-2003, 110

Table 1.3-1, 91RA-Sa-2003 Addenda (PRA Standards), 91, 110, 111. See also

American Society of Mechanical Engineers (ASME)Probabilistic Risk Assessment (PRA) Standards.

SNT-TC-1A (Certification of NDE Personnel), 249American Society of Mechanical Engineers (ASME) Codes and

Standards Redesign Process, 90American Society of Mechanical Engineers (ASME) Code Section

XI Working Group on Flaw Evaluation, 46, 118American Society of Mechanical Engineers (ASME) Code Section

XI Working Group on Operating Plant Criteria, 45American Society of Mechanical Engineers (ASME) Commission paper

COMNJD-002 (On Probabilistic Risk Assessment [PRA]), 108COMNJD-03-0002 (On PRA Quality), 108

American Society of Mechanical Engineers (ASME) constructioncode, 169

American Society of Mechanical Engineers (ASME) Council onCodes and Standards, 90

Project Team, to develop PRA Standard, 90–91American Society of Mechanical Engineers (ASME) Ferrous

Material Specifications (SA specifications), specific typesSA 240, 142SA 312, 142SA 370, 359SA 503 Cl.3 (Reactor Vessel Steel Composition), 243SA-508-2, 44SA-508-CLI, 19SA-533-B1, 44



690 • Index

American Society of Mechanical Engineers (ASME) InnovativeTechnologies Institute, 110

American Society of Mechanical Engineers (ASME) NonferrousMaterial Specifications (SB specifications), SB-166,19

American Society of Mechanical Engineers (ASME) Operations andMaintenance Code for Nuclear Power Plants (O&M Code)

Appendix II (Check Valve Condition Monitoring Program), 105Code cases, 109Inservice Inspection (ISI) code, 108Inservice Testing (IST) code, 103, 109OMN-Code 1995 Edition-1996 Addenda, 105OMN-1, 105OMN-3 (Risk Categorization), 103, 104, 105, 106, 112OMN-4 (Treatment of Check Valves), 103–106, 112OMN-4 White Paper, 105, 112OMN-7 (Treatment of Pumps), 103–106, 112OMN-10 (Snubbers), 103–106, 112OMN-11 (Treatment of Motor-Operated Valves), 103–106, 112OMN-12 (Treatment of Pneumatic and Hydraulic Valves),

103–106, 112risk-informed Code Cases, 90, 94, 106, 108, 109, 112Subsection ISTC, code test, 105Subsection ISTC, LSS check valve code test, 105Subsection ISTD (Inservice Testing of Dynamic Restraints

[Snubbers] in Light-Water Reactor Power Plants), 106Subsection ISTE, 103testing strategies for HSS/LSS components, 103

American Society of Mechanical Engineers (ASME) Operations andMaintenance Code for Nuclear Power Plants (O&M Code)Code Committee, 103–104, 109

Task Group on Component Importance Ranking, 103American Society of Mechanical Engineers (ASME) Performance

Test Codes, PTC25, Section 2, 359American Society of Mechanical Engineers (ASME) Probabilistic

Risk Assessment (PRA) Standard, 89–92, 107–110, 112 Addenda to, 107Addendum b, 111content additions, 109evolution of, 90–91Figure 3.1-1 (Flow Chart for Evaluating PRA Capability), 93for external events, 110flow chart for evaluating capability, 93Independent Decision-Making Panel, 94, 100, 110integrate into other ASME risk-informed Codes and Standards, 107for internal events, 110, 112Level 1, 91, 96Level 2, risk-informed, 91, 96, 110Level 3, risk-informed, 110objectives of, 90, 91 potential new standards, 162quality assurance of PRAs, 95scope, 108, 110Section 1 (Introduction/Scope), 92Section 2 (Acronyms, Terms), 91Section 3 (Application of), 91–92Section 4 (Technical Requirements), 91–92Section 5 (Configuration Control), 92Section 6 (Peer Reviews), 92Table 1.3-1 (Capability Categories for PRA), 91

American Society of Mechanical Engineers (ASME) Pressure VesselResearch Council (PVRC) Workshop on the EnvironmentalEffects on Fatigue Performance, 20

American Society of Mechanical Engineers (ASME) Research TaskForce on risk-Based Inservice Testing Guidelines, CRTD-Vol.40-2, 103, 108

American Society of Mechanical Engineers PVHO-1 (SafetyStandard for Pressure Vessels for Human Occupancy), 169,188

American Society of Mechanical Engineers (ASME) SectionSubgroup on Range, 124

American Society of Mechanical Engineers (ASME) website(www.asme.org), 107

American Society of Mechanical Engineers (ASME) Working Groupon Check Valves, 103

American Society of Mechanical Engineers (ASME) Working Groupon Codes Strategy, 258

American Society of Mechanical Engineers (ASME) Working Groupon Motor-Operated Valve, 103

American Society of Mechanical Engineers (ASME) Working Groupon Pumps, 103

American Society of Mechanical Engineers (ASME) Working GroupPressure (WGP), Standing Committee, 131

guideline for PED, 144American Welding Society (AWS), 163AMP. See Aging management program,AMR. See Aging management review.Anhydrous ammonia service, pressure vessels, 170ANI. See Authorized Nuclear Inspector.ANII. See Authorized Nuclear Inservice Inspector.Annex Z, 147, 149, 193Annulus spacers, 164ANS. See American Nuclear Society.ANSI. See American National Standards Institute.Anticipated transients without scram (ATWS), 31, 42APA. See Administrative Procedure Act.API. See American Petroleum Institute.Appliances burning gaseous fuels, New Approach Directive, 145Approval of Type B Quantity and Fissile Material Packagings,

341–342, 344Architectural Institute of Japan, stress analysis of concrete structures,

288–289Argonne National Laboratory, 21–23, 86Arkansas Nuclear One, Unit 2 nuclear power plant, 97Arrhenius equation, 76Asada, Yasuhide, 112, 257, 276, 292Asbestos removal, 428, 431As Low As is Reasonably Achievable (ALARA), 440, 447, 450, 462,

463, 471As Low As is Reasonably Practical (ALARP) region, 385ASME. See American Society of Mechanical Engineers.Asme Code at Paks Npp, Hungary 589ASNT. See American Society for Nondestructive Testing.Asphalt enamel coatings, for pipeline systems, 409, 412–413Assemblies, in Pressure Equipment Directive, 130, 151, 153, 155Assembly, definition, 218Association Française de Ingenieurs en Appareils à Pression

(AFIAP), 255ASTM. See American Society for Testing and Materials.Atomic Energy Act of 1954 (AEA), 29, 338, 341, 343, 591, 594, 625,

627, 633, 655–659, 662–663, 665, 677 Atomic Energy Control Board, Ottawa, Canada, 188Atomic Energy of Canada Limited, 187Atomic Industrial Forum, 89Attachment weld, 13, 72, 367




ATWS. See Anticipated transients without scram.Audits, age management program/review (AMP/AMR), 34, 36–39Austenitic-ferritic stainless steels

for industrial piping, French codes, 142, 191, 553, 554for pressure equipment, French codes, 252, 253for pressure vessels, French codes, 205, 208for pressure vessels, Japanese codes, 263, 264

Austenitic stainless steelsfor calandria material, 163chloride attack susceptibility, 63for containment vessels, 348for containment vessels for radioactive materials, 345, 346for cryogenic portable tanks, 367dissimilar metal welds, 19, 72environmental fatigue effects, 21, 28fatigue crack growth rate in air environments, 5, 21fatigue crack growth rate in water environment, 4–6, 9, 21–22, 24,

28flaw evaluation, FFS code (Japan), 281–282for industrial piping, French codes, 217, 222, 226, 229, 230irradiation embrittlement of, 59mechanical properties, 63piping flaw evaluation, 116–118, 121, 124piping, fracture evaluation method, Japanese codes, 281, 283piping, intergranular stress corrosion cracking, 25piping, safety factors for evaluating flawed, 19–20piping, structural factors, 118for pressure equipment, French codes, 224, 229, 230, 234, 250for pressure equipment, PD 5500 (U.K.), 311–320, 322–330for pressure vessels, French codes, 224, 237, 253for pressure vessels, Japanese codes, 263–264, 266–268, 271primary system pressure boundary piping, repair of, 19similar metal weld overlays, 19stress corrosion cracking (SCC) analysis, 2, 63wrought, crack extension and plastic collapse, 117

Austenitic steelsallowable stresses, pressure equipment, 139, 157in pressure equipment, 157

Authorized Inspection Agencies (AIA), 101, 366CANDU® nuclear power plants, 172, 175

Authorized Inspectors (AI), 365–637certification, 254

Authorized Nuclear Inservice Inspector (ANII), 102Authorized Nuclear Inspector (ANI), 541, 627, 677Automobiles, high-pressure cylinders for on-board natural gas fuel

storage, Canadian standards, 168, 170Automotive propane vessels, Canadian standards, 168Aviation and Transportation Act (2001), 420AWS. See American Welding Society.Axial flaws (cracks), 14–15, 17–18, 49, 74, 118

applied stress intensity factor for pressure loading, 49causes of, 67in circumferential welds, 49in control rod drive mechanism (CRDM) nozzles, 76deterministic crack growth rates, 76leakage in boiling water reactor (BWR), 74nondestructive testing to determine, 72piping, safety/structural factors, 118in plate material, 15in pressurized water reactor (PWR) inlet/outlet nozzles, 74in primary water SCC in alloy 600 CRDM nozzle, 69propagation, 69

reactor pressure vessel (RPV) outlet nozzle butt weld leakage, 70

repair, 81as small leaks, 73, 96through-wall, in Alloys 82/182 butt weld, 69–70

Axial shell welds, reactor pressure vessel inservice inspection, 8Axial shrinkage, in weld repairs, 18Axial tension, of cylinders, 115

Babcock and Wilcox (B&W) designed PWR power plants, 64–66Back-wall echo criteria, 249Bar, definition, 131Barenblatt model, 113Barlow equation, 401Baseline Assessment Plan, pipeline systems, 376Base metal, welds, examination of, 52Basic Safety Standards (BSS), 290Batelle, 387, 398, 422Beam flexural tests, 186Beam lift-off tests, 186Beltline material, 16Bending moment, 4

of cylinders, 116–117of supports, PD 5500 (U.K.), 319

Bending rupture energy, in pressure equipment, 157Bending stresses, 46, 116

of containment vessels for radioactive materials, 345French codes, 191, 193, 196, 253, 653nuclear power plant piping, 296, 299nuclear pressure vessels, PD 5500 (U.K.), 324

Bending stress intensity factor, 46Bend test

French codes, 253pressure vessel, Japanese codes, 263–264

Bettis Atomic Power LaboratoryWAPD-BT-16, 85WAPD-TM-944, 85

B factor, 251BG Technology. See Advantica, British gas.Bidirectional exercise test, 105Biofouling, 33Blowoff systems, 169Blowoff vessels, 169BMI. See Bottom-mounted instrument nozzle.BNCS. See American Society of Mechanical Engineers (ASME)

Board on Nuclear Codes and Standards.BNQ. See Bureau de normalization du Québec.Boilers. See also Pressure vessels.

Canadian non-nuclear standards, 162Canadian standards, 160–163, 168failure modes, French codes, 198, 218French codes, 191, 193, 196, 253, inservice inspection, Canadian, 181–187in scope of PED, 130–131

Boiling water reactor (BWR)vs. CANDU® design, 163control rod drive stub tube cracking, 12feedwater nozzle, 8–10, 12ferritic stainless steel fatigue crack growth, 22fitness-for-service code (Japanese), 280inclusion criteria (Level A) for high-safety significant (HSS)

snubbers, 106



692 • Index

In-Core Housing, 12intergranular stress corrosion cracking (IGSCC) issue, 74, 81, 83,

85, 94, 97, 106internals, 1–6jet pump recirculation system, 1, 4–7, 13, 24licensing, 15nozzles, 11, 17, 19pressure boundary piping, service-induced degradation in, 24shroud, 24steam dryer, 14–15, 24stress corrosion cracking growth rate, 2–3, 13–14, 24–26, 28, 33,

57, 59, 63weld overlay, 1weld overlay repairs of dissimilar metal welds at nozzles, 81

Boiling water reactor (BWR) environmentaustenitic stainless steel, fatigue crack growth rate in, 21–22ferritic steels, SCC growth rate relationship, 23

Boiling water reactor/2 (BWR/2) material, 12–16Boiling water reactor/2 (BWR/2) plant, shroud support geometry, 14Boiling water reactor/3 (BWR/3) material, 15–16Boiling water reactor/4 (BWR/4) material, 15–16Boiling water reactor/5 (BWR/5) material, 15–16Boiling water reactor/6 (BWR/6) material, 15–16Boiling Water Reactor (BWR) Owners Group

analysis, 16flaw evaluation guidelines, 22objectives, 15pipe cracking in boiling water reactors, 17Topical Report, 15–16, 26, 41

NEDO-32205, Revision 1, 25–26weld-overlay studies, 18

Boiling Water Reactor (BWR) Owners Group Intergranular StressCorrosion Cracking Research Program, 17

Boiling water reactor (BWR) plant, 16design basis for, 20–21personnel radiation exposure, 53plant safety, 53use of alloy 600 base metal, 63

Boiling water reactor (BWR) vessel, 1–6, 16attachment weld cracking, 13–14hydrostatic test temperature, reference temperature, 45service-induced degradation in, 24

Boiling Water Reactor Vessels and Internals Project (BWRVIP), 1–6,24

BWRVIP-03, 25BWRVIP-5 Report (BWR Reactor Pressure Vessel Shell Weld

Inspection Recommendations), 7–8, 10–11, 25 BWRVIP-14 (Evaluation of Crack Growth in BWR Stainless Steel

Internals), 2–3, 23, 25, 28BWRVIP-17, 26BWRVIP-59, 23BWRVIP-60, 23, 26BWRVIP-99, 25BWRVIP-100, 25BWRVIP-108, 11, 26crack growth rate relationship, 23enhanced visual (VT) examinations for managing aging effect in,

60flaw evaluation, 24inspection systems, 7roll expansion repair document, 13

Bolted connectionsalloy 286 failures, 59torqued, 59

Bolted flange connectionsCODAP future specifications, 208CODETI future development, 216

Bolting, for pressure equipment, 138Bolt preload, 45Bolts, 129Boric acid

accumulation, from CRDM nozzle leakage, 69corrosion, 63, 69, 74corrosion, cross-section of Davis-Besse reactor vessel head, 70, 72,

74–75, 84deposit due to leakage, 72preexisting deposits, 71, 75wastage, 84wastage in large leaks, 71

Boronalloy presence and PWSCC, 67in primary coolant water in PWR plants, 68

Boron corrosion, from stub tube cracking leakage, 12Borosilicate glass, 436Bottom-mounted instrument (BMI) nozzles, 65

effect of temperature on PWSCC, 82examination of, 70inspections, 71partial penetrations welds, 72PWSCC leakage, 70–71strategic planning for PWSCC, 84

Boundary collocation methods, 45Bounding assumptions

for crack growth due to IASCC, 60for loss of toughness due to irradiation, 60

Bounding crack growth evaluation, 24Bounding curve, 53Bounding locations, in fatigue monitoring program, 37BPTCS. See American Society of Mechanical Engineers (ASME)Board on Pressure Technology Codes and Standards.Branch Technical Position RSB 5-2, 45, 60Brazing

code compliance, 80joints, 140repair/replacement, 96

Brazing procedures, registration, Canadian, 169Breaking pin devices, 359British Central Electricity Generating Board’s (CEGB) R-6

two-criteria failure assessment program, 119, 128British Gas (BG technology), 387, 397British R-6 method, 114, 119, 121British Standards (BS), specific types, 259

1113 (water-tubesteam generating plant), 311, 314, 3161500 (fusion-welded pressure vessels for general purposes), 309,

314, 3301501, 310–3111501-224-490A or 490B, 3101501-304-S61, 3111503, 3111515, 309, 314–315, 3301560, 3162790 (shell boiler of welded construction), 3113915 (steel vessels for primary circuits of nuclear reactors), 315

Boiling water reactor (BWR) (continued)




4504, 3164975 (prestressed concrete pressure vessels for nuclear

engineering), 3114994 (vessels and tanks in reinforced plastics), 3115169 (fusion-welded steel air receivers), 3115500 (unfired fusion-welded pressure vessels), 309, 311, 321, 324,

330–3317005 (carbon steel vessels for use in vapor compression

refrigeration systems), 311EN 286 (simple unfired pressure vessels designed to contain air or

nitrogen), 311, 314British Standards Guide, BS 7910, 121British Standards Institution (BSI), 309Brittle fracture

allowable pressure-temperature and, 55boilers, French codes, 222codes and regulations for prevention, 44–50at embrittled vessel beltline region, 55ferritic steels at lower shelf, 113French codes, 253industrial piping, French codes, 253of packaging of radioactive materials, 339of pressure equipment, 156of pressure equipment, PD 5500 (U.K.), 312of pressure vessels, 200of pressurized water reactor vessels, 43, 49–50 prevention, Japanese codes, 270temperature and, 50toughness conformance of pressure equipment, 147

Browns Ferry Unit 1, 29BS. See British Standard. BSI.See British Standards Institution.BSS. See Basic Safety Standards.Buckling, 352

containment vessels for radioactive materials, 346, 348dished ends, and PD 5500 (U.K.), 312French codes, 191, 193, 196, 253, 653interstiffener, 313–314, light stiffeners for shells, 318nuclear pressure vessels, PD 5500 (U.K.), 323overall, 313as pipeline failure mode, 374pressure vessels, 200pressure vessels, PD 5500 (U.K.), 313–321, 323–330of radioactive material packaging, 340–341

Buckling strain, theoretical, for a perfectly circular cylinder, 314–315Bugey 3 nuclear power plant, 69Bulk low specific activity materials, 347Bureau de normalization du Québec (BNQ), 160Bureau of Explosives (Association of American Railroads) permits

for radioactive materials packages, 340Burnishing, to reduce potential PWSCC, 82Bursting

boilers, French codes, 222, 253nuclear pressure vessels, PD 5500 (U.K.), 323pressure vessels, 200

Buttering, 65Butt welds, 63

Alloys 82/182, 66, 69CRDM nozzles, examination of, 75dissimilar metal, inspection requirements, 73inlet/outlet nozzle, examination of, 74

large-diameter piping, deterministic crack growth rate predictions,77–78

large diameter piping, residual stress in, 83large-diameter PWR pipe-to-nozzle, circumferential cracking, 74nozzle-to-safe end, 72outlet nozzle, 69–70piping, PWSCC in, 74PWR reactor vessel inlet/outlet, cracks/leaks in, 63, 66residual stress, 83strategic planning for PWSCC, 83weld shrinkage, 67

BWR. See Boiling water reactors.BWRVIP. See Boiling Water Reactor Vessels and Internals Project.

CAA. See Clean Air Act.Cadmium coating, French codes, 248Calculation pressure, 325Calculation temperature, 325Calandria, 163–164

assembly, CANDU® nuclear power plants, 174–175vessel, 174

Californium-252, 343, 351Call before you dig (First call) program, 416Canadian Boiler and Pressure Vessel Standards, 159–160

Category A, 182class 1 components, 175–176, 195–196class 1C components, 178class 2 components, 176class 2C components, 178class 3 components, 176class 3C components, 178class 4 components, 176development and implementation, 160Figure 48.1 (CSA Standard-Developing Process), 160–161Figure 48.2 (CANDU® Primary Heat Transport System), 164Figure 48.3 (CANDU® Primary Heat Transport System), 165Figure 48.4 (CANDU® 6 Fuel Channel Assembly), 164–165Figure 48.5 (Simplified Schematic of CANDU® Fuel Channel

Assembly), 164–165Figure 48.6 (Schematic Overview of CANDU® Online Refueling

System), 176Table 48.1 (CSA B 51 Standard: Classification of Pipe Fittings), 169Table 48.2 (CSA N285.5 and N287.7 Interfaces-Requirements for

Inspection and Testing of Containment System Components),172, 174, 176

Canadian Boiler and Pressure Vessel Standards, specific typesA series (Construction Materials), 163B series (Tolerance Specifications and Pressure Boundary

Standards), 163B51-03, Part 2, 170B51-03, Part 3, 170C series (Electrical Codes and Standards), 163G series (Structural Steel Specification), 163S Series (Construction and Structural Specifications), 163W series (Welding Specifications), 163Z series (Quality Assurance Programs), 163CAN/CSA-B51 (Boilers, Pressure Vessels, and Pressure piping),

162, 168, 172subcommittees for, 168

CAN/CSA-B51-03 (Pressure Vessel Design and Construction),159, 162, 168, 172

CAN/CSA-B52 (Mechanical Refrigeration Code), 162, 170



694 • Index

CAN/CSA-B149.1 (Natural Gas and Propane Installation Code),162, 170

CAN/CSA-B 149.2 (Propane Storage and Handling Code), 162,169–170,

CAN/CSA-B 149.5 (Installation Code for Propane Fuel Systemsand Tanks on Highway Vehicles), 162

CAN/CSA-N285(A) (Pressure-Retaining Components), 163CAN/CSA-N285(B) (Periodic Inspection), 163CAN/CSA-N285.0 (Nuclear Boiler and Pressure Vessel Design

and Construction), Classes 1, 2, 3, 4, 6, 1C, 2C, 3C, 172, 179CAN/CSA-N285.0, revision, 162CAN/CSA-N285.0-95 (General Requirements, Classification,

Registration, and Reporting), 159CAN/CSA-N285.1 (Classes 1, 2, and 3 Components), 171CAN/CSA-N285.2, 172, 174, 175, 179CAN/CSA-N285.2-99 (Mobile Online Refueling Machines

Guidelines), 169CAN/CSA-285.3 (Requirements for Containment System

Components in CANDU® Nuclear Power Plants, 171, 176CAN/CSA-N285.3-88 (CANDU® Containment Systems), 159CAN/CSA-N285.4 (Periodic Inspection of CANDU® Nuclear

Power Plant Components, 174, 181–185, 187 CAN/CSA-285.4-94 (Periodic Inspection of Primary Nuclear

Systems), 159CAN/CSA-N285.5 (Periodic Inspection of CANDU® Nuclear

Power Plant Containment Components), 174, 176, 181, 184,185, 187

CAN/CSA-N285.5-M90 (Periodic Inspection of ContainmentComponents, Metallic and Plastic), 184

CAN/CSA-N285.6 Series (Reactor Core Internals, MaterialFabrication and Testing), 159

CAN/CSA-N285.6 (Material Standards for Reactor Componentsfor CANDU® Nuclear Power Plants), 174, 176–177

CAN/CSA-N285.6, revision, 177CAN/CSA-285.6.1 (Pressure Tubes for Use in CANDU® Fuel

Channels), 177, 187CAN/CSA-N285.6.2 (Seamless Zirconium Alloy Tubing for

Reactivity Control Units), 177CAN/CSA-N285.6.3 (Annealed Seamless Zirconium Alloy Tubing

for Liquid-Injection System (LISS) Nozzles), 177CAN/CSA-N285.6.4 (Thin-Walled, Large-Diameter Zirconium

Alloy Tubing), 177CAN/CSA-N285.6.5 (Zirconium Alloy Wire for Fuel Channel

Spacers), 177CAN/CSA-N285.6.6 (Nondestructive Examination Criteria for

Zirconium Alloys), 176CAN/CSA-N285.6.7 (Zirconium Alloy Design Data), 176–177CAN/CSA-N285.6.8 (Martensitic Stainless Steel for Fuel Channel

End Fittings), 177CAN/CSA-N285.6.9 (Materials for Supports for Pressure-Retaining

Items), 177CAN/CSA-N285.8 (Flaw Evaluation of CANDU® Zirconium

Alloy Pressure Tubes), 159, 181, 187Annex A, 186Annex B, 186Annex C, 186Annex D, 186Annex E, 186

CAN/CSA-N286 Series (Quality Assurance ProgramRequirements), 159, 163, 177–178

CAN/CSA-N286.0, 174, 177CAN/CSA-N286.1 (Procurement Quality Assurance Program

Requirements for Nuclear Power Plants), 174, 177CAN/CSA-286.2 (Design Quality Assurance for Nuclear Power

Plants), 177CAN/CSA-N286.3 (Construction Quality Assurance for Nuclear

Power Plants), 177CAN/CSA-N286.4 (Commissioning Quality Assurance for

Nuclear Power Plants), 177CAN/CSA-N286.5 (Operations Quality Assurance for Nuclear

Power Plants), 177CAN/CSA-N286.6 (Decommissioning Quality Assurance for

Nuclear Power Plants), 177CAN/CSA-N286.7 (Quality Assurance of Critical Computer

Programs, Nuclear Power Plants), 177–178CAN/CSA-N287 (CANDU® Concrete Containment Structures and

Systems), 163, 171, 176, 178CAN/CSA-N287.1 (General Requirements), 178CAN/CSA-N287.1-M82 (CANDU® Concrete Containment

Systems), 159CAN/CSA-N287.2 (Material Requirements), 178CAN/CSA-N287.3 (Design Requirements), 178CAN/CSA-N287.4 (Construction, Fabrication, and Installation

Requirements), 178CAN/CSA-N287.5 (Examination and Testing Requirements), 178CAN/CSA-N287.6 (Preoperational Proof and Leakage Rate

Testing Requirements), 178, 186CAN/CSA-N287.7 (Inservice Examination and Testing

Requirements for Concrete Containment Structures forCANDU® Nuclear Power Plants), 178–179, 184, 185–186

Appendix (Annex) A, 186Appendix (Annex) B, 186Appendix (Annex) C, 186Appendix (Annex) D, 186

CAN/CSA-N287.7-96 (Periodic Inspection of ContainmentComponents, Concrete and Structural), 159

CAN/CSA-N288 (Environmental Radiation Protection), 163CAN/CSA-N289 (Seismic Qualification of CANDU® Nuclear

Power Plant Structures and Systems), 159, 163, 171, 179CAN/CSA-N289.1 (General Requirements for Identification and

Qualification), 179CAN/CSA-N289.2 (Ground Motion Determination), 179CAN/CSA-N289.3 (Design Procedures), 179CAN/CSA-N289.4 (Testing Procedures), 179CAN/CSA-N289.5 (Instrumentation, Inspection, and Records), 179CAN/CSA-N290 (Safety and Safety-Related Systems), 163CAN/CSA-N291 (Safety-Related Structures), 163CAN/CSA-N292 (Waste Management), 163CAN/CSA-N293 (Fire Protection), 163CAN/CSA-N294 (Decommissioning), 163CAN/CSA-Z180.1 (Compressed Breathing Air and Systems), 162,

170CAN/CSA-Z299 (Canadian Quality Control Program), 168CAN/CSA-Z305.1 (Nonflammable Medical Gas Piping Systems),

162, 170CAN/CSA-Z305.3 (Pressure Regulators, Gauges, and Flow-Metering

Devices for Medical Gases), 170CAN/CSA-Z662 (Oil and Gas Pipeline Systems), 162, 170, 403,

405–406Appendix N, 376–377

CAN/CSA-Z662-03 (Oil and Gas Pipeline Systems), 159, 170–171Appendix K, 400

Canadian Boiler and Pressure Vessel Standards, specific types (continued)




CAN3-Z305.4 (Qualification Requirements for Agencies TestingNonflammable Gas Piping Systems), 170

CAN/ULC-S603.1 (Galvanic Corrosion Protection Systems forUnderground Steel Tanks), 169, 189

RMA IP-2 (Rubber Manufacturer’s Association), 170Canadian Deuterium Uranium (CANDU®) reactor design and

licensing basis, 159CANDU® Owners Group (COG), 187CANDU® 6 reactor, 163–165, 169concrete containment structures, 178design, 163fuel channel assembly, 164–165power reactors, 171

Canadian General Standards Board (CGSB), 160testing and construction materials standards, 162–163

Canadian non-nuclear standards, 162Canadian Nuclear Safety Commission (CNSC), 159, 171Canadian nuclear standards, 163Canadian Registration Numbers (CRNs), 159, 169Canadian standards

NRCC 38726 (National Building Code of Canada), 179, 189NRCC 38727 (National Fire Code of Canada), 189

Canadian Standards Association (CSA), 160annexes, nonmandatory and mandatory, 162boiler and pressure vessel standards development, 161CSA-SDP-2.1-99, 188CSA-SDP-2.2-98, 188format and structure of standards, 161–162headquarters address, to obtain standards, 168non-nuclear boiler, pressure vessel, and piping design and

construction standards, 168–171nuclear boiler and pressure vessel design and construction

standards, 171–181nuclear boiler and pressure vessel inservice inspection standards,

181–187nuclear standards, 161–162Nuclear Strategic Steering Committee (NSSC), 163publications and updates, 162standards developing process, 160–161technical committees (TCs), 160

CSA N285B Technical Committee, 181CSA N287 Technical Committee, 180, 185

Web site and headquarters address, 160Canadian Transportation Safety Board, 372CANTEACH Web site, 166, 189Capacity certification test report, 173Carbides, 16–17

PWSCC and density of, 68Carbon, alloy presence and PWSCC, 67Carbon-manganese steels

fast breeder reactor material, 251for industrial piping, French codes, 223–224for pressure equipment, French codes, 201–202, 236–237for pressure equipment PD 5500 (U.K.), 311–312

Carbon steelsboiling water reactor piping, 16for containment vessels for radioactive materials, 346dissimilar metal welds, 63environmental fatigue effects, 21fatigue life in high-temperature reactor water, 21for industrial piping, French codes, 223–224J estimation, 114–115

piping, circumferential flaws, 118–119piping, flaw evaluation, 118, 127pressure equipment, EN 13445, 328for pressure equipment, French codes, 201–202, 236–237, 242, 250for pressure equipment, PD 5500 (U.K.), 311, 324, 329for pressure equipment, U.K. and PED codes, 311for pressure vessels, Japanese codes, 282–283, 287, 297reactor vessel head, 84for transport tanks, 359, 365weldments, piping, circumferential flaws, 118–119

Carbon steels, specific typesSA-106, Grade B, 16SA302Bmod, 16SA 302B, plates, 16SA-333, Grade 6, 16, 115SA-376 304N, 81A508, 28SA-508 class 2, 81SA-516, Grade 70, 16A533B, 28SA533B, plate, 16

Carlsbad, New Mexico pipeline incident, 371, 374Carrier, 337Cask code, Japanese codes, 289–290Cask design

for Type B radioactive materials, 348Casks, shipping, for radioactive materials, 340, 343–345, 347CASS. See Cast austenitic stainless steel.Cast austenitic stainless steel (CASS)

irradiation embrittlement in, 59thermal aging embrittlement in, 59

Castings, European standards, 236Cast stainless steel, piping, fracture evaluation (Japanese codes), 281,

284Cast steel, in pressure equipment, 157Categorization of components strategy, 59Category 0, 131, 133, 134Category I, 131, 133–136, 138, 143, 147, 153Category II, 131–138, 140, 143, 147, 152, 154Category III, 133–138, 140, 142-143, 147, 152, 154Category IV, 133–138, 140, 142–143, 145, 152, 154Cathodic protection

for pipeline system assessment, 391–394for pipeline systems, 413, 410–411, 412–414for pipeline systems, calculation of resistance values, 413–414for pipeline systems, monitoring of, 415–416

CAVS. See Crack arrest/advance verification system.CCDP. See Core damage probability.CCV. See Concrete containment vessels.CDA. See Copper Development Association.CDF. See Core damage frequency.CE. See Combustion Engineering.CEA. See Commissariat à l’Energie Atomique.CEDM. See Control element drive mechanism.CEGB. See British Central Electricity Generating Board.CEN (European Standardization Body for Mechanical Equipment),

144, 150represented in Working Group Pressure Standing Committee, 144standard, 324

Central Electricity Generating Board ReportsR/H/R6, 114, 127R/H/R6, Revision 3, 121, 127



696 • Index

CERCLA. See Comprehensive Environmental Response,Compensation, and Liability Act.

Certificate of Authorization (ASME), 169Certificate of Compliance (CoC), 350–355Certificate of Conformity, 136, 144, 156Certification, CANDU® nuclear power plants, 174Certified Individual (CI), 365–366CFER Technology, 384, 385CFR. See United States Nuclear Regulatory Commission (USNRC)

Code of Federal Regulation.CGR. See Crack growth rate.CGSB. See Canadian General Standards Board.Change-in-risk evaluation, 96Charpy energy curve, 125Charpy KV tests, pressure equipment, French codes, 235Charpy V-notch (CVN) absorbed energy, 118Charpy V-notch (CVN) impact test, 54

carbon steels, PD 5500 (U.K.), 312fracture toughness transition due to temperature, 50monitoring changes in fracture toughness, 45of pressurized water reactor vessel materials, 44surveillance data, 50–51transport tanks, 359–360

Charpy V-notch (CVN) upper shelf energy (USE), 15, 121, 124pipeline systems, 397, 400

Check valveshigh-safety significant (HSS), testing strategy for, 100low-safety significant (LSS), 100RI-IST, 100

Chemical attack, of pressure equipment, 153Chemical plants, 110, 168Chemical resistance, pressure equipment conformance, 143Chemical testing, 248Chemical Volume and Control System (CVCS), risk-informed safety

significance, 100Chinese Daya Bay 1 and 2 contract, 193–194Chinese nuclear power plants, 193–194, 293Chloride-induced stress corrosion cracking, 63–64Chromium, solution heat treatment (SHT) and, 17Chromium alloys, for use in PWR vessels, 63–64Chromium carbide, 66

boundary deposition in PWHT, 63Chromium concentration, susceptibility to PWSCC and, 66–67Chromium-molybdenum steels

fast breeder reactor material, 251for industrial piping, French codes, 222–223for pressure equipment, French codes, 201–203,

236–237, 241Chromium-molybdenum-vanadium steels

for industrial piping, French codes, 222for pressure equipment, French codes, 201–203, 236–237, 251

Chromium steels, for pressure equipment, Japanese codes, 287CI. See Certified Individual.Circumferential cracks (flaws), 4, 15, 18, 49, 69–70, 118–120

in boiling water reactor (BWR), 74crack growth predictions for, 76–77on control rod drive mechanism (CRDM) nozzles, 69flaw size for nozzle failure, 79nondestructive testing to determine, 73in plate material, 15in PWR RPV CRDM nozzles, 74–75, 77–78in PWR RPV inlet/outlet nozzles, 74

tensile strength causes, 67in top head nozzles, 72–73

Circumferential reference flaw, 48–49Circumferential welds

assumed axial flaws in, 49BWR reactor pressure vessel examination requirements, 8BWR reactor pressure vessel failure frequency, 8BWR shroud cracking, 2

Civil Aeronautic Law, 260Cladding, 52, 56, 65, 66, 153, 460

alloy 65, 66, 69alloys 82/182 crack detection, 70alloys 82/182 used for, 65corrosion-resistant, 17stainless steel, on inside of pressurized water reactor vessel top

head, 74stress corrosion cracking initiation, 25

Class 1 components, 174ASME Code requirement development, 103austenitic stainless steel piping, 26BWR intervals, 1design control provisions, 102as high-safety significant (HSS), 100–101piping, flaws and continuing service, 18piping, RI-ISI requirements, 94–97piping, Section XI inspections, 94piping, structural factors, 118reactor coolant pressure boundary structures, systems, and

components, 99transport tanks, 366–367

Class 2 componentsASME Code requirements development, 103austenitic stainless steel piping, 26high-safety significance, 101piping, RI-ISI requirements, 94–97piping, structural factors, 118reactor coolant system SSC makeup, 99transport tanks, 366–367

Class 3 componentsASME Code requirement development, 103austenitic stainless steel piping, 26high-safety significance, 101piping, RI-ISI requirements, 94–97piping, structural factors, 118reactor coolant system, SSC function of removing heat from

support system, 99transport tanks, 366–367

Class CC (concrete containment) components, 99Class MC (metal containment) components, 99Cleanliness, French codes, 253Cleanup cost, crude oil pipeline break, 371CLERP. See Conditional large early release probability.Cleavage crack, from local brittle zone, 52Clock Spring(tm) repair, 404–405Closure plugs, 174CNRM. See American Society of Mechanical Engineers (ASME)

Committee on Nuclear Risk Management (CNRM).CNSC. See Canadian Nuclear Safety Commission.Coal tar enamel coatings, for pipeline systems, 409–413Coatings, 143, 409–413

for pipeline systems, property tests, 411–412pressure vessels, PD 5500 (U.K.), 319




CoC. See Certificate of Compliance.CODAP. See Code de Construction des Appareils à Pression.Code. See American Society of Mechanical Engineers Boiler and

Pressure Vessel Code.Code allowable stress, 113Code Cases, 197

on environmental fatigue effects, 21evaluation and repair of stainless steel pipe cracking, 21French codes, 197proposed, environmentally assisted fatigue crack growth in a BWR

environment, 22risk-informed, 90

Code Cases, specific typesN-47, 193N-XXX (Alternative Acceptance Criteria and Evaluation

Procedure for Ferritic Steel Component Flaws, Upper ShelfRange), Revision 4, 124, 128

N-432, 19, 27N-463, 118, 127N-494, 119, 121, 128N-504-2, 27, 86 (Revision 3), 21N-512, 15–16, 26N-560 (Alternative Examination Requirements for Class 1,

Category B-J Welds), 94, 95, 111Method B (Risk-Informed Process), 95Table of Examining Categories for Category B-J, Class 1 Piping,

96N-560-1, Method A (Risk Ranking of Pipe Segments), 94N-560-2 (Alternative Examination Requirements for Class 1,

Category B-J Welds), 94, 106N-577 (Risk-Informed Requirements for Classes 1, 2, & 3 Piping,

Method A), 94–97, 99, 100, 106, 111Table of Risk-Informed Piping Examinations, 97

N-577-1 (Appendix I), Method A, 94–97, 106, 111N-578 (Risk-Informed Requirements for Classes 1, 2, and 3

Piping)Method B, 94, 95, 97, 111

N-578-1 (Risk-Informed Requirements for Classes 1, 2 and 3Piping)

Method B, 94, 97, 106, 111N-578-1 (Appendix 1), Method B, 97N-588, 61N-606-1, 13, 26N-629, 53–54N-631, 53–54N-638, 19, 27N-640, 51, 53, 61N-641, 50N-643, 22, 28N-648-1, 10, 26N-660 (Risk-Informed Safety Classification for Use in

Risk-Informed Repair/Replacement Activities), 98, 99, 100,108, 109, 112

N-662 (Alternative Repair/Replacement Requirements for ItemsClassified in Accordance with Risk-Informed Processes), 98–103, 113

N-662(3)(a), 101N-720, 108N-722 (Additional Examination for PWR Pressure-Retaining

Welds in Class 1 Components Fabricated with Alloys600/82/182 Materials), 71

N-722, para. 44.5.1, 72

1514, 442260, 362OMN-1, 106OMN-3 (Risk Categorization), 103–106, 112OMN-4 (Treatment of Check Valves), 103–105, 112OMN-7 (Treatment of Pumps, Alternative Requirements for Pump

Testing), 103–106, 112OMN-10 (Snubbers), 103–106, 112OMN-11 (Treatment of Motor-Operated Valves), 103–106, 112OMN-12 (Treatment of Pneumatic and Hydraulic Valves),

103–106, 112Code de construction des Appariels à Pression (CODAP), 138, 139,

190, 191, 195–198, 201–203, 207–208, 254–255, 314, 316Code de construction de générateurs de VAPeur (COVAP), 191–193,

195, 217–218, 220, 224, 226, 234, 254–255Annexes, 201, 222links with PED, 217, 229nominal design stress, 213, 221Part G, overheating risk of boilers, 217vs. Section I, 224

Code de construction de Tuyauteries Industrielles (CODETI), 139,191–193, 195–220, 254–255

annexes, 212Category A, 212, 220–222Category B, 212, 220–222Category C, 212, 220–222Category D, 212, 220–222Category Ex, 212, 216, 220–222conformity assessment procedures, 222, 233design, 208, 213, 215–223Division 1, 216Division 1, scope, 209, 210Division 2, 216Division 3, 216fabrication, 216, 229materials, 212, 223, 228Pressure Equipment Directive links, 209, 217Testing and inspection, 223–224, 231–232

CODETI. See Code de construction de Tuyauteries Industrielles.Code of Federal Regulations (CFR). See United States Nuclear

Regulatory Commission (USNRC) Code of FederalRegulations (CFR).

Code of Record of ASME Section III (Nuclear Vessels), design risk-informed safety classification, 108

Code stamped devices, on Code Stamped Transport Tanks, 359Code year, 174COG. See CANDU® Owners Group.Cold leg temperatures, 67, 83Cold springing stress, 18Cold work, and Alloy 60 susceptibility to PWSCC, 68Collapse, 403

by limit load, 113of radioactive material packaging, 340–341

Collars, 131Combination impact group assessment, 95Combined tension and bending, of cylinders, 115–116Combustion Engineering (CE)-designed PWR plant, 64–66Combustion Engineering (CE) Marking, 129–131, 133–134, 136,

138, 141, 143, 149Commercial grade classification, 98Commissariat à l’Energie Atomique (CEA), 193, 195Committee of Enquiry into the Pressure Vessel Industry (U.K.), 309



698 • Index

Competent Authority, regulation of hazardous material transportation,334

Componentcategories, 104failure, consequences of, 99failure potential, 99metal fatigue of reactor coolant pressure boundary, 21

Component designer, CANDU® nuclear power plants, 172Component risk categorization, 104Component safety categorization, 104Companion, 25, 41, 333, 347, 355, 513Composite coatings, for pipeline systems, 409–414, 416, 422Composite wrap repairs, pipeline systems, 404Compressive residual stress, 84Compressive stress(es), 67, 83

French codes, 246Computational fluid dynamics analysis, 6Computational pipeline monitoring, 416Computer programs, quality assurance, for CANDU® nuclear power

plants, 177–178CONCAWE. See Conservation of Air and Water Environment.Concrete components

load and resistance factor design (LRFD)/risk-informed methods,108–109, 112, 400

use of load and resistance factor design methods, 108use of risk-informed methods, 110

Concrete casks, 269–270, 290, 443Japanese codes, 44, 257–259

Concrete containment, tendon prestress, 31, 34Concrete containment vessels (CCV), Japanese codes, 269–270,

287–288Conditional large early release probability (CLERP), 100Condition monitoring programs, 59

effects of aging and, 56, 58–59Conductivity, crack growth rate and, 24Confidence/tolerance bound, 53Confirmatory direct assessment, pipeline systems, 376Conformity assessment modules, 131, 135–136, 310

without quality assurance, 137, 172, 177, 178, 380with quality assurance, 137, 172, 177, 178, 380

Conformity assessment procedures, 129–131, 133, 135–137sboilers, French codes, 224, 253industrial piping, French codes, 142, 191, 554manufacturer responsibility, 138–140, 144, 168

Conical shells, 312–313EN 13445 vs. PD 5500 (U.K.), 328–330

Consequence assessment, ranking, 328Conservation integrals, 328Conservation of Air and Water Environment (CONCAWE), 372Constant amplitude stress, pressure vessels, PD 5500 (U.K.), 320Construction

boilers, French codes, 222industrial piping, French codes, 142, 191, 553–554nuclear boiler and pressure vessels, Canadian standards, 168, 171,

253, 337, 655, 679Construction code, 101

alternatives, 102fracture toughness requirement, 102technical requirements of replacement, 102

Construction materials, Canadian standards, 163Construction of spent nuclear fuel storage, 268–269, 345Construction products, New Approach Directive, 145

Construction specifications, Canadian standards, 163Consejo de Seguridad Nuclear (CSN), 568Contact damage, causing pipeline incidents, 373Containment boundary, definition, 185Containment building, 165, 461, 469, 528, 638, 644Containment performance assessment, 95Containment surfaces, liner plate, 31, 34Containment system, 348, 351, 502, 524, 651Containment vessels

ferritic steel shipping containers, fracture toughness, 348, 356

for radioactive materials, 345–346, 355structure, cutting access openings, 84

Contaminated material, 339disposal of, 438, 681earth, transportation of, 348

Contamination, 355identifying sources of, 507, 523, 529, 531, 630incidents in European countries, 434, 568, 604

Continued crack growth condition, 115Continuous venting, prohibition for radioactive material packages,

336, 352–353, 355Control element drive mechanisms (CEDM)

location, 65nozzle, 64, 71

Control rod drive (CRD), 31, 63, 65, 169boiling water reactor, cracking, 8boiling water reactor, return line nozzles, 9–10stub tube design, 12

Control rod drive mechanism (CRDM), replacement, 84Control rod drive mechanism (CRDM) nozzles

alloy 600 use, 63in Babcock & Wilcox, Westinghouse-designed power

plants, 65boric acid leakage from, 69butt welds, examination of, 71circumferential cracks, 67, 69, 70, 72examination methods, 71leak, 69nickel-based alloys used, 63–64partial penetration welds, 72probabilities of leakage and failure, 79–80PWSCC in, 69severe volume leakage,70, 73–74small volume leakage, 69, 72–73top head, PWR vessel PWSCC, 63type of cracking observed, 71–73

Control rod drive mechanism (CRDM) nozzle-to-head welds,PWSCC of, 69

Coolantfor CANDU® reactor, 159–160in hypothetical accident conditions, 288, 292, 336, 338, 339, 341,

345Coordinated Research Project (CRP) of IAEA, 352Copper

content, probability of vessel failure and, 55content, upper shelf life and, 15for pressure equipment, Japanese codes, 258–259for pressure equipment, PD 5500 (U.K.), 311transition fracture toughness temperature shift, 54

Copper-64, 363Copper-67, 363




Copper alloysfor pressure equipment, Japanese codes, 258for pressure equipment, PD 5500 (U.K.), 311Copper Development Association (CDA), 162

Core damageassessing risk of, 84reducing risk, 83

Core damage frequency (CDF), 8component high-safety significance and, 100estimating, 99Level 1 PRA analysis for, 96probabilistic risk-assessment (PRA) Standard, 111ranking according to contribution to, 90

Core damage probability (CCDP), 100, 525Core flow, 6

predicted crack lengths for, 6Core meltdown

initiators, 89from loss of coolant accident, 89operator error, 89from transients, 89

Core plate, BWR reactor intervals, 1, 6Core reactivity control, 74 Core spray safe end to safe-end extension weld overlay, 19Core support attachments, 66Core weld, BWR shroud designations, 3Correction factor, 34, 37–38Corrective maintenance, 84Corrosion

age evaluation of, 33CANDU® nuclear power plant components, 173causing pipeline incidents, oil and gas pipelines, 419, 421piping failure, 96pressure vessels, EN 13445, 324pressure vessels, PD 5500 (U.K.), 319zirconium alloys, 159

Corrosion allowance, in pressure equipment, 152Corrosion control

Canadian standards, 172, 405-407, 409, 423Pipeline systems, 405–407, 409, 423

Corrosion defects, in pipeline systems, assessment, 395–400, 403Corrosion fatigue

as pipeline failure mode, 387, 507, 511–513of pressure vessels, 11

Corrosion inhibitors, 372Corrosion protection, for pressure vessels, 169Corrosion-related cracking, CANDU® nuclear power plant

components, 181Corrosion resistance

of alloy 600 in high temperature water environments, 63of pressure equipment, 151pressure equipment conformance, 142

Corrosion-resistant cladding (CRC), of stainless steel piping, 17Cost modeling software, 87Costs

certificate of compliance holders, for recordkeeping and reportingregulations, 354

of Davis Besse RPV head wastage, 84of decommissioning a nuclear facility, 439, 590, 661of NRC to monitor certificate holders and applicants, 354pipeline corrosion damage, 399pipeline system assessment methods, 386, 395-396

pipeline system breaks, 372reporting minimal changes vs. preparing license amendments, 387

COVAP. See Code de construction des générateurs de VAPeur.Covers, 131Crack. See also Flaws.

hydrogen water chemistry for, 17stress improvement remedies for, 17through-wall circumferential in pipe, 115–117

Crack arrest, 53, 69, 245Crack arrest/advance verification system (CAVS), 24Crack detection, in boiling water reactors, Japanese codes, 278Crack driving force J, 115, 122–123Crack growth, 115–116

attachment weld to vessel material, 13BWR evaluations, 6changes in pH and, 68due to cyclic loading, 33, 647due to irradiation-assisted stress corrosion cracking

(IASCC), 60due to SCC, 5, 22dynamic, 52–53environmentally-assisted, 22, 28in feedwater nozzle, 8-9fracture mechanics analysis, 24hydrogen concentration, 67–68lithium concentration, 68predicting, 84, 126, 397, 450PWSCC in alloy 600 in PWRs, 69, 79vessel-to-shroud support weld, 14welding residual stress contributing to, 77–78

Crack growth rate (CGR)in alloys 82/182 reactor vessel outlet nozzle butt welds, axial

cracks, 70in BWR jet pumps, 4BWR stainless steel intervals, 2–3in BWR water environment, 2–3effect of hydrogen on PWSCC, 80effect of lithium on, 82effect of temperature reduction on, 82effect of zinc on PWSCC, 83irradiation and, 1monitoring, 24and plant monitoring, 20prediction model, 22probability of PWSCC on alloy 600 in PWRs, 86reduction in, 23

Crackingfrom aging, 57-58causing pipeline incidents, 373detecting effects of, due to aging, 59

Crack initiation, 9, 20, 25, 33, 55, 67–68, 113, 303compared to P-T limits for normal cooldown transient, 55in feedwater nozzles, 9irradiation embrittlement and, 59and primary water stress corrosion cracking, 63, 86-87, 94, 684P-T limit and deterministic analysis of conditional vessel failure,

56rate in alloy 82/182 PWSCC in butt welds, 70residual stresses and, 67steam-dryer-support-bracket, 14

Crack length, 115, 126, 396Crack-like discontinuities, French codes, 245–247, 264, 266



700 • Index

Crack mouth opening displacement, 115Crack propagation, 4, 25, 28, 77, 509, 516

due to fatigue in low alloy and stainless steels, 69dynamic loading and, 51fatigue, 4to through-wall, 73

Crack stability, 44, 74, 122, 123Crack tip

plastic deformation, 66strain rate, 22-23

Crack tip opening displacement (CTOD), 113–114, 400Crack tip stress intensity factor, 44Crack tip temperature, 124Creep, 139, 152, 157, 159, 186, 448Creep design, pressure vessels, EN 13445, 324Creep regimes

French codes, 252Japanese codes, 275

Creep ruptureboilers, French codes, 212industrial piping, French codes, 198pressure vessels, 188

Creep rupture strength, industrial piping, French codes,213, 220Creep rupture stress, French codes, 312Creep-strain laws, French codes, 251Crevice corrosion, age evaluation, 33Crevice corrosion cracking, 17Critical flaw size, 59

feedwater nozzle, 9Criticality safety, 342, 350, 351, 353, 447Criticality Safety Index (CSI), 350, 351, 353Critical stress, ferritic steels at lower shelf, 113, 374Critical zones, of pressure equipment, 152CRC. See Corrosion-resistant cladding.CRD. See Control rod drive.CRDM. See Control rod drive mechanism.CRN. See Canadian Registration Numbers.CRP. See Coordinated Research Project.Crush test, 350, 353

for Type B radioactive materials packaging, 345–350, 447, 478,681

Cryogenic Cargo Tanks, 358Cryogenic portable tanks, 358, 361, 364, 367Cryogenic temperatures, fusion reactors, Japanese codes, 293CSA. See Canadian Standards Association.CSA Info Update, 158CSI. See Criticality Safety Index.CTOD. See Crack tip opening displacement.CUF. See Cumulative usage factor.Cumulative usage factor (CUF), 20, 34, 180, 306Curies, 342–344Current licensing basis (CLB), 32–33, 35, 39

design load, 58detection of aging effects under, 58extended operation maintenance and, 58

Cushion tanks, Canadian standards, 169–170CVCS. See Chemical Volume and Control System.CVN. See Charpy V-notch upper-shelf energy (USE).CVN. See Charpy V-notch energy.CWA. See Clean Water Act.Cyclic bending stress, 14Cyclic events, design specifications and fatigue, 20

Cyclic loading, 20crack initiation/growth, 33for nuclear reactor vessels, 65

Cyclic pressure tests, 397Cylinders, circumferential through-wall flaws, 115–116Cylindrical shells

PD 5500 vs. EN 13445, 327as pressure equipment, PD 5500 (U.K.), 299, 300

D&D. See Decontamination and decommissioning.Dampers, 169Damping constant, 295, 300Dangerous goods, definition, 357Data analysis, 93Davis Bacon Act (DBA), 202, 222, 326, 488, 494, 528Davi-Besse nuclear power plant

costs of RPV head wastage, 84cross-section through reactor vessel head, 75top head boric acid corrosion, 69–70, 73–76

Daya Bay nuclear power plant, China, 193–194, 255DBA. See Davis Bacon Act.DBA. See Design by analysis.DBE, 32DBF. See Design by formula.DC. See Direct current.DCRC. See Design and Construction Rules Committee.DCVG. See Direct current voltage gradient method.DE. See Designated equipment.Deactivation (or Transition) Plan, for decommissioning, 425Dead weight loading, 67Declaration of Conformity, 311Decommissioning, definition, 661Decommissioning of nuclear facilities, 656, 661

characterization planning, 428clearance criteria, 642decontamination, 637demobilization, 390demolition of nonradioactive structures, 479evaluation of alternatives, 681evaluation of technologies, 388facility characterization, 428license termination, 420life cycle factors, 425management team, 428, 430–431operations phase, 344operations phase, NRC license termination plan, 344–345operations phase, work scope activities, 344phase of, 344post-decommissioning, 425pre-planning, 378pre-planning regulatory requirements, 378–379public and stakeholders participation program, 357–358quality assurance requirements, 352plan, decommissioning cost planning, 590plan, decommissioning project plan (DP), 643, 661plan, identification of detailed activities, 523plan, preparation of, 531plan, project schedule, 530plan, project scope, 530planned decommissioning details, 590records, 541removal/dismantling, 567




risk analysis, 533Security program, 421site restoration, 257Table 55.1 (NRC and OSHA Regulatory Delegations), 299transition planning, 511waste management, 685worldwide liability, 655

Decommissioning Operations Contractor (DOC), 659Decommissioning Plan (DP), 590, 659Decommissioning Trust Fund (DTF), 681DECON alternative, 636DECON plan, 637Decontamination, of nuclear facilities, 590, 656, 661Decontamination and decommissioning (D&D), 590, 656, 661Deep water immersion test, radioactive material packaging, 390, 425,

590, 656, 661Defect removal, 101Defense-in-depth principles, 58Deformation, weld-induced PWSCC of alloy 600, 66Deformation plasticity-based J-integral analysis, 126Deformation plasticity failure assessment diagram (DPFAD) method,

114, 118–121, 123Defueling, as part of decommissioning of nuclear facilities, 656, 661De-gas line nozzle, 65Degradation mechanisms of piping, 96

inservice inspection, 94piping classification and active, 94piping segments, 97

Degradation predictions, 76–78DEIR. See Designated equipment inspection regulations.Delayed hydride cracking, 175, 183–184, 187Delivery pipeline, 372Delta ferrite

content, thermal aging embrittlement and, 59requirements in weld reinforcement, 18

Demands for Information, 392Demolition, of nonradioactive structures, 479Dents, in pipeline systems, assessment, 396–397Department of Public Safety, New Brunswick, Canada, 168Department of Trade and Industry (DTI) (U.K.), 309Department of Transportation Act, 338–343Design

boilers, French codes, 217, 222-224, 225Canadian oil and gas pipeline systems, 170CANDU® nuclear power plants, 172–174, 176CANDU® nuclear power plants, seismic qualification, 179criteria for the facility, 33expansion bellows, 208explicit safety factor of PWR reactor vessels, 43EN 13445, 324–329fast breeder reactors, French codes, 250French pressure equipment, 191–192, 208industrial piping, French codes, 209, 226nuclear boiler and pressure vessels, Canadian standards, 168, 171,

181pressure equipment, PD 5500 (U.K.), 310, 312–318, 320–323pressure tubes, 174pressure vessels, Japanese codes, 253–254, 257–260of pressurized water reactor vessels, 43transport tanks, 360–364tubesheet heat exchangers, 208

Design and Construction Rules Committee (DCRC), 193

Design and Construction Rules for Civil Works of PWR NuclearIsland (RCC-G), 191, 194, 255

Design and Construction Rules for Electrical Equipment of NuclearIslands (RCC-E), 191, 194, 255

Design and Construction Rules for Fire Protection (RCC-I), 194Design and Construction Rules for Fuel Assemblies of Nuclear

Power Plants (RCC-C), 191, 194, 255Design and Construction Rules for Mechanical Components of FBR

Nuclear Islands (RCC-MR), 191–196, 245, 250–253, 255appendices, 243Appendix A10, 251Appendix A11, 251Appendix A1 2, 251Appendix A1 6, 251class 1 pressure components, 251class 2 pressure components, 251

Design and Construction Rules for Mechanical Components of PWRNuclear Islands (RCC-M), 253–254

vs. American Society of Mechanical Engineers Code, 254, 264,281

ANNEX ZK (Inservice Surveillance Modification), 228Appendix S.I, 248Appendix S.II, 248Appendix S.III, 248Appendix Z IV, 243Appendix ZA (Reinforcement of Opening), 238, 242Appendix ZD, 244, 245Appendix ZE, 244, 245–246Appendix ZF, 244Appendix ZG, 244, 245Appendix ZH, 244–245Appendix ZS (Inservice Surveillance Provisions), 228, 247class 1 pressure components, design, 226, 237–247class 2 pressure components, 226, 233, 242, 246–247class 3 pressure components, 226, 233, 242, 246–247design and construction rules, 191–192, 194–197, 226, 255documents covered in A.3000, 241fabrication of parts, 247–2481974 Order, 237–238, 239, 244–246vs. PED, 248

Design and Construction Rules for System Design, French Codes(RCC-P), 194, 228

Designated equipment (DE), 261, 263Classes 1 and 2, 261

Designated equipment inspection regulations (DEIR), 259, 261, 267Design basis, analysis, 104

acceptance criteria of AMP taken directly from, 58component requirements, 105events, 33, 173for pressurized water reactors, 43

Design-by-analysis, 20, 202, 222, 246calculation of loads of shells, PD 5500 (U.K.), 320concrete casks, 289concrete casks, Japanese codes, 290EN 13445, 326–327Nuclear pressure vessels PD 5500 (U.K.), 323

Design by formula (DBF), 202, 222Design-by-rule method, 314

PD 5500 (U.K.), 312Design damping constant, 294–295Design documentation, CANDU® nuclear power plants, 172Design factor (Japanese), 267, 269, 271, 275, 287, 307



702 • Index

Design fatigue analysis, 35Design fatigue usage factor, 35Design pressure, 314, 316–317, 325Design of indian pressurized heavy water reactor components 316Design specification, 20

use for dismantling planning, Design stress, 314, 315

ferritic piping, 119PD 5500 vs. EN 13445, 327

Design stress intensityof containment vessels for radioactive materials, pressure equipment, Japanese codes, 286–287

Design temperature, 325Design tensile strength, pressure equipment, Japanese codes, 287Design tensile stress, pressure equipment, Japanese codes, 286–287Design transients, 35Design yield strength, pressure equipment, Japanese codes, 286–287Destructive examination

CANDU® nuclear power plant components, 187of commercial reactor pressure vessel welds, 52joint coefficients allowed, 139, 159

Destructive failure analysis, of alloy 82/182 butt weld leak, 69Deterministic fracture mechanics (DFM), 11

as alternative nozzle weld evaluation, 11–12predicting PWSCC on Alloy 600 in PWRs, 76–78

Deterministic insightsfor component safety categorization, 103low-safety significance classification, 100–103

Det Norske Veritas (DnV), 400DnV RP-F101 method, 400

Deuterium ingress, zirconium alloys, 159DFM. See Deterministic fracture mechanics.DIAL. See Differential Absorption LIDAR.Differential Absorption LIDAR (DIAL), 417Differential thermal expansion

allowable stresses for reactor vessel components/structures, 67clad-base metal, 51

Diffusion treatment, French codes, 248Dimensionless parameter h1, 115DIN standards, 259Direct assessment, pipeline systems, 376–377, 385, 393Direct current (DC) potential technology, 24Direct current voltage gradient (DCVG) method, for pipeline system

assessment, 395Directive (97/23/CE). See Pressure Equipment Directive.Direct use of spent pressurized water reactor fuel in CANDU®

(DUPIC), 163Discontinuities, welds, PD 5500 (U.K.), 323Discontinuity formation, CANDU® nuclear power plant components,

185Discontinuity stresses, 124Dished ends, 312, 317

EN 13445, 327–328Dissimilar metal welds (DMW)

butt weld inspection requirements for, 72examination methods, 72inspection of, 81, 83MSIP applied to PWR vessel nozzle, 83weld overlays,

Distribution pipeline, DMW. See Dissimilar metal weld.DN. See Nominal diameter.DnV. See Det Norske Veritas.

DOC. See Decommissioning Operations Contractor.Documentation

justification of solutions adopted for ESRs, 147material certification of pressure equipment, 143

DOD. See United States Department of Defense.DOE. See United States Department of Energy.DOE/OCRWM. See United States Department of Energy, Office of

Civilian Radioactive Waste Management.Donnell’s formula for cylindrical shells, 262–263DOT. See United States Department of Transportation.Double containment rule

for plutonium, 344for plutonium, proposed rule elimination (1997), 349–352for plutonium, rule elimination (1998), 349–350for plutonium vitrified high level waste, elimination (1998 final

rule), 349–350DP. See Decommissioning Plan.DPFAD. See Deformation plasticity failure assessment diagram.DTF. See Decommissioning trust fund.DTI. See Department of Trade and Industry.Dual-purpose packages, 354–355Ductile cast iron, for metal casks, Japanese codes, 289Ductile collapse, 113Ductile crack extension, 122–123, 124–125Ductile fracture, of pressurized water reactor vessels, 48Ductile overload, 113Ductile tearing, 125–126Ductility

of pressure equipment, 143, 156temperature and, 50

Dugdale elastic plastic strip yield model, 398DUPIC. See Direct use of spent pressurized water reactor fuel in

CANDU®.Dupont, E.I., Savannah River Plant, 425Dye penetrant testing, 72Dynamic/arrest fracture toughness, crack propagation and, 51Dynamic crack, 53Dynamic crush test, 353. See also Crush test.

of Type B radioactive material packages, 342, 345, 355Dynamic loads

crack propagation and, 51of transport tanks, 357–359, 365, 366, 368

Dynamic load test, bend of pipe, 297

EA. See Environmental Assessment.EAF. See Environmentally assisted fatigue evaluation.EAM. See European Approval of Materials.Earthquake(s)

piping seismic design codes, 294PRA Standards and, 110

Earthquake loads, public health risk, 14, 90EC. See European Commission.Economic and Social Council of the United Nations, 333ECP. See Electrochemical corrosion potential.Eddy current inspection/examination, CANDU® nuclear power plant

components, 177French codes, 252as surface examination, 72of wetted surface of each J-groove weld and RPV head penetration

nozzle, 72zirconium alloy components, 170–171

EDF. See Electricit È de France.




EDM. See Electrodischarge machining.EDYs. See Effective degradation years.Effective crack size, 113Effective degradation Years (EDYs), 72“Effective” fatigue life correction factor, 37Effective flaw depth, 120Effective full-power years (EFPY), 72Effective stress intensity factor, 120Efficiency diagram method, 236, 237EFPY. See Effective full-power years.EFR. See European Fast Reactor studies.EGIG. See European Gas Pipeline Incident Data Group.EJMA (Expansion Joint Manufacturers Association, Inc.) Standard,

264Elastic analysis of nuclear reactor vessels, 509Elastic compensation method, 226Elastic component of J, 115Elastic instability pressure, 314Elastic-plastic analysis, 119, 272–275, 511

of nuclear reactor vessels, 113–126of piping, 295–296

Elastic-plastic ductile tearing, 118Elastic-plastic fracture mechanics (EPFM), 113

austenitic stainless steel piping, Japanese codes, 281evaluation, 117flow diagram, 119–120techniques, 124–125

Elastic stress analysismethods for design analysis of concrete containment vessels, 287–288nuclear pressure vessels, PD 5500 (U.K.), 309

Elastic stress intensity factor for an effective crack size, 115Elastomer degradation, age evaluation, 33Elastoplastic analysis, French pressure equipment, 243, 251Electrical codes and standards, Canadian standards, 163Electrical equipment, New Approach Directive, 145Electrical equipment, nuclear power plant, environmental

qualifications, 31, 34Electric Power Research Institute (EPRI), 22

categorizing systems and components for inservice inspection (ISI)programs, 100

conditions causing high boric acid corrosion, 74, 75evaluation of draft radiation embrittlement trend equations, 54NPV economic modeling software developed, 84pipe cracking in BWRs, 17piping reliability study, 294primary water stress corrosion cracking causal testing, 68probable rate of corrosion of low-alloy steel by boric acid, 74researching effect of zinc on crack growth, 80risk-informed inservice testing, pilot program for snubbers, 100testing mechanical remedial measures for PWSCC of alloy 600

nozzles, 83weld overlay repair studies, 18white paper (Reactor Vessel Integrity Requirements for Levels A

and B conditions), 46Electric Power Research Institute (EPRI) Boric Acid Corrosion

Guidebook, 74Electric Power Research Institute (EPRI) Ductile Fracture Handbook, 48Electric Power Research Institute (EPRI)/General Electric (GE)

project, 21, 22, 24Electric Power Research Institute (EPRI) J estimation scheme, 114Electric Power Research Institute (EPRI) Materials Reliability

Program (MRP)

Reactor Internals Issue Task Group (RI-ITG), 57, 59Electric Power Research Institute (EPRI) Nondestructive (NDE)

Center, 72Electric Power Research Institute (EPRI) Piping and Fitting

Reliability Program (PFDRP), 295, 296, 298Electric Power Research Institute (EPRI) Pressurized Water Reactor

(PWR) Primary Water Chemistry Guidelines, 82Electric Power Research Institute (EPRI) Reports, 115

EPRI 1003557, 26EPRI NP-719-SR (Flaw Evaluation Procedures), 61, 127EPRI NP-1406-SR, 126EPRI NP-1931, 127EPRI NP-2431, 128EPRI NP-2671-LD, 25EPRI NP-3319, 61EPRI NP-3607 (Advances in Elastic-Plastic Fracture Analysis),

127EPRI NP-4273-SR, 26EPRI NP-4443, 27EPRI NP-4665S-Sr, 85EPRI NP-4690-SR (Flaw Evaluation in Austenitic Steel Piping),

127EPRI NP-4767, 25EPRI NP-4824M (Evaluation of Flaws in Carbon Steel Piping),

127EPRI NP-4824SP (Evaluation of Flaws in Carbon Steel Piping),

127EPRI NP-5151, 61EPRI NP-5596 (Cylinders, Elastic-Plastic Fracture Analysis of

Flaws), 127EPRI NP-6045 (Evaluation of Flaws in Ferritic Piping), 127EPRI NP-6301-D (Circumferential Throughwall Cracks), 25, 61,

127EPRI NP-6927-D, 26EPRI NP-7085-D, 27EPRI NP-7103-D, 27EPRINP-7492, 128EPRI NP-7493, 86

Electric Power Research Institute (EPRI) Technical ReportsEPRI TR-100251 (White Paper), 61EPRI TR-100852, 86EPRITR-101971, 28EPRITR-103345, 85EPRITR-103566, 86EPRITR-103696, 86EPRITR-103824, 86EPRI TR-104030, 86EPRI TR-105396 (PSA Applications Guide), 111EPRI TR-105406, 86EPRITR-105696, 25EPRITR-105873, 25EPRI TR-106589-VI (PWR Steam Generator Examination

Guidelines, Revision 4, Vol.1: Guidelines), 189EPRITR-106712, 26EPRI TR-108390-RI (Application of Master Curve Fracture

Toughness Methodology for Ferritic Steel), 61EPRI TR-108709, 26EPRITR-110356, 28EPRI TR-112657 Rev. B-A (Revised Risk-Informed ISI

Procedure), 96–97, 111, 572Electric Utility Industry Law, 258–260, 263–264, 266, 268

Ordinance 51, 268



704 • Index

Electrochemical corrosion potential (ECP), 3, 17, 23–24, 67Electrode boilers, 156Electrodes, 17, 188, 286, 395, 597, 675Electrodischarge machining (EDM), skim cutting, as remedial

measure for PWSCC, 83Electroless nickel plating, as remedial measure for PWSCC, 83Electrolytic plating, French codes, 253Electromagnetic compatibility, New Approach Directive, 145Electro mechanical nickel brush plating, as remedial measure for

PWSCC, 83Electroslag, 15–16Elevated temperatures

boilers, French codes, 237concrete containment vessels, 287conformance of pressure equipment, 148fast breeder reactors, French codes, 250Japanese fast breeder reactor, 275pressure equipment conformance, 143–144, 149

Elongation after rupture, in pressure equipment, 157, 311Embrittlement, 50–51, 57Embrittlement trend curve prediction, 43, 54–55Emergency/faulted conditions, structural factor, 118Emergency Operating Procedures (EOP), 100Emergency responders, 352EN. See Euro Norm.Enbridge Pipeline, 388End fittings, 164–165, 174, 177Energy release rate, 114Energy Resource and Development Agency (ERDA), 334Enhanced immersion test, 323Enquiry Case, 311Envelope defect, 245Environmental effects

causing PWSCC, 66–68, 77–78concrete casks, Japanese codes, 290concrete containment vessels, Japanese codes, 287on fatigue life, 39fuel-handling equipment, CANDU® nuclear power plants, 176on high-CUF components, 37to initiate PWSCC in PWR, 76Japanese codes, 276pipeline systems, 372–374of reactor coolant on components, 34–35, 37seismic design, Japanese codes, 290, 294–296, 298water and fatigue of pressure vessels, 212

Environmentally-assisted crackingfatigue crack initiation/growth, 24as pipeline failure mode, 375, 390, 409of pressurized water reactor (PWR) vessels, 44stress-corrosion cracking, 24, 67–68

Environmentally assisted fatigue (EAF) evaluation, 35, 37Environmental Standard Review Plan, 31EOP. See Emergency Operating Procedure.EPA. See United States Environmental Protection Agency.EPFM. See Elastic-plastic fracture mechanics.EPR studies (project), 23, 193, 228, 245, 247

ETC-M (Paper 2488), 253EPRG. See European Pipeline Research Group.EPRI. See Electric Power Research Institute.EPU. See Extended power uprate.Equivalency recommendation, 393Equivalent flat-bottom hole criteria, 249

Equivalent margin, review summary, 16Equivalent margin analysis, 15–16ERDA. See Energy Resource and Development Agency.Erosion

of pressure equipment, 153provision for, 102

Erosion/corrosion, fuel channel feeder pipes, 183ES&H. See Environment, safety, and health.ESRs. See Essential safety requirements.Essential safety requirements (ESRs), 129–130, 142–144, 147

compliance with, 149–156of Pressure Equipment Directive, 130–131, 133–134, 137–138

ETC-M (EPR Technical Code-Mechanical Components), 193, 228Ethylene oxide, 260 EU. See European Union.Euler’s formula for tubes, 262–263Euro Norm (EN), 255Euro Norm (EN) Standards, 192, 255

revisions and new developments, 330for steel, 149

Euro Norm (EN) Standards, specific typesEN 286, 314EN 287 (Personnel), 140, 148, 312, 330EN 287-1, 248EN 288 (Procedures), 330EN 288 Part 3 (Procedure Testing for Steels), 212, 222EN 473 (Qualification of NDE Personnel), 140, 249, 526, 570,

590, 595, 596EN 571-1 (Penetrant Testing), 249EN 583 1-2-5 (Ultrasonic Testing/Examination (UT)), 249EN 895 (Tensile Tests), 249EN 1043-1, 249EN 1092, 316EN 1418, 326EN 1591 (Piping Flanges), 208, 216, 329EN 1594 (Functional Requirements for Pipeline Incident

Statistics), 372EN 1713 (Ultrasonic Examination, Welded Joints), 249EN 1759, 316EN 2650, 248EN 10002-1, 249EN 10028 (Flat Products Made of Steels for Pressure Purposes,

French Codes), 142, 236, 310, 311, 554, 555EN 10160 (Ultrasonic Examination), 249EN 10204, 143, 554, 555EN 10204: 2004-3.1, 143EN 10204: 2004-3.2 Inspection Report, 143EN 10213, French codes, 236EN 10216, 142EN 10222 (Steel Forgings for Pressure Purposes) (French codes),

236, 310EN 10246 6-7 (Ultrasonic Examination, Tube Defects), 249EN 10307 (Ultrasonic Examination, Flat Products), 249EN 12223 (Ultrasonic Examination), 249EN 12668 1-2-3 (Ultrasonic Examination), 249EN 12952, 142, 311EN 12953, 142, 311EN 13445, 142, 208, 254, 309, 310, 314, 316, 320, 324–330, 537,

556, 558Annexes B and C, 326Annex G, 329, 330Annex J, 329




Annex K, 329Annex Z, 325

EN 13445-2, 326EN 13445-3, 208, 312, 326EN 13480 (European Harmonized Standard for Piping), 142, 209,

556EN 15614-1 (Procedure Testing for Steels), 148EN 29001, 253EN 45004, 138, 570EN 45012, 138EN ISO 3452-2 (Testing of Penetrant Materials), 249EN ISO 6506 1-2-3, 249EN ISO 6507 1-2-3 test, 249EN ISO 6508 1-2-3, 249EN ISO 9000, 136, 137, 228, 253, 538EN ISO 9001, 554, 555EN ISO 9606-4, 248EN NF 287-1, 248, 554, 555EN NF 288-1, 554, 555EN NF 288-3, 212, 222EN NF 10028-2 (Pressure Vessel Steels), 236NF EN 1591, 216NF EN 9606-4, 248

European Approval of Materials (EAM), 139, 142–144, 148, 310European Parliament and Council, 131, 563European Network for Inspection Qualification (ENIQ)519, 520, 524,

529–532, 568–570, 574, 590, 629European Commission (EC), 130, 138, 142, 144, 375, 529, 530, 531,

536, 593chairing Working Group Pressure Standing Committee, 144Guiding Principles issued by, 142Web site, 130

European Commission (EC) Declaration of Conformity, 133, 144European Commission (EC) Design Examination Certificate, 136,

137European Commission (EC) Type Examination Certificate, 136, 557European Commission (EC) unit verification, 136European consortium of gas pipeline companies (GERG), 417European Court, legal authority for implementing PED Guidelines,

144European Data Sheet, 310EU 6th Framework Programmes, 582European Fast Reactor (EFR) studies, 193European Federations, represented in Working Group Pressure

Standing Committee, 144European Gas Pipeline Incident Data Group (EGIG), 372European Pipeline Research Group (EPRG), 397European Union (EU), 129–131, 133, 138, 148, 149, 310, 545, 553

New Approach concept, 129Examination categories, specific types

B-A (Vessel Welds), 6, 96–98B-D (Full Penetration Welded Nozzles in Vessels), 10, 11, 673B-F (Pressure-Retaining Dissimilar Metal Welds in Vessel

Nozzles), 72, 96–98, 572, 573B-J (Pressure-Retaining Welds in Piping), 94–98, 573B-N-1 (Interior of Reactor Vessel), 72B-N-2 (Core Support Structures), 1B-N-3 (Removable Core Support Structures), 60B-O (Control Rod Housing Welds), 72B-P (Pressure Retaining Components), 72, 98C-F-1, 96, 97, 98C-F-2, 96, 97, 98

Executive Summary of the Reactor Safety Study, 89Expansion bellows

EN 13445 standard, 329French code design rules, 206–208

Expansion joints, 131, 216, 226, 262–264, 267Explosive atmospheres, equipment and protective systems, New

Approach Directive, 146Explosives for civil uses, New Approach Directive, 145Extended operation

existing nuclear facilities, 679age management programs (AMPs) and, 58, 59license renewal application for, 29, 57standards to evaluate programs, 41

Extended power uprate (EPU), 6External events, 91, 104

probabilistic risk assessment for treatment of, 105

FabricationCANDU® nuclear power plants, 174failure due to defects in, 102industrial piping, French codes, 228–229inspections, Canadian standards, 169pressure equipment, EN 13445, 330pressure vessels, French codes, 206–208shipping containers for radioactive materials, 347transport tanks, 358, 365

FAC. See Flow-assisted corrosion.Factor MF, 122FAD. See Failure assessment program.Failure analysis, 69, 96, 264, 265Failure Assessment Diagram Procedure, 121Failure Assessment Program (FAD), two-criteria (CEGB), 117Failure modes

pipeline systems, 374pressure vessels, PD 5500 (U.K.) (EN 13445), 326

Faraday’s Law, 23Fast breeder reactors (FBR), 59, 193

French codes,193, 226, 250–253rules on design and construction (Japanese), 275

Fast fracture, French codes, 245, 247Fatality rate, from pipeline incidents, 371–373, 421Fatigue, 33, 34, 124

crack propagation in low alloy and stainless steels, 69environmental effects, 21, 34, 39nozzles, PD 5500 (U.K.), 314as pipeline failure mode, 374in pressure equipment, 152of pressure equipment, PD 5500 (U.K.), 310, 323of supports, PD 5500 (U.K.), 319

Fatigue analysis, 34, 37of containment vessels for radioactive materials, 346, 347French codes, 208, 243–245, 251Japanese codes, 271, 275, 287–289seismic design, nuclear power plant piping, 296, 298

Fatigue crack growth, 24in boiling water reactor jet pumps, 4–5vessel-to-shroud support weld, 14

Fatigue crack growth analysisevaluation methods, 5–6reference curves: austenitic stainless steels in air environment, 21reference curves: austenitic stainless steels in water environment,

21–22



706 • Index

reference curves: ferritic stainless steels in air environment, 22reference curves: ferritic stainless steels in water environment, 22

Fatigue crack growth model, 402Fatigue crack growth rate, 5–6, 9, 281Fatigue cracking

boilers, French codes, 212CANDU® nuclear power plant components, 177fitness-for-service code evaluation of BWR (Japan), 280, 286industrial piping, French codes, 198pressure vessels, 198, 205

Fatigue crack initiation, 20–24Fatigue failure, 6, 20, 102Fatigue life, 34, 37, 141

EN 13445 standard, 330piping, Japanese codes, 295–297pipeline systems, 397pressure vessels, PD 5500 (U.K.), 310, 323reactor component, 15, 21

Fatigue life correction factor, 37Fatigue monitoring program (FMP), 35, 37–38Fatigue reduction factor, pressure vessels, PD 5500 (U.K.), 322Fatigue strain correction factors, 245Fatigue strength, of containment vessels for radioactive materials,

346–347Fatigue strength/life-property curves (S/N curves), pressure vessels,

PD 5500 (U.K.), 320–321Fatigue strength reduction factor, 245, 596Fatigue usage, 21, 34, 37, 245Fatigue usage factor

calculation of, 34–35seismic stress limiting requirements, 295, 298

FAVOR Code, pressurized thermal shock events, 56FBE. See Fusion bonded epoxy coatings.FBR. See Fast breeder reactor.Federal Aviation Act of 1958, 339Federal Aviation Administration, 339Federal Aviation Agency, 336–337Federal Aviation Regulations, 339Federal Highway Agency, corrosion costs and effects study in U.S.,

407Federal Register. See United States Nuclear Regulatory Commission

(USNRC) Federal Register(FR).Feedwater nozzle, 8–9, 11–12, 24Feedwater System (FWS), 99Ferrite content, French tests for, 249Ferritic-austenitic stainless steels, for industrial piping, French codes,

203Fiber-optic sensors, 417

50 ft.-lb. regulatory requirements, 121Fillet weld, 322Filling stations, Canadian standards, 168Film/rupture model, 23Film thickness measurement, coatings and liners of CANDU®

nuclear power plant components, 186Filters, in scope of PED, 131Final assessment/inspection, of pressure equipment, 154Final safety analysis report (FSAR), 30, 32–33, 58–59, 539Financial planning, for decommissioning, Fine-grained steels

allowable membrane stress, 144for industrial piping, French codes, 212

in pressure equipment, 157for pressure equipment, French codes, 223–224

Finite element analysis, 296, 323, 507, 510, 644, 649of boiling water reactor steam dryer failure, 6for calculating semi-elliptical crack depth, 47for calculating stresses of high-pressure vessels, 47vessels, PD 5500 (U.K.), 323for nuclear reactor vessels, 67of pipeline systems, 400–402

Finite element models (FEM), 5, 18, 35, 644Finland nuclear power plants, surveillance programs, 229Finland Olkiluoto 3 project, 193Fire conditions, 91

damage-limitation requirements, pressure equipment, 154fire protection equipment, 30, 32–33, 104probabilistic risk assessment (PRA) standard, 92public health risk, 89–90

Fired-heater pressure coils, 168, 170Fired or otherwise heated equipment, categorization, 134Fired pressure vessels, 157Firetube boilers, French codes, 217, 222, 236. See also COVAP.Fissile Classes, 336–338Fissile Class I, 338Fissile Class II, 336, 337Fissile Class III, 336, 337Fissile material, 335–338, 341,–342, 344, 347, 349–350, 352–354, 356Fissile-to-nonfissile mass ratio, 354Fitness-for-service code (Japanese Nuclear Safety), 276–281Fitness-for-service demonstrations, of pressurized water reactor

vessel internals, 59FIV. See Flow-induced vibration.Ferritic steels

allowable membrane stress, 144allowable stresses, pressure equipment, 139dynamic or crack arrest condition, 53flaws in components operating in upper shelf range, 125fracture toughness, 45, 50, 53in pressure equipment, allowable stresses, 157for pressure equipment, PD 5500 (U.K.), 311radiation embrittlement, 43shipping container fracture toughness, 339stress corrosion crack growth rate relationships, 26

Ferritic stainless steelsenvironmental fatigue crack growth rate, 22failure mechanism prediction, 116flaw evaluation, FFS code (Japan), 114for industrial piping, French codes, 191piping, default material properties and Z factor, 212piping, dissimilar metal weld overlays, 19piping, flaw evaluation, 123piping, Japanese codes, 243piping, structural factors, 118for pressure equipment, French codes, 181for pressure equipment, U.K. codes, 135for transport tanks, 366

Ferrous materialsfor industrial piping, French codes, 191for pressure equipment, French codes, 202for pressure equipment, Japanese codes, 258

Flanges, 66, 131, 286, 312, 316, 317, 319, 329, 359, 363, 405, 416bolted joints, PD 5500 (U.K.), 316, 319, 324EN 13445 standard, 324

Fatigue crack growth analysis (continued)




reactor vessel lower closure, use of alloy 600, 63–66welded to alloy 600 nozzle, 65, 72, 82, 83

Flapper wheel surface polishing, as remedial measure for PWSCC,83

Flaring test, French codes, 249Flat ends, EN 13445 standard, 329Flat plates and covers, for pressure equipment, PD 5500 (U.K.), 317Flattening test, French codes, 249Flaw (crack)

allowable flaw size, 56, 78, 116, 280axial, 14, 15, 17, 18, 49, 74, 118, 648axial, leakage from, 17axial, repair of, 81axial, through-wall in alloys 82/182 butt weld, 69, 73circumferential, 4, 15, 18, 49, 72, 74, 79, 118–120, 279, 281, 648crack growth rate monitoring, 24critical size, component failure and, 73depth, 9, 13, 47, 55, 118, 122, 123, 400distribution, 12, 80effect on integrity of PWR vessels, 66elliptical, 265, 401end-of-life size, 45evaluation methods, 24, 46, 112evaluation methods of Section XI, 46, 76fatigue crack growth, 4inspection and, 258, 263length, 3, 4, 11, 49, 451, 452as predicted failure mechanisms, 82reference, 44–49, 55, 56, 59, 262semi-elliptical surface, 44–46shape factor, 48size, 44, 51, 53, 56, 59, 72, 77, 112, 114, 115, 116, 118, 120, 124,

175, 262, 263, 448, 452surface, 10, 15, 47, 48, 50, 52, 56, 78, 120, 444through-wall, 4, 9, 114–116, 262, 450

Flaw acceptance criteria and evaluation, 10, 15, 124, 280of carbon steel piping, 16, 573Class 1 ferritic piping, 118of ferritic piping, 118, 125, 281Japanese nuclear power plant components, 269of stability/instability, 126

Flaw removal, 80Flexure forces, concrete containment vessels, 269, 270, 287–288Flow-assisted corrosion (FAC), ISI programs for, 89Flow-induced vibration (FIV), 4–6Flow-sensitization, of piping failure and, 96Flow stress, 4, 18, 113, 117, 118, 265, 281, 374, 396, 398, 399, 400,

513definition, 118ferritic piping, 119

Fluence, 3, 12, 15, 43, 45, 51, 54, 55, 57, 124, 499, 500, 643crack tip, 59, 124embrittlement and prediction-trend curves, 43, 50neutron, 45, 55, 450

Fluoride-induced stress corrosion cracking (SCC), 2–5, 13, 20, 22,23, 24, 57, 63, 64, 83, 160, 161, 278–281, 290, 373, 374, 386,387, 499, 500, 512, 526, 572

Flux welds, failure mechanism, 118FMP. See Fatigue monitoring program.Folias factor M (MT), 396, 398, 401, 402“For a Use of Nuclear Energy in 21st Century of Japan”, 257Foreign national competent authority, 342, 343

Forge welding, 365Forgings, European standards, 236Form NIS-2 (Owner’s Report for Repair/Replacement Activity), 102,

541, 543Forms, sample, Canadian standards (Annex D), 163Form X, 311FR. See United States Nuclear Regulatory Commission (USNRC)

Federal Register.Fracture, as pipeline failure mode, 374Fracture design analysis, based on J integral, 114Fracture design handbook, 114Fracture mechanics analysis, 8, 9, 55, 400, 592, 646

for crack growth due to IASCC, 60flaw tolerance evaluation, 9for loss of toughness due to irradiation, 60plant-specific, 9of pressurized water reactor vessels, 45

Fracture resistance, 113, 121, 124, 512, 596“Fracture-safe” design, 43Fracture test specimen geometries, 114Fracture toughness

determining median, 53of ferritic steel, 53, 348French codes, restrictions, 230irradiation and, 1–3, 45, 57, 59, 60in light-water reactors, 45lower bound curves, 52–54Master Curve approach, 43, 53, 581, 583monitoring changes in, 45of nuclear pressure vessel steels, 114pressure vessels, Japanese codes, 281protection against pressurized thermal shocks, 33reduction in, due to aging, 57static loading and, 51temperature dependent, 50testing, 53transition temperature, 44, 53

Fracture toughness curve index, 44Fracture toughness curves, referenced, 50–53Framatome-ANP, 677Framatome-EDF teams, 194Freeze plugging, 405French association of design, construction, and inservice inspection

rules for nuclear island components (AFCEN), 191, 193–197,246

working groups, 197French codes dealing with pressure equipment, 191

Annex FA 1 (Permissible welded joints), 205design, 193Figure 49.1 (CODAP Committee Structures), 192Figure 49.2 (Initial Pragmatic Approach for Establishing RCC-M),

193Figure 49.3 (Organization of French Nuclear Codes), 194Figure 49.4 (AFCEN Structure), 195Figure 49.5 (Structure of Subsections of RCC-M and Relations

among Sections), 196Figure 49.6 (Hazard Category Determination of a Vessel

Containing Dangerous Gas), 199Figure 49.7 (Hazard Category Determination for Piping

Containing a Dangerous Gas), 218Figure 49.8 (Installation of Expansion Joints (Extracted from

Annex C3.A3)), 226



708 • Index

Figure 49.9 (Supports, Industrial Piping), 228Figure 49.10 (Buried Piping, Industrial Piping), 228Figure 49.11 (Dimensional Tolerances for Prefabricated Spools),

229Figure 49.12 (Set-on Branches with Reinforcing Ring), 230Figure 49.13 (Hazard Category for Heated Pressure Equipment,

Overheating Risk), 234Figure 49.14 (Firetube Boiler: Tubesheet Arrangements), 236Figure 49.15 (RCC-M Section 2 Structure), 242Figure 49.16 (Class 1 Design Rules and Nonmandatory

Appendices), 244Figure 49.17 (Welding Qualification and Acceptance Criteria), 247links with PED, 198surveillance programs, 229Table 49.1 (French Pressure Equipment Codes), 192Table 49.2 (RCC-M Structure), 195Table 49.3 (Comparison of ASME Code Structure and French

Codes), 196Table 49.4 (Contents of the RCC-MR Code), 197Table 49.5 (Risk Assessment for Pressure Vessels Falling Within

the Scope of the PED), 200Table 49.6 (Risk Assessment for Pressure Vessels Beyond PED

Scope), 201Table 49.7 (Determination of the Construction Category in

CODAP), 202Table 49.8 (Nominal Design Stress in CODAP® 2000 (Excerpt)),

203Table 49.9 (Weld joint Efficiency, Pressure Vessels), 203Table 49.10 (Material Grouping), 204Table 49.11 (205), 205Table 49.12 (Design Rules), 206Table 49.13 (Tolerances on Branches), 207Table 49.14 (Permissible Joints (Exerpt)), 210Table 49.15 (Extent of Nondestructive Examination), 213Table 49.16 (Conformity Assessment Interventions), 214Table 49.17 (Conformity Assessment Procedure Selection), 217Table 49.18 (Risk Assessment for Piping Falling Within the Scope

of PED), 219Table 49.19 (Risk Assessment: Additional Criteria, Industrial

Piping, Beyond PED Scope), 220Table 49.20 (Construction Category Determination in CODETI),

220Table 49.21 (Nominal Design Stress, Industrial Piping), 221Table 49.22 (Nominal Design Stress/Piping Categories, Beyond

Material Creep Range), 221Table 49.23 (Welded Joint Coefficient, Industrial Piping), 222Table 49.24 (Material Grouping, Industrial Piping), 223Table 49.25 (Steel Grades and Maximum Permissible Thickness in

Relation to Construction Categories), 224Table 49.26 (Safety Factors), 224Table 49.27 (Components, Industrial Piping), 225Table 49.28 (Flexibility Characteristic, and Flexibility and Stress

Intensification Factors), 227Table 49.29 (Dimensional Tolerances for Prefabricated Spools),

229Table 49.30 (Nondestructive Testing, Industrial Piping), 231Table 49.31 (Conformity Assessment Procedures), 233Table 49.32 (Nominal Design Stress in COVAP), 234Table 49.33 (Boiler Acceptance, Design Stresses, and Welded Joint

Efficiency by Category), 234Table 49.34 (Material Grouping, Boilers), 235

Table 49.35 (Materials for Boilers, Elevated TemperatureConsiderations), 237

Table 49.36 (Testing, Boilers), 238Table 49.37 (Testing, Boilers), 239Table 49.38 (Nondestructive Examination, Boilers), 240Table 49.39 (Conformity Assessment Procedure, Boilers), 241Table 49.40 (RCC-M: Correspondence Among Safety Class,

RCC-M Class, and Operating Conditions), 241Table 49.41 (RCC-M Section 2 Structure), 242Table 49.42a (Reactor Vessel Steel Comparison of Chemical

Properties), 243Table 49.42b (RCC-M Mandatory Appendices), 243Table 49.43a (Reactor Vessel Steel Comparison of Impact Tests),

244Table 49.44 (Correspondence among RCC-M B 3200, B 3500, and

B 3600 Criteria), 246Table 49.45 (RCC-MR Technical Appendices), 250Table 49.46 (RSE-M Appendices), 252testing and inspection, 216

French Association of Pump Manufacturers, 246French-German ETC-M rule, 253French 1984 Order on Quality, 253French Pressure Vessel Regulation, 254French Safety Authority, 193

Fundamental Safety Rule, 197Fundamental Safety Rule RFS V.2.C, 193Safety Authority Decision, 197

French Standardization Organization (AFNOR-Association Françaisede Normalisation), 192

AFNOR/SNCT Codes, 195French Standard Series NF E 32-100, 217French transposing regulations, Decree 99-1046, 222FSAR. See Final safety analysis report.Fuel bundle design, 163Fuel channels, 174Fuel cladding, 164Fuel deposit, reduction in, 83Fuel fabrication plants, 343Fuel reprocessing, 343Full penetration welds, 52Functionality analysis, as aging management strategy, 60Fuse holder, age management program (AMP), 39Fusion bonded epoxy (FBE) coatings, for pipeline systems, 410Fussell-Vesely (FV), 106FV. See Fussell-Vesely.FWS. See Feedwater System.

Gadolinium-159, 475GALL Report. See Generic Aging Lessons Learned Report.Galvanic corrosion

age evaluation, 33as pipeline failure mode, 374pressure equipment conformance, 143

GAO. See Government Accounting Office reports.Gas, in sense of PED, 133Gas distribution systems, Canadian standards, 171Gaskets, 131Gas pipeline systems, 170Gas Research Institute (GRI), 395Gas tungsten arc welding (GTAW), 18Gas Utility Industry Law, 258Gathering system, definition, 372

French codes dealing with pressure equipment (continued)




GE. See General Electric.GEIS. See Generic Environmental Impact Statement.General corrosion, as pipeline failure mode, 407General Electric (GE), 27

Pipeline Solutions, 388General Electric (GE) Reports

GEAP-24098, 28GE-NE-523-A71-0594-A, Revision 1, 26NEDE-20364, 25NEDO-21000, 26NEDO-21821-A, 26

General Electric (GE) Service Information Letter 644, Supplement 1, 27

General Instruction (Canadian publication), 162General membrane stress, of pressure equipment, 157Generic Aging Lessons Learned (GALL) Report (USNRC), 21,

28–30, 42, 57, 61. See also United States Nuclear RegulatoryCommission (USNRC) Regulatory Guides, NUREG 1801(GALL Report).

age management programs, 33, 35–38age management review (AMR), 32, 35Appendix X.M1, 28elements, 58-59exceptions to, 33purpose, 41reactor coolant environment effects on component fatigue

life, 34, 37second GALL Report, 30system groupings, 32Vol. 2 System, 38

Generic Environmental Impact Statement (GEIS) for LicenseRenewal of Nuclear Plants, 31

Geographic information systems (GIS), database, 379Geometry (deformation, caliper or band pigs), for pipeline system

assessment, 391Geotechnical issues, causing pipeline incidents, 373GERG. See European consortium of gas pipeline companies.German KTA provisions, 193Girth weld, 17, 49GIS. See Geographic information systems. GL. See United States

Nuclear Regulatory Commission (USNRC) Generic Letter.Glass, structural factors, 125Gouges, in pipeline systems, assessment, 396–397Grain size, French codes, 249Grandfather clauses, radioactive material packaging, 340

for special form radioactive material encapsulation, 340TS-R-1 provisions, 349–353, 355

GRI. See Gas Research Institute.Grinding, for removal of surface flaws, 80Ground storage vessels, compressed natural gas, Canadian standards,

168, 170Group 1 gas, 133Group 2 gas, 133Group VII transport group, 340GTAW. See Gas tungsten arc welding.Guidance Documents, 333Guiding Principles, 142Half-pipe coils, 318Hanford and West Valley, 434Hardness testing

CANDU® nuclear power plant components, 181French codes, 253

Harmonized European Product Standard, 139Harmonized standards, 129, 137–142, 147, 150

Annex Z, 147definition, 150

Hazard analysis, 140Hazard categories

boilers, French codes, 208, 212, 217, 222–224industrial piping, French codes, 191

Hazard classes, transport tanks, 358Hazard identification, 138Hazardous material, 340

disposal of, 438Hazardous Materials Regulations of the Department of

Transportation (49CFR171–178), 339Hazardous Materials Transportation Regulations (49CFR170–190),

338Hazardous Material Transportation Act of 1990 (HMTA), 350Hazard zone radius, 384HB test, 249HCAs. See High consequence areas.Heads

reactor pressure vessel bottom head, 73, 86reactor pressure vessel top-head, boric acid corrosion, 86, 114reactor pressure vessel top-head, PWSCC cracking issue, 86, 124replacement, 84

Head vent nozzle, 65Health physics (HP), 461Heat-affected zone (HAZ)

cracked welds and, 18cracking, 1-2impact testing, 249joining procedure qualifications, 140local brittle zones, 52stress corrosion cracking initiated in cladding, 25toughness properties of circumferential welds, 49weld sensitization in., 16of welds of pressure vessels, 26

Heaters for chemical processes, 156Heat exchanger examination, 108Heat exchangers

Canadian standards, 187design, EN 13445, 324Japanese codes, 257for pressure vessel components, French codes, 329in scope of PED, 131, 147

Heat exchanger tubesheetsEN 13445 standard, 329Japanese codes, 257

Heat-sink welding (HSW), of stainless steel piping, 17Heat treated steel, 144Heat treatments

French codes, 253of pressure equipment, 138, 141of pressure vessels, Japanese codes, 257–261, 263

Heatup/cooldown limit curves, for pressurized water reactors, 43, 49–51High alloy steels, 360, 365

for pressure vessels, Japanese codes, 259High consequence areas (HCAs), of pipeline rupture, 375High consequence assessment, 100High cycle fatigue, 6, 245High-fatigue lines, limiting welds, 38High-level requirements (HLR), PRA Standards, 93



710 • Index

High-level waste (HLW) borosilicate glass, 349, 356High-level waste (HLW) containing plutonium, 349, 355–356High Level Waste (HLW) Packages, 355High-pressure cylinders, for on-board storage of natural gas fuel for

automobiles, Canadian standards, 168High-pressure gases, definition, Japanese codes, 260High-Pressure Gas Safety Law (HPGSL), 258, 260High Pressure Institute of Japan, 271High-pressure polyethylene units, reactors for, 261High-safety significant (HSS) components, 90, 96High-strength steels, for pressure equipment, PD 5500 (U.K.), 312High temperature gas-cooled reactors (HTGR), 505–506High-yield strength steels, 224HLR. See High-level requirements.HLW. See High-level waste.HMR. See United States Hazardous Materials Regulations.HMTA. See Hazardous Material Transportation Act of 1990.Holiday, pipeline systems, 395Homeland security, risk-informed methods for protection, 89Homeland Security Act of 2002, 420Hoop stress, 67, 315, 402

pipeline systems, 413Hopper diagram, 324Hot cell test, 60Hot cracks, as PWSCC initiators, 67Hot isostatic pressing unit, 261Hot-leg nozzle, weld overlay repair, 81Hot-leg pipe, 69Hot leg welds, 72Hot-water boilers, 156

New Approach Directive, 144Hot-water generators, 156Hot-water tanks, Canadian standards, 169HP. See Health physics.HPGSL. See High-Pressure Gas Safety Law. HRC test, 249HSA. See Historical Site Assessment.HSS. See High-safety significant segments.HSW. See Heat-sink welding.HTGR. See High temperature gas-cooled reactor.Human factors, 104Human reliability analysis, 93Hungary’s sole Nuclear Power Plant 589Hungarian Atomic Energy Authority (HAEA).589Hungarian regulatory rules 590–591HV 10 for welds test, 249HV test, 249HWC. See Hydrogen water chemistries. H/X. See Hydrogen to fissile

material atomic ratio.Hybrid containment vessels, Japanese codes, 288Hydraulic pressure test

EN 13445 standard, 329Japanese codes, 257of pressure vessels, PD 5500 (U.K.), 319

Hydraulic valves, inservice testing using risk-insights, 105Hydride blister formation, 183, 187Hydride cracking, 159Hydrogen

crack growth rate and concentration, 67–68pressurized water reactor primary coolant concentration, 85refrigerated, hazard class, 358use in dissimilar metal weld overlays, 19

Hydrogen cyanide, liquefied, 260Hydrogen-induced cracking, 140Hydrogen to fissile material atomic ratio (H/X), 335Hydrogen water chemistries (HWC), 3, 17Hydropneumatic tanks, Canadian standards, 169Hydrostatic pressure, PWR test limits, 45, 103, 155

CANDU® nuclear power plant components, 184in pressure equipment, 157pressure vessels, Japanese codes, 263

Hydrostatic testing, 45, See also Hydro testing.detecting PWSCC, 69pipeline systems, 376, 393transport tanks, 366

Hydro testing, 376French codes, 253pipeline systems, 376, 379, 393of pressurized water reactor vessel, 45

Hypothetical accident conditions, radioactive material incidents, 336,338, 341

Hypothetical Accident Condition test, 339nHysteresis tests, piping, Japanese codes, 296

IAEA. See International Atomic Energy Agency.IASCC. See Irradiation-assisted stress corrosion cracking.IATA. See International Air Transport Association.ICAO. See International Civil Aviation Organization.ICC. See Interstate Commerce Commission.ICDA. See Internal corrosion direct assessment.ICI. See Incore instrument nozzle.ICRP. See International Commission on Radiological Protection.ID. See Inside diameter examination.Idaho National Engineering Laboratory (INEL) Oversight Program,

506IDCOR. See Industry Degraded Core Rulemaking program.Identification, for Canadian standards, for pressure

equipment, 169IDP. See Integrated decision-making panel.IEC. See International Electrotechnical Commission.IEC. See International Energy Consultants, Inc.IEEE. See Institute of Electrical and Electronics Engineers.IGSCC. See Intergranular stress corrosion cracking.Inhibitors, for pipeline systems, 416IHSI. See Induction heating stress improvement.ILI. See In-line inspection.Immersion tests, of radioactive material packaging, 416Immersion-type electrically heated boilers, 156IMP. See Integrity management plan.Impact limiters, 345Impact resistance, pressure equipment conformance, 141–142Impact strength, pressure equipment conformance, 149 Impact testing, 144

French codes, 253pressure vessels, Canadian standards, 169pressure vessels, Japanese codes, 263–264, 266radioactive material packaging, 292, 350–356transport tanks, 365

Imperfection, RCC-MR code, 251Inaugural inspection, CANDU® nuclear power plant components, 182INCO. See International Nickel Corporation.Inconel alloys, 17, 63, 85. See also Nickel alloys, specific types;

Nickel-chromium alloys; Weld metals, specific types.Incore instrument (ICI) nozzles, 65




Incremental collapseboilers, French codes, 222industrial piping, French codes, 209pressure vessels, 200

Indirect assessment, pipeline systems, 380, 386 Indian phwr 635Industrial code in korea:korea electric power industry code (kepic)

674Individual plant examination (IPE), for severe accident

vulnerabilities, 33Individual plant examination of external events (IPEEE), 33Induction heating stress improvement (IHSI), 17

as remedial measure for IGSCC in boiling water reactors, 83Industrial hazards, Japanese codes and standards

preventing, 259Industrial piping

buried, French codes, 216, 228design, French codes, 213, 216, 221–222failure modes, 212French codes (CODETI), 197–198, 212, 218, 220materials, French codes, 216, 222nominal design stress, French codes, 213, 221risk assessment, French codes, 212in scope of PED, 147

Industrial Safety and Health Law, 267Industry Degraded Core Rulemaking (IDCOR) program, 89INEL. See Idaho National Engineering Laboratory. INGAA. See Interstate Natural Gas Association.Initiating event impact group assessment, 95Inlet/outlet nozzles

primary water stress corrosion cracking in, 79projected repair weld cracking, 81

In-line inspection (ILI), of pipeline systems, 374, 378, 387, 401Inside diameter (ID) examination, of boiling water reactors, 7Inservice inspection (ISI), 110

access problems, 7as aging management strategy, 60of boiling water reactor jet pump, 10CANDU® nuclear power plants, 171code, 108and crack growth rate monitoring, 24French codes, 253French pressure equipment, 252Frequency/coverage, 59implementation, 95Japanese codes, 257–259nuclear boiler and pressure vessels, Canadian, 181as part of age management program (AMP), 57of piping, 108plant-specific, risk-informed decision-making, 103, 106of pressure-retaining RPV shell welds, 6–7, 9of pressurized water reactor nozzles, 10primary water stress corrosion cracking detection, 79probabilistic risk assessment, 111of reactor pressure vessel axial shell welds, 8of reactor pressure vessel nozzles, 11of reactor pressure vessel-to-shroud support plate weld, 13for reactor vessel nozzles, 11risk-informed (RI-IST), 90–92, 94–96, 98–112of small bore piping, 39transport tanks, 358–359, 366, 368

Inservice Inspection Rules for Mechanical Components of PWRNuclear Islands (RSE-M), 254–255

appendices, 251–252classes 1, 6, components, 251

Inservice testing (IST), 677plant-specific risk-informed decision making, 94, 106of pumps, 108risk-informed, 90–92, 94–96, 98–112of snubbers, 105–106of valves, 105

Inspectionof boilers, French codes, 217, 222–224of boiling water reactors, 10Canadian requirements, 170effects on probability of crack growth leakage to failure, 79French codes, 253improved capability by using weld overlay repair, 81industrial piping, French codes, 191nuclear boilers and pressure vessels, Canadian, 181of nuclear reactor vessels, 86PD 5500 (U.K.), 319pressure equipment, EN 13445, 331pressure vessels, French codes, 205transport tanks, 365–366

Inspection frequencyfeedwater nozzle, 9future Section XI changes, 94

Inspection interval, 13, 96alternate inspection frequency, 11CANDU® nuclear power plants, 184–185for feedwater nozzle/sparger, 10for high susceptibility plants, 73for low susceptibility plants, 73for moderate susceptibility plants, 73and probability of leakage from a top-head nozzle, 79–80socket welds, 95

Inspection personnel radiation dose expenditure, CANDU® nuclearpower plants, 185

Inspection schedules, 97Instability criterion, 116Installer, CANDU® nuclear power plants, 172Institute of Electrical and Electronics Engineers (IEEE), 110Institute of Universality of Japan, 109Instrument nozzles, with PWSCC, 63, 68Instrument selection guide (ISG), 186Instrument Society of America, 163Integrated decision-making panel (IDP), 100Integrated plant assessment (IPA), 30, 32Integrity management plan (IMP), 380

development steps, 378elements of, 378–379

Integrity Management Program, pipeline systems, 407Intelligent (smart) pig tool, 387Interferometry, 418Intergranular corrosion, French codes, restrictions, 233Intergranular corrosion test, French codes, 249Intergranular stress corrosion cracking (IGSCC), 16

boiling water reactor issue, 10initiation and propagation, 17, 66inservice (ISI) inspection program for, 94piping, cracking conditions, 16–18piping, remedial measures, 17, 82



712 • Index

repair/replacement/mitigation of, 17stainless steel for resistance to, 4stub tube, 12weld overlay repair, 18, 81

Interim staff guidance (ISG), 40Interim storage facilities, 685 Internal components of BWR vessels, 1–2, 6Internal corrosion direct assessment (ICDA), 395 Internal events, 104

probabilistic risk assessment (PRA) standard for, 111Internal initiators, 90,Internal pressure, allowable stresses for reactor vessel

components/structures, 67Internal pressure test, radioactive materials, 338International Atomic Energy Agency (IAEA), 333–334, 338–341,

349–353, 355–35610CFR71 (1988 proposed changes), 347–348guidance documents, 42labeling system for radioactive materials, 339no double-containment requirement, 355revision cycle of two years, 355Safety Series No. 6 (‘Regulations for the Safe Transport of

Radioactive Materials’), 338, 350, 352–353, 356Safety Series No. 6 (Cross Index to Present and Proposed

Regulations), 338 Appendix II, 543 Appendix I, 180 TS-G-1.1 (Advisory Material for the Regulations for the Safe

Transport of Radioactive Material), 338, 341TS-R-1 (ST-1) standard, 349, 350–353, 355–356Appendix A, 351

International Atomic Energy Agency certificate, 342 International Civil Aviation Organization (ICAO), 353International Conference on Nuclear Engineering (8th), Proceedings,

ICONE-8, 26International Congress on Advances in Nuclear Power Plants,

Proceedings, ICAPP03, 25International Electrotechnical Commission (IEC), Canadian

participation, 160International Energy Consultants, Inc. (IEC), 349International Nickel Corporation (INCO), 63International Organization for Standardization (ISO), 162

Canadian participation, 160special form radioactive materials, 335

International Organization for Standardization (ISO) Registrar, 160, 162

International Organization for Standardization (ISO) standards, specific types

7195 (Packaging of Uranium Hexafluoride for Transport), 3519000 (Quality Assurance Rules of French Codes), 228, 9001 (Quality Control Program), 16911439: 2000 (Gas Cylinders—High-Pressure Cylinders for

On-Board Storage of Natural Gas for Automobiles), 17017020, 138/DIS 2694 (International Pressure Vessel Standard), 309, 312

International System of Units (SI), 350–351International Thermo-Nuclear Experimental Reactor (ITER) Code, 269

Committee, 269project, 269

International Trade and Industry Ministerial Ordinance 51 (MITI MO 51), 270

International Trade and Industry Ministerial Ordinance 123 (MITIMO 123), 271–272

International Trade and Industry Ministry, Notification 501, 272–273 Interstate Commerce Commission (ICC), 334–338, 340–341

Notice No. 58 in Docket No. 3666, 338Order No. 70, 339Order No. 74, 339

Interstate Commerce Commission (ICC) Regulations, 338Interstate Natural Gas Association (INGAA), 420 Iodine-131, 437Iodine-133, 478IPA. See Integrated plant assessment. IPE. See Individual plant examination. IPEEE. See Individual plant examination of external events.Iron castings, for pressure equipment, French codes, 252Irradiated fatigue curves, 174Irradiated metals, risk of piping failure and, 124Irradiated steels, 5, 44–45, 50Irradiation, 1–3

loss of toughness due to, 60personnel exposure, 13shift in nil ductility, reference temperature due to, 51

Irradiation-assisted stress corrosion cracking (IASCC, 1, 57, 59Irradiation embrittlement, 15, 45, 126as aging mechanism, 57Irradiation-enhanced stress relaxation, as aging mechanism, 59Irradiation-induced void swelling, as aging mechanism, 59Irwin plastic zone correction, 113ISG. See Instrument selection guide. ISG. See Interim staff guidance. ISI. See Inservice inspection. ISMS. See Integrated safety management system.ISO. See International Organization for Standardization.IST. See Inservice testing. ITER. See International Thermo-Nuclear Experimental Reactor.

Jacketed vessels, as pressure equipment, PD 5500 (U.K.), 318 JAERI. See Japan Atomic Energy Research Institute. James A. Fitzpatrick nuclear power plant, 97Japan Atomic Energy Research Institute (IAERI), 291Japan Electrical Association (JEA), 294

Technical Guidelines for Seismic Design of Nuclear Power Plant,294

Supplement (1984), 294–295Japan Electrical Association Code (JEAC 4205-2000), 294Japan Electrotechnical Standards and Codes Committee (JESC), 271Japanese boiler and pressure vessel codes and standards, 259

class 1 components, flaw evaluation, 246class 2 components, 247class 3 components, 247Figure 50.1 (Laws/JIS under the Mandatory Laws [Pressure Vessel

Standards]), 266Figure 50.2 (Organization of JSME Committee on Power

Generation Facilities Codes), 269Figure 50.3 (JSME Design and Construction Code Structure), 273Figure 50.4 (Plastic Analysis Results for Nuclear Power Plant

Components), 274Figure 50.5 (JSME Fitness-for-Service Code Structure), 276Figure 50.6 (Flow Chart of Rules on Inspection and Flaw

Evaluation), 277 Figure 50.7 (Flow Chart to Determine the Extent of Ultrasonic Test

in ISI for General Inspection), 278

Intergranular stress corrosion cracking (IGSCC) (continued)




Figure 50.8 (Extent of Examination and Inspection Period,Determination of), 279

Figure 50.9 (Minimum Required Cross-Sectional Area, Fitness-forService), 279

Figure 50.10 (Flow of Flaw Evaluation), 280 Figure 50.11 (Flaw Evaluation Procedure for Ferritic Vessel), 282 Figure 50.12 (Fracture Evaluation Method Selection for Pipe), 283Figure 50.13 (Relation Between Design/Construction Codes and

Welding Codes), 284Figure 50.14 (Test Apparatus Sketches [Testing of Bend in Pipe]), 297 nuclear power plant components, 268 nuclear-specific material specifications, 286 seismic design codes, 290Table 50.1 (Suggested Codes), 258 Table 50.2 (Publication of JIS on Construction of Pressure Vessel),

267 Table 50.3 (Publication of JSME Committee on Power Generation

Facility Codes), 270 Table 50.4 (Comparison Between MO 51 and JSME Code on

Power Boilers), 271 Table 50.5 (Comparison Between MO 123 and JSME Code on

Power Boilers), 272 Table 50.6 (Ultrasonic Examination in the JSME FFS Code), 278Table 50.7 (Loads Posed on Concrete Portions), 288 Table 50.8 (Classification of Major ITER Components), 293 Table 50.9 (ITER Metallic Components, Requirements and

Technical Rules), 294 Table 50.10 (Technical Guidelines for a Seismic Design of Nuclear

Power Plant Allowable Stress of Piping), 296 Table 50.11 (Philosophy for the Future Revision of the Piping

Allowable Stress Standards), 299 welding, 262–264, 266–270

Japanese Industrial Standards (JIS), 257, 259, 266 Japanese Industrial Standards, specific types

B 8265-2000 (Pressure Vessel Structure), 261–264 class 2 vessels, 267 class 3 vessels, 366

B 8265-2003, 266, 286 B 8266 (Construction of Pressure Vessels, Requirements),

267–268, 286–287 class 1 vessel, 266–267

B 8270-1993 (Pressure Vessels, Basic Standard), 266–267 class 1 vessels, 266–267 class 2 vessels, 267class 3 vessels, 366

B 8271 (Pressure Vessel Shell and End Plate), 267 B 8273 (Bolting Flange of Pressure Vessel), 267 B 8274 (Pressure Vessel Tube Plate), 267B 8275, 267B 8277-1993 (Expansion Joint for Pressure Vessels), 267 B 8278 (Horizontal Pressure Vessel with Saddle Type Support), 267B 8279 (Pressure Vessel Jacket), 267 B 8280 (Noncircular Shell Pressure Vessel), 267B 8281, 267 B 8282, 267B 8283, 267B 8284 (Pressure Vessel Head Cover Quick Closing Mechanism) , 267B 8285-1993 (Pressure Vessel Welding Procedures Qualification

Tests) , 267class 1 vessel, 267class 2 vessel, 267class 3 vessel, 366

G 3106-1999, 262, 264 G 3114-1998, 264G 3115-2000, 264G 3126-2000, 264 Z 3014 (Radiographic Testing and Classification of Steel Welds),

263–264Z 3801-1997 (Qualification Procedure for Manual Welding

Technique), 263–264Z 3805-1997 (Welding Technique of Titanium), 264 Z 3811-2000 (Welding Technique of Aluminum and Aluminum

Alloys), 264Z 3821-2001 (Welding Technique of Stainless Steel), 264 Z 3841-1997 (Semiautomatic Welding Procedure), 264

Japan Maintenance Standard, 22–23Japan Power Engineering and Inspection Cooperation (JAPEIC),

fitness-for-service code, 280Japan Society of Mechanical Engineers (JSME), 258

code and rule endorsement by government, 258Codes Committee, 258Committee on Power Generation Facilities Codes, 268–269Concrete Containment Vessel Code, 287 Design and Construction Rules, 268–270, 272–273 Guideline on the Approval of new Materials (Nuclear Materials

Code), 286Nuclear Materials Code Appendix 1, 286 Power Generation Facility Codes, 268–270, 272, 276 Rules on Concrete Casks, Canister Transfer Machines, Canister

Transport Casks for Nuclear Fuel, 289–290 Rules on Concrete Containment Vessels for Nuclear Power Plants,

287 Rules on Construction of Nuclear Power Plant Components, 286 “Rules on Design and Construction for Nuclear Power Plants”,

268, 270, 272 Rules on Design and Construction for Thermal Power Generation

Facilities, 286 Rules on Fitness-for-Service for Nuclear Power Plants (2000), 268Rules on Materials for Nuclear Use, 287 Rules on Metal Casks, 289 Rules on Nuclear Design and Construction, 269 Rules on Nuclear Power Generation Facilities, 268 Rules on Thermal Power Generation Facilities, 268, 270, 286 Rules on Transportation/Storage Packaging for Spent Nuclear

Fuel, 269Subcommittee on Fusion Power Generation Facilities, 268 Subcommittee on Fusion Reactors, 291–292 Subcommittee on Nuclear Codes, 276 Subcommittee on Nuclear Power Generation Facilities, 268, 276 Subcommittee on Thermal Power Generation Facilities (SC-TP),

268–269Subgroup on Environmental Fatigue, 275 Subgroup on Materials (SG-M), 270 Subgroup on Structures Design (SG-SD), 270 Subgroup on Welding (SG-W), 270

thermal and nuclear plant component codes, 268Welding Technical Standard, 281–282

Japan Society of Nondestructive Inspection, 267Japan Standard Association, 268JAPEIC. See Japan Power Engineering and Inspection Cooperation. J applied, evaluation procedure for calculation, 122J-controlled crack growth, 115JEA. See Japan Electrical Association.JEAC. See Japan Electrical Association Code.



714 • Index

JESC. See Japan Electric Standards Committee.Jet pump, 4

boiling water reactor internals, 10 J-groove welds, 65

PWSCC detected, 65 J integral, 15 J integral estimation method, 114 J -integral fracture resistance curve equation, 124 J -Integral Material Resistance Curve, 126 J -Integral/Tearing Modulus Curve, 123 J -Integral/Tearing Modulus Methodology, 114 J -Integral/Tearing Modulus (J-T) Procedure, 114 J -integral testing, 113 JIS. See Japanese Industrial Standards. Joining personnel, 140 Joining procedures, 148

materials for pressure equipment, 212Joining procedures qualifications, of pressure equipment, 140 Joint coefficients, 139

French codes, 253in pressure equipment, 157

Joint efficiency, pressure vessels, Japanese codes, 262Joint efficiency factor, 312Joints

for boilers, testing, French codes, 224bolted flanged, PD 5500 (U.K.), 316brazed, 140French codes, 253lattice tube-to-calandria tubesheet, 175mechanical, French codes, 248oil and gas pipeline systems, Canadian, 170permanent, 140in pressure equipment, 157pressure vessels, French codes, 205pressure vessels, Japanese codes, 263soldered, for air piping, 170welded joint coefficient, industrial piping, French codes, 222welded, PD 5500 (U.K.), 318

Joint tensile test, pressure vessels, Japanese codes, 264 JR curve, 115J-R curve, 15J-R curve Crack Driving Force Diagram Procedure, 123 J-R curve test, 122JSME. See Japan Society of Mechanical Engineers. J-T. See J-Integral/Tearing Modulus Procedure.

KI, stress intensity factor, 46KIA, reference fracture toughness curve, 53. See also KIR reference

fracture toughness curve.as lower bound, 51–53

KIC, reference fracture toughness curve, 53lower bound, 52

KIm, applied pressure stress intensity factor, 49 KIR, reference fracture toughness curve, 53

index for, 44lower bound curve and high-rate loading, 53

KIt, applied thermal stress intensity factor, 46KJC, reference static fracture toughness curve, 51 KAPA spread sheet, 400Ke coefficient, piping, Japanese codes, 300Ke factor (simplified elastic-plastic analysis method), 269 Ke’ factor, 275

Ke501 factor in Notification 274KeA0, 275Kellogg Company method, 264KHK. See Koatsu Gas Hoan Kyokai.Koatsu Gas Hoan Kyokai (KHK), 260Korean nuclear power plants, surveillance programs, 229Korean regulatory system and codes of nuclear boiler and pressure

vessels 655Korean Ulchin 9–10 project, 193

allowable release limit in a hypothetical accident, 336

Labeling, of pressure equipment, 141Labeling system, for radioactive materials packages, 340Labor cost, of decommissioning a nuclear facility, 460Lamellar iron castings, for pressure equipment, French codes, 252Lame’s equations, 312Large-diameter butt welds, 77Large early release, definition, 91Large early release frequency (LERF), 8

piping, failure probability/PRA consequence, 96PRA Standards for, 90probabilistic risk-assessment (PRA) Standard, 107

Large quantity, 335“Large quantity”, of licensed material, definition, 335 Large quantity shipments, radioactive materials, 345Laser cladding, 83Laser weld repair, 83Last-pass heat sink welding (LPHSW), 17LBB. See Leak-before-break analysis. LCM. See Life cycle

management. LDM. See Low Dispersible Material.Lead-201, allowable release limit in a hypothetical accident, 348 Lead shielding, 345 Leakage, 9

boiling water reactor stub tube cracking, 12boric acid, 69bottom-head nozzles PWSCC crack in, 70from control rod drive housing, 26from CRDM nozzle PWSCC, 87radioactive soil and water remediation, 486probabilistic analysis to determine PWSCC behavior in alloys

600/82/182, 70top head, from flange gaskets, 73

Leak-before-break (LBB) analysis, 9of CANDU® nuclear power plants, 170fast breeder reactors, French codes, 250French pressure equipment, 252fusion reactors, 292

Leak testing (LT), 45of boiling water reactor components, 45of CANDU® nuclear power plants, 185French codes, 253pressure vessels, Japanese codes, 263of pressurized water reactor components, 263risk-informed initiatives, 45as Section XI provision, 45

Leak tightness, 329of radioactive material packaging, 356

Leckie/Penny calculation, 320 Leckie/Penny formulation, 316 LEFM. See Linear-elastic fracture mechanics. LEFM/EPFM. See Linear-elastic fracture mechanics/elastic plastic

fracture mechanics analysis.




LER. See Licensee event report. LERF. See Large early release frequency. Level 1 Probabilistic Risk Assessment (PRA), 89 Level 2 Probabilistic Risk Assessment (PRA), 91 Level 3 Probabilistic Risk Assessment (PRA), 110Level of confidence, maximum, 148Levels A, B, C, D. See Service Level A, Service Level B, Service

Level C, Service Level D.LFRD. See Load and Resistance Factor Design methodology. License applications, 336 Licensee event reports (LERs), 30, 32 License-exempt contractors, 342License renewal

age management program and, 58–59aging effects, 57environmental review, 29, 31guidance documents, 29, 40–41requirements, 31TLAA identification/update, 31–32, 34, 35, 38, 41

License renewal application (LRA), 29–35, 37–38, 41, 57aging management program, 30–32, 35, 58aging management review (AMR), 30–32, 57Appendix A (Final Safety Analysis Report Supplement), 32Appendix B (Aging Management Programs and Activities), 32final safety analysis report (FSAR), supplement, 38guide for, 41requirements list, 31reviewing process, 39–40safety assessments for, 30scoping and screening methodology, 31–33Section 2.0 (Identifying Structures/Components Subject to Aging

Management Review), 31Section 3.0 (Aging Management Review Results), 32Section 4.0 (Time-Limited Aging Analyses), 32Table 3.X.1 (“Further Evaluation Recommended” and

“Discussion” Columns), 38USNRC review, 40–42

License renewal guidance (LRG) document, 40–41License termination, 428 License Termination Order, 428License Termination Plan (LTP), 428 Licensing requirements, 21

transfer of responsibilities from DOT to AEC, 341Licensing restrictions, plutonium air transport, 352LIDAR. See Light Detection and Ranging. Life cycle management (LCM) approach, 84Lifting and tiedown device requirements, radioactive materials,

336Lifting attachments, radioactive materials, 338Lifting eyes, 329–330Lifts, New Approach Directive, 147Ligament, evaluation with multiple indications, 3Ligament efficiency factor, 316Light Detection And Ranging (LIDAR), 417Light-water reactor (LWR), 104–105

construction, French codes, 253environment effects on fatigue crack growth rate, 21–22fitness-for-service code, 264monitoring changes in fracture toughness, 45piping systems, flaw evaluation, ferritic piping, 118–119use of alloy 600 base metal, 63

Light-water reactor (LWR) nuclear power plant

inservice testing (IST), 104–105Japanese codes, 257–259PRA Standard for, 90–92, 95, 97RI-IST of check valves, 105

Limit load, 3, 113, 115–119, 121, 125Limit load analysis, 113, 116, 279, 329–330

austenitic stainless steel piping, Japanese codes, 280–281 Limit load equation, 3–4Limpet coils, 318Linalog magnetic flux leakage (MFL) pigs, 387Linde 0091 flux, 54Linde 80 weld material, 121Linear elastic analysis, of containment vessels for radioactive

materials, 345–346Linear-elastic fracture mechanics (LEFM), 3, 12, 55, 113–114, 116,

118–119, 124assessing flaws effects on nuclear components, 113evaluation, 124methodology, 114predicting conditions for brittle failure, 55pressure vessels, Japanese codes, 257technique, 113

Linear elastic fracture mechanics/elastic plastic fracture mechanics(LEFM/EPFM) analysis, of irradiated stainless steel fracturetoughness, 3

Linearized stress method, 46–47Line loads, EN 13445 standard, 329–330Liners, 153Ling Ao nuclear power plant, China, 255Ling Ao 1 and 2 contract, 193Linings, 143

pressure vessels, PD 5500 (U.K.), 311Liquefied gases, 170, 260 Liquid, in sense of PED, 138Liquid-injection system (LISS) nozzles, 177Liquid natural gas systems, 159, 168

Canadian standards, 168Liquid penetrant test (PT)/examination, 400

CANDU® nuclear power plant components, 159, 163–167, 171

of feedwater nozzle/sparger, 10French codes, 196, 253pressure vessels, Japanese codes, 257, 264transport tanks, 366for vessel-to-shroud support weld cracking, 13zirconium alloy components, 176–177

Liquids rule, 376LISS. See Liquid-injection system nozzles. Lithium, 68, 82LLS. See Low level solid radioactive material. LLW. See Low level waste. LMFBR reactors, 193Load

pressure vessels, EN 13445 (PD 5500, U.K.), 309–310stress on welds, PD 5500 (U.K.), 309

Load and resistance factor design (LRFD)for concrete components, 108Level 2 analysis, 91methodology, 400reliability based, for piping, 108, 112use for nuclear service concrete components, 108with risk-informed safety classification, 108



716 • Index

Load capacity, 173of pressurized water vessels, 43replacement component, requirements and, 104

Load capacity ratings, 53Load category system, Japanese codes, 288Load line displacement, 115Load per unit thickness, 115LOCA. See Loss of coolant accident. Local brittle zones, 52Localized corrosion, piping failure, 96Log-normal distributions, in predicting initiation of PWSCC in

PWRs, 76Log-normal model, 76 Log secant model, 398 Longitudinal flaw sizes, piping, 118Longitudinal upper-shelf energy (USE), 15Longitudinal welds, 48Long-lived structures/components, 31–32, 41 Loss of alternating current power, 33Loss of coolant accident (LOCA), 80, 89, 172

determining consequence categories, 100large break, 99small-break, 99

Loss of material, from aging, 57Lost production, 84Low-alloy steel, 69

boric acid wastage, 74, 84butt welding, 65–66, 69for containment vessels for radioactive materials, 269corrosion rate by high temperature borated water onto a hot

surface, 74dissimilar metal welds, 19, 72fatigue effects, 13, 21, 69for pressure equipment, French codes, 234for pressure equipment, PD 5500 (U.K.), 311, 325for pressure vessels, Japanese codes, 263–264for reactor coolant piping, 66repair/replacement/mitigation for IGSCC in, 17SCC initiated in cladding, 24for transport tanks, 358, 366weld cracking, 13

Low-carbon stainless steel, 1 Low consequence assessment, 99Low-cycle fatigue, 245, 269, 296

Japanese codes, 257Low-cycle fatigue tests, piping, Japanese codes, 296, 298Low Dispersible Material (LDM), 350, 352Lower bound fracture toughness curves, 51Low level solid (LLS) radioactive material, 350–351 Low-level waste (LLW), 436 Low potential stress corrosion cracking (LPSCC), 64Low power, probabilistic risk-assessment standard and, 91Low-safety significance (LSS) component classification as, 90,

96–97, 104–105components, testing requirements for, 105exclusion criteria (Level B) for snubbers, 106non class, 101piping segments, 95–97, 100repair/replacement codes and, 96safety-related items having, 100

Low-specific activity (LSA) material, 340, 347LSA-I, 353

LSA-II, 353LSA-III, 353

Low temperature(s), for industrial piping, French codes, 223 Low-temperature overpressure (LTOP) protection system

pressure-temperature (P-T) limits, 49pressurized water reactor, 45, 60, 63, 85, 99setpoints, 49–50transient, 8–9, 14–15, 49, 51

Low upper shelf energy (USE) evaluation, 15, 26, 121, 128 Low upper-shelf toughness, 121, 124–126Low voltage directive, 138, 141Low water cut-off, 169LPHSW. See Last-pass heat sink welding. LPSCC. See Low potential stress corrosion cracking. LRA. See License renewal application. LRFD. See Load and resistance factor design. LRG. See License renewal guidance. LSA. See Low specific activity material. LSS. See Low-safety-significance. LT. See Leak testing.LTOP. See Low-temperature overpressure. LTP. See License Termination Plan. LWR. See Light-water reactor.

Machinery directive, 136, 141New Approach Directive, 129, 131, 137–138, 144

Machiningfor removal of surface flaws, 81of repair surface for nondestructive examination (NDE), 94

Magnesium-molybdenum-chromium-nickel steels, for industrialpiping, French codes, 223

Magnesium-molybdenum steelsfor industrial piping, French codes, 223, 224for pressure equipment, French codes, 129, 130, 131, 133–141,

Magnetic flux leakage (MFL), 387–388Magnetic particle examination, 248, 263

CANDU® nuclear power plant components, 181French codes, 253pressure vessels, Japanese codes, 263transport tanks, 366

Maintenance, 90Maintenance pigging, 390 Management Board, French codes, 191 Manganese-nickel-molybdenum steels, for pressure equipment,

French codes, 129 Manhole, sizing minimum, 169 Manufacturer. See also Certificate of Conformity.

accreditation by ASME for meeting PED requirements, 149achieving overall level of safety, 138application to Notified Body, 152assembly of pressure equipment, 131conformity assessment categories in PED, 310conformity assessment modules without QA, 136conformity assessment modules with QA, 136conformity assessment of pressure vessels, French codes, 320conformity assessment procedures for pressure equipment, 154data report detailing inspections, for Canadian regulating authority,

181defining testing type and extent, pressure vessels, French codes, 253drawing up technical documentation, 137industrial piping, testing and inspection, French codes, 142, 223inspection of boilers, French codes, 260, 253material specifications of pressure equipment, 141




New Approach principles for pressure equipment, 147non-nuclear-specific equipment, French codes, 230PED requirements, 148pressure equipment manufacturing procedures, final assessment,

marking, labeling, and operating instructions, 154product quality assurance, 137responsibility for design manufacture and conformity assessment,

pressure equipment, 144self certification, 136testing definition and extent, boilers, French codes, 224

Manufacturers Standardization Society of the Valves and FittingIndustry (MSS), 162,

MAOP. See Maximum allowable operating pressure. Marine Self Defense Force, 260 Market surveillance, 144 MARSSIM. See Multi-Agency Radioactive Site Survey and

Investigation Manual. Martensitic stainless steel

for containment vessels for radioactive materials, 346for end fitting material, 642for fuel channel end fittings, 177for industrial piping, French codes, 191for pressure equipment, French codes, 212

Master Curve reference fracture toughness, 44, 53Material certification, 143Material fatigue, primary water stress corrosion cracking, 63Material flow stress, 118. See also Stresses.Material manufacturer, of pressure equipment, 139, 153Material reference fracture toughness, 43Material Reliability Program (MRP) (EPRI sponsored), 57

MRP-86 Materials

for boilers, French codes, 224, 253for construction, PED vs. ASME code, 147industrial piping, French codes, 191, 253PD 5500 (U.K.), 311for pressure equipment, 156pressure equipment, Japanese codes, 259for pressure equipment, PD 5500 (U.K.), 264, 311for pressure vessels, French codes, 191, 201transport tanks, 365

Material specifications, of pressure equipment, 141 Material surveillance program, monitoring changes in fracture

toughness, 45Material Tables and Allowable Stress Tables (Japanese codes), 275Material transition temperature, 53MAWP. See Maximum Allowable Working Pressure. Maximum

allowable bending stress, of pressure vessels, Japanese codes, 262

Maximum allowable buckling stress, of pressure vessels, Japanese codes, 262

Maximum allowable longitudinal compressive stress, pressurevessels, Japanese codes, 262

Maximum allowable longitudinal stress, of pressure vessels, Japanese codes, 257

Maximum allowable operating pressure (MAOP), 397 Maximum allowable tensile stress, pressure vessels, Japanese codes, 263 Maximum Allowable Working Pressure (MAWP), 368

of transport tanks, 366Maximum elastic stress, nozzle reinforcement, 316 Maximum membrane stress, 139Maximum normal operating pressure (MNOP), 488

Maximum shear stress theory, 320 Maximum transport index, 351 MC. See Metal containment vessels, MDMT. See Minimum Design Metal Temperature. Mean fracture toughness curve, 53 Measured material toughness, 44 Measuring instruments, New Approach Directive, 147 Mechanical stress improvement (MSIP), 17

as remedial measure for IGSCC in boiling water reactors, 83as remedial measure for PWSCC in PWRs, 76

Mechanical testing, 283Medium consequence assessment, 99Medium voltage underground cable testing, 39Membrane stress, 46–49. See also Stresses.

circumferential reference flaw, 48, 49computing for pressure and thermal loading, 45, 122concrete containment vessels, 31nozzle reinforcement, 315of pressure equipment, PD 5500 (U.K.), 330, 331

Membrane stress intensity factor (Mm factor), 47Memorandum of Understanding (MOU) (7/2/79), 337between Interstate Commerce Commission and Atomic Energy

Commission, 338Mercury 428Metal fatigue, 9, 21, 28, 34, 35

environmental impact on nuclear power plant components, 42Metallography, crack detection by, 13Metallurgical analysis, 14Metal cracks, Japanese codes, 258, 259Metal containment (MC) vessels, Japanese codes, 288Metal structure examination, French codes, 249Methane, detection near pipeline systems, 417Methyl-bromide, liquefied, 260METI. See Ministry of Economy, Trade and Industry.Metrication Policy, 351MF factor, 124Mm factor. See Membrane stress intensity factor.MFL. See Magnetic flux leakage.MIC. See Microbial influenced corrosion.Microbial influenced corrosion (MIC), as pipeline failure mode, 408Microcleavage pop-in, 53Midland reactor pressure vessel, 61Milliroentgen per hour or equivalent, 337Miner’s Rule, 320 Mine Safety Law, 260 Minimum Design metal Temperature (MDMT), transport tanks, 360Minimum holding temperatures, Japanese codes, 263Minimum holding time, Japanese codes, 263Ministerial Council on Economic Measures (Japan), 266 Ministry of Economy, Trade and Industry (METI), 258

Notification 97Notification 408 (Technical Standards on Structure for Concrete

Reactor Containment), 287Notification 97, 147, 258, 162, 187, 260, 261, 266Ordinance 258Ordinance 270 (Technical Standards for Nuclear Power

Generation), 270Ordinance 287Ordinance 287, 288Standard Department, 258

Ministry of Education, Culture, Sports, Science and Technology,fusion reactor safety, Japanese codes, 292



718 • Index

Mitigation programs, effects of aging and, 58MITI MO 51. See International Trade and Industry Ministerial

Ordinance 270. MITI MO 123. See International Trade and Industry Ministerial

Ordinance 272.Mitigation programs, effects of aging and, 58, 59Mixed waste, classification of, 519 Mixers, 141MNOP. See Maximum normal operating pressure. Model PAT-1 package, 334Modulus of elasticity, of containment vessels for radioactive

materials, 363Molybdenum-250, 236, 351 Molybdenum stainless steel casting grades, for French pressure

equipment, 236 Molybdenum steels

for industrial piping, French codes, 191for pressure equipment, French codes, 212, 222

Moment loading, 320Monitoring, of age management program, 34 Monitoring devices, for pressure equipment, 153 Motor-operated valves (MOV), esting requirements for HSS and

LSS, 105MOU. See Memorandum of Understanding. MOV. See Motor-operated valves. MRP. See Material reliability program. MSIP. See Mechanical stress improvement process.MSS. See Manufacturers Standardization Society of the Valves and

Fitting Industry.Multi-Agency Radioactive Site Survey and Investigation Manual

(MARSSIM), 431Multiple flaw indications, 3–4. See also Flaws.

N4 studies, 245N289 Technical Committee (TC), 180Nameplate, 169, 366NASA. See national Aeronautics and Space Administration.National Accreditation Body, 138National Aeronautics and Space Administration (NASA), predicting

effect of small component failures, 89National Association of Corrosion Engineers Standards,

RP05052–2002 Item No. 21097 (Pipeline External CorrosionDirect Assessment Methodology), 394, 422

National Board Inspection Code, 366National Board Inspection Code (ANSI/NB-23), 366National Board of Boiler and Pressure Vessel Inspectors (Canada),

162National Board of Boiler and Pressure Vessel Inspectors in the United

States, 168National Board Owner/User “R” Certificate of Authorization, 366National Board “R” Stamp, 366–367National Building Code (Canada), 163, 179, 189National Energy Board (NEB) (Canada), 374, 376National Energy Board, Canada’s Safety and Performance Indicators,

annual report, 372National Energy Board Act (Canada), 420National Environmental Policy Act (NEPA), 31National Fire Code (Canada), 163, 189National Fire Protection Association (NFPA), 162National Fire Protection Association Fire Code, 163National Regulations, 129–130National Standards of Canada, 160–161

Natural gas fuel, high-pressure storage cylinders, automotive,Canadian storage, 168

Natural gas liquids, 170NB. See Notified Bodies.NCT. See Normal conditions of transport.NDE. See Nondestructive evaluation/testing.NDT. See Nil-ductility temperature.NDTT. See Nil-ductility transition temperature.NEB. See National Energy Board (Canada).NEI. See Nuclear Energy Institute.NEPA. See National Environmental Policy Act; National

Environmental Protection Act.Net positive suction head (NPSH), 45, 50Net present value (NPV) cost, 84–85Net present value (NPV) economic modeling software, 84Net-section collapse, 84

prediction, 116–117Neuber correction, 245Neutron efficiency, 164Neutron embrittlement, 31, 34, 50, 512

pressure vessels, Japanese codes, 281of PWR vessel materials, 50, 55

Neutron fluence, 59New Approach concept, 129New Approach Directives, 138, 144–145New Approach to Technical Harmonization and Standards, 129

fundamental principles, 129New/one-time inspections, detecting aging effects, 58NFPA. See National Fire Protection Association.NG18 surface flaw equation, 396Nickel

embrittlement prediction-trend curves, 50–51, 54–55for pressure equipment, Japanese codes, 286for pressure equipment, PD 5500 (U.K.), 311for pressure vessels, French codes, 203–206

Nickel 201, EAM approvals issued, 310Nickel alloys

CODAP future specifications, 208EAM approvals issued, 310for industrial piping, French codes, 222–223for pressure equipment, Japanese codes, 286, 295for pressure equipment, PD 5500 (U.K.), 311for pressure vessels, French codes, 203–206

Nickel alloys, specific typesalloy 142, 201, 287, 311, 500. See also Nickel-chromium alloys,

specific types; Weld metals, specific types.alloy 600 (NC 15 Fe), 12, 15, 19–20, 23, 28, 85, 236alloy 600, applications, 63–66alloy 600, coordinated maintenance program, 84alloy 600, inspection methods/requirements, 71–72alloy 600, locations in PWR vessel, 64–65alloy 600, primary water stress corrosion cracking of, 39, 63–64,

66, 68–69, 76–78alloy 600, properties, 63alloy 600, properties compared to austenitic stainless steel, 63alloy 600, related weld materials, 63–64alloy 600, repair processes for PWSCC, 80–82alloy 690 (NC 30 Fe), 64, 67, 84, 236alloy 690, for pressure equipment, Japanese codes, 287alloy 690, resistance to PWSCC, 67alloy 800, 64, 250SB-166, 16




Nickel-base alloysfor fast breeder reactor material, 251irradiation embrittlement, 59for pressure equipment, French codes, 242

Nickel-chromium-iron alloyschromium content and PWSCC, 66for containment vessels for radioactive materials, 346fatigue curve, 14intergranular stress corrosion cracking, 1

Nickel-chromium-iron alloys, specific typesalloy 52, 19–20alloy 82, 19–20. See also Weld metals, specific types.alloy 182, 12–15, 17, 19, 26, 28. See also Weld metals, specific types.alloy 600, 12, 15, 19–20, 23, 28, 39, 63–66, 68–69, 71–72, 76–80,

236. See also Nickel alloys, specific types.Nickel-chromium-molybdenum steels, for pressure equipment,

French codes, 242Nickel-chromium steels, for pressure equipment, French codes, 242Nickel steels, for pressure vessels, Japanese codes, 262, 286Nil ductility reference temperature, 44, 48–51, 53, 55Nil-ductility reference temperature index, 43–44Nil-ductility temperature (NDT), steel containers for radioactive

materials, 346Nil-ductility transition temperature (NDTT), 50Niobium, alloy presence and PWSCC, 67Niobium alloys, UNS R60901, for pressure tube material, 164NISA. See Nuclear and Industrial Safety Agency.Nitric acid, 436Nitrogen, addition to stainless steel for structural strength, 1NKK, 388Nobel metal, addition to mitigate cracking, 3Nominal design strength, of pressure equipment, PD 5500 (U.K.), 312Nominal design stress

boilers, French codes, 222, 234industrial piping, French codes, 212, 221

Nominal diameter (DN), 132, 134Nominal pipe size (NPS), 4, 72NON. See Notices of Nonconformance.Non-alloy quenched-tempered steels, for industrial piping, French

codes, 223Non-alloy steels

for industrial piping, French codes, 221, 223for pressure equipment, French codes, 203–204, 234–235

Nonaustenitic stainless steelsfor pressure equipment, EN 13445, 327for pressure equipment, French codes, 203–206, 224, 234

Non-austenitic steels, for industrial piping, French codes, 212, 216,223

Non-class classification, 101, 103Non-Cryogenic Portable Tanks, 358Nondestructive examination/testing (NDE), 85, 95, 101

of alloys 600/82/182 locations, 71boilers, French codes, 224, 240of BWR nozzles and their welds, 11of crack depth, 3to detect vessel flaws, 55of effects of fatigue on nuclear power plant components, 38French codes, 240, 253–254frequent, to prevent boric acid corrosion, 84future Section XI changes, 94industrial piping, French codes, 223, 231–232information in technical documentation, 137

joints, 139–140, 152, 154–155, 157personnel, 137–138mechanical components, French codes, 252of piping segments, 100personnel approval in PED, 147personnel qualification and certification, French codes, 248of pressure vessels, French codes, 205, 213pressure vessels, Japanese codes, 264–267, 281reference flaw size, 44, 56requirements, 102of RPV nozzles, 71, 72selection, 95transport tanks, 358, 364–366of weld replacement repair, 82welds, EN 13445, 326–327zirconium alloy components, CANDU® nuclear power plants, 176

Nondimensional tearing moduli, 116None consequence assessment, 99Nonferrous materials

for pressure equipment, Japanese codes, 262–263for pressure equipment, PD 5500 (U.K.), 311, 324for pressure vessels, French codes, 203–205for transport tanks, 365

Non-Linde 80, 16Non-nuclear boilers, Canadian standards, 168Non-nuclear pressure vessels, Canadian standards, 168Non safety related (NSR) classification, low-safety significance

(LSS), 100Non-stainless alloy steels, for industrial piping, French codes, 212,

214, 221Non-stainless steels

for pressure equipment, French codes, 224, 234for pressure vessels, French codes, 201–204

Notified Bodies (NB), 129–131, 133, 136–139, 146, 149appraisal of material for boilers, French codes, 212approval of design procedures of pressure equipment, 139–140,

144, 148, 152conformity assessment of boilers, French codes, 224, 241conformity assessment of pressure vessels, French codes, 208,

214–216conformity assessment procedure, industrial piping, French codes,

212experimental design approval of pressure equipment, 139–140,

142, 147, 156identification number, 136–137, 141lists, and their scope of approval, 138monitoring by, 136–137represented in Working Group Pressure Standing Committee, 144Web site providing lists and scope of approval, 138

Normal Conditions of Transport, 337–338of containment vessels for radioactive materials, 346–347

Normal form, 343of radionuclides, Type A package limits, 334

Normalized steelallowable membrane stress, 144in pressure equipment, 157for pressure vessels, French codes, 201–202

Normal operation/upset conditions (Levels A/B conditions), 14structural factors, 118

North Anna Unit 2 nuclear power plant, 74–75NOV. See Notices of Violation.Novatome, 193–194



720 • Index

Novetech, N 14–3, 61Nozzles, 68, 131, 320

alloy 600 use in PWR vessels, 65bottom-mounted instrument, 63, 65control-element drive mechanism, 65control rod drive mechanism, 63, 65cracking, 8–10, 22, 74de-gas line, 65dissimilar metal weld overlay, 19effect of temperature on PWSCC, 82ejection danger, 74, 77–78, 82examination methods, 9–12, 71–72head vent, 65improved thermal sleeve design, 11incore instrument (ICI), 65inlet/outlet, 63–64, 66, 74, 78J-groove welds for, 65liquid-injection shutdown system, 175mechanical remedial measures for PWSCC, 82–83predicting time to PWSCC, 84probabilities of leakage and failure, 79reinforcing, PD 5500 (U.K.), 314–316repair/replacement, 80–82residual stresses and crack initiation, 67stainless steel, SCC in BWRs, 64subsequent leakage following repair, 82thermocouple, 65top-head, 63, 84water-jet conditioning, 84

Nozzle to safe end socket welds, examination methods, 72Nozzle-to-safe end butt welds, surface method

examinations, 72NPS. See Nominal pipe size.NPSH. See Net positive suction head, 45NPV. See Net present value.NQA. See American Society of Mechanical Engineers (ASME)

BNCS Nuclear Quality Assurance Committee.NRC. See United States Nuclear Regulatory Commission.NRMCC. See American Society of Mechanical Engineers (ASME)

Nuclear Risk Management Coordinating Committee.NSC, 79NSNRC, installation of LTOP systems, 49NSR. See Non safety related classification.NSSC. See Canadian Standards Association, Nuclear Strategic

Steering Committee.NTD ASI Code for VVER Reactor Components, 577Nuclear and Industrial Safety Agency (NISA), 258–259Nuclear regulatory organizations, 655Nuclear boilers and pressure vessels, inservice inspection, Canadian,

181–187Nuclear cranes, 107, 109, 112–113Nuclear energy, history, 29–30Nuclear Energy Institute (NEI), 110

NEI-00–02 (Peer Review Process), 107NEI-00–02, Rev. A3 (Probabilistic Risk Assessment Peer Review

Process Guidance), 91, 111NEI-00–04 (Draft-Rev. D), 10CFR50.69 SSC Categorization

Guidelines, 98, 112NEI-95–10 (License Renewal Rule Guidance Document), 29,

31–32, 41–42Section 3.0 (Identify the SSCs Within the Scope of License

Renewal and Their Intended Function), 32

Section 4.1 (Identification of Structures and ComponentsSubject to an Aging Management Review and IntendedFunctions), 32–33

USNRC endorsement, 33Nuclear industry, risk-informed codes and standards, 107Nuclear power plants (NPPs), 433Nuclear Power Engineering Corporation (NUPEC), 295–297Nuclear power plant

aging research, NRC, 29–30detection of age effects in, 58extended operation period, 29–31license renewal, 40, 58maintenance program, 30–31onsite NRC inspectors, 30outage extension, 63plant shutdown, 63, 69, 83, 99–100, 104–106, 112–114, 165seismic design guidelines, Japanese codes, 290–300

Nuclear Power Plant Components (ASME BPV Code Section III)age management program (AMP), 33–37age management review (AMR), 33–35flaw evaluation during inservice inspection, 113–128indications evaluated from inservice inspection, 113–128metal fatigue, 34–35passive/long-lived structures/components, 32repair/replacement, 37–38time-limited aging analysis, 31

Nuclear reactor core, 164–165irradiation embrittlement and, 59

Nuclear reactorsinternal loose-parts monitoring program, 39plutonium recovery from fuel, 343risk analysis and security of, 110

Nuclear reactor vesselsbeltline welds, examination of, 72flaw evaluation during inservice inspection, 113–128indications evaluated from inservice inspection, 113–128low upper-shelf energy evaluation, 121–124time-limited aging analysis and, 31, 34

Nuclear Regulatory Commission (NRC). See United States NuclearRegulatory Commission.

“Nuclear Regulators Working Group” (NRWG-TF-NDTQ) 530, 568, 574

Nuclear Risk Management Coordinating Committee (NRMCC), 108, 110

Nuclear power plants in Spain, 567, 570NUPEC. See Nuclear Power Engineering Corporation.NWC conditions, boiling water reactor crack growth rate,

23–24NWPA. See Nuclear Waste Policy Act of 1982.NYSEARCH group of the North East Gas Association, 417

Oak Ridge National Laboratory (ORNL), 506finite element stress analyses, 48ORNL/NRC/LTR-93/15, 61ORNL/NRC/LTR-93/33, Revision 1, finite element analysis for

inside surface flaws, 61ORNL/NRC/LTR-94/26, 61

testing for microcleavage pop-ins, 53updated FAVOR code, 56

Obrigheim steam generators, U-bend cracking, 68Oconee Unit 2 nuclear power plant, 73Office of pipeline Safety (OPS), 373–376




Office of the Federal Register, 26, 42Official Journal of the European Communities, 138, 142, 146, 148, 310

97/23/EC, 130Offshore steel pipelines, Canadian standards, 171Oilfield steam distribution pipelines, Canadian standards, 171Oil pipeline systems, 170–171Oil spill, 528Oil Refineries and Petrochemical Plants, 565On-power refueling, 165Onshore Piping Regulations (OPR) (Canada), 376Operability, requirements for LTOP protection and, 50Operating condition stress/fabrication residual stress leading to

PWSCC, 67Operating heatup and cooldown limit curves, 49–51Operating instructions, 148

of pressure equipment, 141, 155Operating pressure, use in predicting crack growth rate, 77Operating temperature

effect in predicting crack growth rate, 77, 79PWSCC in pressurized water reactors, 79–80, 83

Operating timecorrections for predicting time to PWSCC, 83and probability of nozzle cracking/leakage in RPV head, 79

Operating transientsdesign fatigue analysis and, 34level A, static load conditions, 51level B, static load conditions, 51level C, static load conditions, 51level D, static load conditions, 51

Operational insights, for component safety categorization, 104OPR. See Onshore Piping Regulations.OPS. See Office of Pipeline Safety.Order on Life Cycle Asset Management, U.S. Department of Energy,

485ORNL. See Oak Ridge National Laboratory.OSHA. See Occupational Safety and Health Administration.Overlay weld metal, 17–18Overpressure protection, French codes, 249Overpressure Protection Devices, Canadian standards, 168Overpressure Protection Report, 15, 172

CANDU® nuclear power plants, 172“An Overview of R6 Revision 4”, 121, 128Owner/Licensee

CANDU® nuclear power plants, 172, 174repair program, 103

Owner’s Design Specification, fatigue, 20Oxygen service, cryogenic portable tanks, 364

PAA. See Price Anderson Indemnification Act.PACE. See Petroleum Association for the Conservation of the

Canadian Environment.Paris law crack growth model, 401Partial penetration nozzles, examination method, 72Partial penetration welds, 67

for BMI nozzles, 73for control rod drive mechanism nozzles, 73

Particular Material Appraisal (PMA), 137, 142–143, 146, 310Part wall defect, 310Passivation, 23Passive power plant structures/components, 41

aging management, 31, 57identification, 31–32

PAT. See Plutonium Air Transport package.PCCV. See Prestressed concrete containments vessels.PD 5500. See Published document (PD) 5500.PDD-63. See Presidential Decision Directive 63.Peak stress, 125–126

welds, PD 5500 (U.K.), 322–323Peak stress strength, piping, Japanese codes, 300–301Pearson method, 395PED. See Pressure Equipment Directive.Peer review process, 92, 104Pellini test, French codes, 249Penalty factors, 117–118Penetrant testing (PT), 9, 72Penetration assembly, sleeve fatigue, 31, 34Performance monitoring programs, 104

effects of aging and, 57–58Performance testing, static and dynamic, of motor-operated valves,

105Permanent joints, pressure equipment, 154Personnel

joining, qualified, 140radiation exposure, 53, 182, 352, 355–356

Petroleum Association for the Conservation of the CanadianEnvironment (PACE), 162

Petroleum gas, liquefied, 170, 260Petroleum plants, fired-heater pressure coils, 168PFM. See Probabilistic fracture mechanics.PGE. See Portland General Electric Company.pH

crack growth rate and changes in, 68of PWR primary coolant, 82

Phased array ultrasound, 388Phosphorus

alloy presence and PWSCC, 67causing hot cracks, 67

PHTS. See Primary heat transport system.PHWR. See Pressurized heavy water reactor.Physical testing, 248Physicochemical testing, 248Pigging, 386Pipe fittings, Canadian standards, 168–169Pipeline security, 371–410Pipeline systems, 371–410. See also Piping.

assessment methods, 386–394cathodic protection, 391–394, 409, 410–411, 413–415coatings, 409–414corrosion control, 405–408defect assessment methods, 395–402defect assessment models, 384, 395–402direct assessment, 376, 386, 390, 394–395emergency response plans, 421environmental protection, 372–374event tree model, 384–385environmental protection, 372–374failure modes, 373–374, 377Figure 54.1 (Gas Pipeline Explosion in Carlsbad, N.M., August

2000), 371Figure 54.2 (Natural Gas System Network), 372Figure 54.3 (Pipeline Construction [by decade]), 373Figure 54.4 (Causes of Pipeline Incidents on U.S. Pipelines in

2000), 373Figure 54.5 (Buckling, Gouging and Denting, Corrosion), 374



722 • Index

Figure 54.6 (Frequency of Occurrence of Various Threats to GasPipelines), 375

Figure 54.7 (Integrity Management Process Flow Diagrams[ASMEB31.8S]), 377

Figure 54.8 (API 1160 Managing System Integrity for HazardousLiquid Pipelines), 377

Figure 54.9 (Simplified Risk Hierarchy), 381Figure 54.10 (Example of Relative Ratings of Potential Threats),

382Figure 54.11 (Risk Matrix), 383Figure 54.12 (Risk Assessment and Mitigation Process Template),

383Figure 54.13 (Calculating the Failure Probability from a Limit

State Analysis), 384Figure 54.14 (Simple Event Tree to Predict Ignition Probability

Following Rupture), 384Figure 54.15 (Possible Scenarios Following a Gas Pipeline

Rupture), 385Figure 54.16 (ALARP Figure), 385Figure 54.17 (Effect of Three Integrity Strategies on Risk

Reduction), 386Figure 54.18 (Hydrotest Aftermath for Driving Out SCC), 386Figure 54.19 (Defect Assessment Curve), 387Figure 54.20 (Magnetic Flux Leakage), 387Figure 54.21 (Ultrasonic Tool in a Liquid Batch), 388Figure 54.22 (Four-Step Direct Assessment Process), 389Figure 54.23 (Part Wall [A] and Through Wall [B] Defects), 389Figure 54.24 ([a]Dimensions of a Longitudinal and [b]a

Circumferential Through Wall Crack Defect), 394Figure 54.25 (Dents Under Pressure), 396Figure 54.26 (Method of Determining Longitudinal Extent of

Localized Corrosion and Interaction Distances), 396Figure 54.27 (Determination of Nondimensional Variable B), 397Figure 54.28 (Simplified and Detailed RSTRENG Profiles), 398Figure 54.29 (Profile of Corrosion Depth Along the “River

Bottom” Path), 399Figure 54.30 (Remaining Strength Assessment Representation of

Metal Loss), 399Figure 54.31 (Type A and Type B Sleeves), 400Figure 54.32 (A Composite Wrap Repairs), 402Figure 54.33 (Clock Spring(tm) repair), 403Figure 54.34 (Stopple(tm) Bypass Repair Method), 404Figure 54.35 (Schematic Showing a Differential Corrosion Cell on

a Pipeline Surface), 404Figure 54.36 (Factors Affecting Corrosion), 405Figure 54.37 (Timeline of Coating Development), 407Figure 54.38 (Special Purpose Multilayer Coatings), 409Figure 54.39 (Cause of Pipeline Coating Breakdown in Australian

Pipelines), 409Figure 54.40 (Vertical Anode Arrangement), 410Figure 54.41 (Helicopter-Borne LIDAR Used for Surface

Topography and Leak Detection), 410Figure 54.42 (Buried Fiberoptic Detection Device), 415Figure 54.43 (Synthetic Aperture Radar Scanning Swaths from

Orbiting Satellites), 417Figure 54.44 (Vandalized Attack on the Alyeska Pipeline Causing

Millions of Dollars of Environmental Damage), 418Figure 54.45 (Gas Pipeline System Dependencies Source Argonne

national Laboratories), 419hydro testing, 376, 386–392inhibitors for protection, 409, 416

integrity assessment methods, 386–395integrity management plans, 375–378line marking and locating, 416liquid hydrocarbon, 372long-term repairs to pressure boundary piping, 19magnetic flux leakage for assessment, 387–389natural gas, 372pressure boundary risk, 19, 96probability of segment failure, 96ranking process, 95regulations, 374–377remote sensing of encroachment, 417–418remote sensing of leaks, 416–417repair, 402–406right of way patrols, 416risk assessment of failures, 376, 378–384risk-informed-inservice inspection (RI-ISI) process, 95risk mitigation, 384–386safety, 372–374security management programs, 418–420Table 54.2 (Fatality Rate by Mode, 2000), 373Table 54.3 (Major Threats to Transmission Pipelines ASME

B31.8S), 377Table 54.4 (Index Methods for Rating Annual Probability of

Occurrence), 382Table 54.5 (Matching Risk Severity with Level of Response), 383Table 54.6 (Defect Detection Capability of Various Inspection

Tools), 388–389Table 54.7 (Attributes of Various Pipe Protection Methods),

390–393Table 54.8 (Methods for Assessing Corrosion), 401Table 54.9 (Codes and Standards for Making Repairs, Gas

Pipelines and Oil Pipelines), 403Table 54.10 (Permissibility of Corrosion Repair Technique), 405Table 54.11 (Permissibility of Crack Repair Technique), 406Table 54.12 (Permissibility of Mechanical Damage Repair

Technique), 406Table 54.13 (Pipeline Corrosion Prevention), 408Table 54.14 (Galvanic Series of Common Commercial Metals and

Alloys in Brine [approx. 25°C]), 408Table 54.15 (Advantages and Disadvantages of Pipeline Coatings),

412Table 54.16 (Classification of Pipeline Coating Tests), 413–414third party damage awareness and control, 416–418ultrasonic testing, 388–389, 391, 394, 400

Pipeline Research Council International (PRCI), 395pipeline repair manual, for gas pipeline repairs, 403pipeline repair manual, oil pipeline repairs, 403

Pipeline Safety Improvement Act of 2002, 374, 419Pipeline transportation, advantages and purposes, 371–372Pipe rupture, circumferential cracking and, 74Pipe steel

J estimation, 115NRC/BCL 4111–1, 115

Pipetronix, 388Pipe welds

class 1, 96–97class 1, Category B-F, 96–97class 1, category B-J, 94–97class 1, category C-F-1, 96–97class 1, category C-F-2, 96–97class 2, 96–97

Pipeline systems (continued)




class 2, category B-F, 96–97class 2, category B-J, 96–97class 2, category C-F-1, 96–97class 2, category C-F-2, 96–97class 3, 96–97large-diameter PWR pipe-to-nozzle butt weld, circumferential

cracks in, 74repair/replacement/mitigation activities, 17

Piping. See also Pipeline systems.alloy 600 maintenance program, 84austenitic steel, flaw evaluation, 121austenitic steel, IGSCC, 17austenitic stainless steel, flaw evaluation, 116–118austenitic stainless steels, weld overlay repair as long-term, 19axial flaws, safety/structural factors, 18–19, 118butt welds, cracks/leaks, 63Canadian standards, 162, 166–167, 171carbon steels, flaw evaluation, 118, 127categorization, 134circumferentially flawed, stress ratio, 118crack growth rates, 22, 24degradation mechanisms, 96design, load and resistance factor design use, 108environmental fatigue effects in a BWR, 21failure rate, 96, 99ferritic stainless steel, flaw evaluation, 118–119, 127of ferritic stainless steel, structural factors, 117fracture evaluation, Japanese codes, 281–284fracture evaluation method selection, Japanese codes, 281, 284fuel channel feeder, 183hazard categories if containing a dangerous gas, 133–134high-consequence category, 100high-safety significant classification, 100inservice inspection standards, 94intergranular stress corrosion cracking with stainless steels, 1load and resistance factor design, 109longitudinal flaw sizes, allowable, 118low consequence category, 96managing internal corrosion, 59–60medium consequence category, 96nominal diameter (DN), 134none consequence category, 100pressure boundary, 19, 24, 35, 96primary water stress corrosion cracking in butt welds, 74probabilistic EPFM, 125–126probability of crack growth propagating to through-wall, 78reactor coolant, 66risk-informed applications, 109risk-informed-inservice inspection, 98–100reliability-based load and resistance factor design, 108, 111remedial measures for IGSCC in BWRs, 83remedial measures for PWSCC in PWRs, 83residual stress in large diameter butt welds, 83risk-informed classification and exam requirements, 108in scope of PED, 130–131, 133–134, 142segment degradation risk categories, 96seismic design, Japanese codes, 290–300small bore, inservice inspection, 40stainless steel, 17–18, 69, 116–117submerged arc weld, crack instability, 117through-wall circumferential crack calculation, 115–117true-stress true-strain curve, 115

ultrasonic examination in FFS code (Japan), 278–279welded, joining procedure qualifications, 140weld overlay repair, 81–82

Piping and Fitting Dynamic Reliability Program, 294Piping element tests, Japanese codes, 297Piping seismic evaluation methodology, Japanese codes, 269Pitting corrosion

age evaluation, 33as pipeline failure mode, 390–391, 395, 405–408

Plane strain, in elastic component of J, 115Plane stress, in elastic component of J, 115Plant expert panel, 95, 104–106Plant overall safety, 53Plastic collapse, 116–118, 387Plastic deformation, 113, 198

industrial piping, French codes, 212prevention, French codes, 243

Plastic instabilityboilers, French codes, 222bursting, pressure vessels, 198industrial piping, French codes, 209

Plasticity theory, 323Plastic load line displacement, 115Plastic pipelines, Canadian standards, 171Plastic strain correction factor (Ke), 245

French codes, 247Plastic zone size, 47–48, 113, 122Plastic zone size correction, 113Plate-and-shell theory, 317Plates

center-cracked, loaded to failure, 113–114construction materials, 149European standards, 236J-R curve parameters, 124for pressure equipment, Japanese codes, 263

Plutoniumdouble containment, 343–348, 354–355double containment rule elimination petition (1998), 349–350double containment rule elimination proposed (1997), 349–352double containment rule for high level waste, elimination (1998

final rule), 349–354, 349packaging of fissile material, 341sea transport, 352shipment and quality assurance, 343–348solid exemptions from double containment requirements,

344–345vitrified high level waste (1997 proposed rule), 349–352

Plutonium isotopes, 337Plutonium nitrate, 342–347Plutonium oxide, 343PMA. See Particular Material Appraisal.Pneumatic testing, 141

transport tanks, 366Pneumatic valves, inservice testing using risk insights, 105P-No. 3, weld procedure, 13Poisson effect, 396Polyethylene tape coatings, for pipeline systems, 409–413Portable tanks, 357–367Postulated flaw size, 122Post-weld heat treatment (PWHT), 13

for low-alloy steel parts, 63of pressure equipment, PD 5500 (U.K.), 312



724 • Index

pressure vessels, Japanese codes, 262of welded components, 12, 17, 19

Power boilers, Japanese codes, 271Power Generation Facilities Codes Committee, 258Power uprate and license renewal, 685PRA. See Probabilistic risk assessment/analysis.PRA Standard. See American Society of Mechanical Engineers

Probabilistic Risk Assessment (PRA) Standard.PRCI. See Pipeline Research Council International.PRDs. See Pressure relief devices.Precipitation-hardened austenitic steels, fast breeder reactor material,

251Precipitation hardening alloys, in pressure equipment, 157Precracked specimen tests loaded to failure, 44Predicted instability load, 116–117, 125Predicted time to crack initiation, 84Predictive model, in determining PWR component performance, 60Preheating, requirements of welds, 19Preliminary Safety Analysis (PSA) Applications Guide, 96, 106Preload, 57, 59Preservice examination, 103Presidential Commission, to investigate Three Mile Island, 89Presidential Decision Directive 63 (PDD-63), 420Pressure

of cylindrical shell, 313–314designing for fluctuations in, 20–21low upper-shelf energy evaluation, 122maximum allowable (PS), 132, 134

Pressure accessories, 134in scope of PED, 130–132

Pressure-area method, 314, 316, 329Pressure boundary piping, 1, 35, 55, 100Pressure coils, fired-heater, 168Pressure cookers, in scope of PED, 133, 154Pressure equipment

components included, 310European system vs. U.S. System, 149hazard, level of, 131–133, 135

Pressure Equipment Directive (PED) ((97/23/EC), 129–155,183–184, 196, 199, 206, 217–218, 220–221, 308, 310, 321

Annex I (Essential Safety Requirements), 130–131, 133, 138–143,147, 151, 157, 198, 209, 212, 218, 259, 310, 554

basic principles, 135, 138, 599, 613design, 138–139, 151–154manufacturing, 139, 154, 173, 187, 226

Annex II (Conformity Assessment Tables), 131, 133, 134, 198,209, 217, 218, 310, 452,

Annex III (Conformity Assessment Procedures), 130, 131, 135,198, 209, 217

Annex IV (Criteria of the Notified Bodies), 138Annex V (Criteria of the User Inspectorates), 144Annex VI (CE marking), 144Annex VII (Declaration of Conformity), 144Annex Z, 147, 149, 193, 310, 325annexes, 131, 133, 153, 271, 361, 364, 368, 665, 668Article 1 (Scope and Definition), 131, 133, 153, 153, 271Article 2 (Market Surveillance), 131, 260, 271, 272, 665Article 3 (Technical Requirements), 131, 134, 138, 156, 201, 271,

310, 368, 667, 668Article 4 (Free Movement), 131, 171, 368, 369, 668, 670

Article 5 (Presumption of Conformity), 131, 138, 270, 271, 272,368, 369, 561, 668

Article 6 (Committee on Technical Standard and Regulations),271, 668

Article 7 (Committee on “Pressure Equipment”), 271, 272, 668Article 8 (Safeguard Clause), 131, 271, 668Article 9 (Classification of Pressure Equipment), 131, 133, 198,

209, 217, 271Article 10 (Conformity Assessment), 131, 135, 198, 209, 217Article 11 (European Approval for Materials), 131, 142, 156Article 12 (Notified Bodies), 131, 554, 557, 663Article 13 (Recognized Third-Party Organizations), 154, 272Article 14 (User Inspectorates), 271, 272Article 15 (CE Marking), 155Article 16 (Unduly Affixed CE Marking), 272Article 17 (Appropriate Measures), 272Article 18 (Decisions Entailing Refusal or Restriction), 272Article 19 (Repeal), 272Article 20 (Transposition and Transitional Provisions), 131, 272Article 21 (Addressees of the Directive), 272articles, 272category A, 201category B, 201category C, 201category D, 201category Ex (Exceptional), 201classification of pressure equipment, 131comparisons with ASME Code, 147conformity assessment categories (I to IV), 310conformity assessment modules, 131, 135, 136, 310conformity assessment procedures, 129, 130, 131, 133, 135, 136,

137, 156, 198, 209, 212, 217, 222, 310definition, 146development of, 324and EN 13445, 129Figure 47.1 (PED Flow chart), 131, 132Figure 47.2 (Hazard Categories for a Vessel Containing a

Dangerous Gas), 134Figure 47.3 (Determination of Hazard Category for a Piping

Containing a Dangerous Gas), 134, 135final assessment and proof test, 141flow chart, 120, 277, 278, 380, 570,591Fluid Group 1, 310Fluid Group 2, 310vs. French codes, 191, 193, 196, 253, 653guidelines, 144hazard categories, 133, 134, 138, 198, 209, 212, 217, 218, 220,

627industrial piping, 142, 191, 553, 554industrial piping risk assessment, 219link with COVAP, 217link with codes and standards, 192, 193material specifications, 141, 143, 148, 163, 177, 188New Approach Directives, 129, 131, 138, 144, 145, 147Notified Bodies, 129, 137, 138, 142, 310objectives and requirements, 130vs. published document (PD 5500 (U.K.)), 309, 531vs. RCC-M, 248risk assessment of pressure vessels, 320scope, equipment covered, and exclusions, 137technical documentation, 136Table 47.4 (Selection of Conformity Assessment Procedures), 136

Post-weld heat treatment (PWHT) (continued)




Table 47.5 (List of New Approach Directives (as of January2005)), 144, 145, 146

Table 47.6 (European System vs. U.S. System PressureEquipment), 148–149

Web site, 130Pressure Equipment Regulations 1999, 309, 330Pressure excursions, 45Pressure Equipment Directive (PED) 97/23/EC, 547, 553, 561, 563Pressure gauges, in scope of PED, 129, 148, 191, 259, 308Pressure hazard level, 131, 133Pressure limiting devices, in pressure equipment, 141, 157Pressure load, 44, 47, 49Pressure piping, Canadian standards, 162, 168, 170, 288, 422Pressure regulations, in scope of PED, 626, 628, 633Pressure relief devices (PRD), 169, 192, 357, 358,, 368, 370, 666,

677Canadian standards, 160

Pressure stress intensity factor, 47, 49Pressure Systems Safety Regulations 2000, 310Pressure-temperature (P-T), rate of temperature change affecting, 49Pressure-temperature (P-T) limit, 49Pressure-temperature (P-T) limit curves, 49

pressurized water reactor heatup and cooldown curves, 45, 60, 63Pressure test, minimum internal, 249Pressure testing, 18, 108, 148, 153, 171, 283, 308

Canadian standards, 160French codes, 191, 193, 196, 253pipeline systems, 372risk categories and, 96, 526risk-informed initiatives, 107, 108transport tanks, 357, 358

Pressure tube, 159containing pressurized coolant in CANDU® design, 163

Pressure vesselboiling water reactor, 10, 25boiling water reactor probabilistic fracture mechanics for

inspection exemption, 10burial, Canadian standards (Annex A), 159, 160Canadian non-nuclear standards, 162Canadian standards, 159, 160, 161, 162, 163, 168categorization, 134French codes, 191, 193, 196, 253, 653hazard categories if containing a dangerous gas, 133inservice inspection, Canadian,181inspection, French codes, 206Japanese codes, 257, 258, 259multilayer, Japanese codes, 262nondestructive examination, French codes, 240risk assessment, French codes, 198in scope of PED, 130, 131, 133, 134

Pressure Vessel Research Council (PVRC), EPRI/GE methodology,adoption of, 21

Pressure Vessel Research Council (PVRC) Task Group, 54Pressure Vessel Research Council Task Group on Toughness

Requirements, 44Pressurized food processing equipment, 156Pressurized heavy water reactor (PHWR), 163Pressurized thermal shock (PTS), 30, 32, 44, 51, 56

fracture toughness requirements, 33Pressurized water reactor (PWR), 1

austenitic stainless steel, fatigue crack growth rate, 21–22brittle fracture protection, 45, 49–50

in CANDU® design, 163environmental fatigue effects, 21–22“feed and bleed” items, safety significance of, 95flaw effect on integrity of nuclear components, 43–50, 53, 55French codes, 226hydrogen concentration in primary coolant, 82inclusion criteria (Level A) for (HSS) high-safety significant

snubbers, 106large-diameter pipe weld repair, 81lithium concentration and pH of primary coolant, 82LTOP for brittle fracture protection, 43, 49nozzle cracking, 10operating cycle, 74passive structural components, 57pressure-temperature heatup and cooldown curves, 43, 45–49primary water stress corrosion cracking and, 69, 73, 78primary coolant water chemistry, 67, 78reactor coolant water chemistry changes, 67, 80risk-informed process, 98top-head nozzles, repairs, 81–82zinc added to coolant, 82

Pressurized Water Reactor (PWR) Owner’s Group, aging mechanism study programs effects, 57

Pressurized water reactor plantpersonnel radiation exposure, 53plant safety, 53use of alloy 600 base metal, 63

Pressurized water reactor (PWR) vessel(s)absence of inner surface flaws, 51–52alloy 600 applications, 63–66beltline material, 44–45, 49beltline region, brittle failure at, 43, 55beltline weld, 48degradation predictions of PWSCC, 76–79embrittlement, 50failure/fracture, 43–44Figure 44.1 (Locations with Alloys 600/82/182 Materials in

Typical PWR Vessel), 63–64inspection methods of PWSCC and requirements, 71–72integrity analysis, 50, 54–55primary water stress corrosion cracking (PWSCC), 63, 66–68primary water stress corrosion cracking of alloy 600 material,

operating experience, 68–71probability of failure as a function of pressure temperature, 55remedial measures of PWSCC, 80–81repairs of PWSCC, 79–82safety considerations of PWSCC, 73–74strategic planning for PWSCC, 83–84surveillance program, 50top head insulation, 74, 75toughness level of plates, 50

Pressurized water reactor (PWR) vessel internalsaging management of, 57–60aging management strategies, 59–60aging mechanisms, 57enhanced visual (VT) examinations, 60irradiation-assisted stress corrosion cracking, 59irradiation embrittlement, 59stress corrosion cracking, 59stress relaxation, 59structure/component, loss of material due to aging, 58support, in event of structural failure, 66



726 • Index

thermal aging embrittlement, 59void swelling, 59

Pressurizer heater sleeve, 84–85use of alloy 600, 63, 68

Pressurizer welds, 73Prestressed concrete containment vessels (PCCV), Japanese codes,

287Primary bending stress, 18, 118–119, 121. See also Stresses. of

containment vessels for radioactive materials, 346–347transport tanks, 362

Primary bending stress intensity, 272Primary coolant system

alloy 600/82/182 cracks/leaks, 63boron added in PWR plants, 73hydrogen concentration, 82leaks, 63, 69, 72lithium concentration and pH, 82zinc addition, 82, 84

Primary heat transport system (PHTS), 163–165, 174, 177Primary loading, 109Primary membrane stress, 18, 118–119, 121. See also Stresses.

French codes, 242, 253nuclear power plant piping, 295, 299nuclear pressure vessels, PD 5500 (U.K.), 324

Primary membrane stress intensity, 273Primary stress, 125Primary water stress corrosion cracking (PWSCC), 64

of alloys 600/82/182 in PWR plants, 63–82as axial, 67causes: environmental, 66–68causes: material susceptibility, 66–67causes: tensile stresses, 66–67conditions of PWSCC susceptibility, 68crack arrest, 69crack growth, 67crack growth behavior in alloy 600, 79cracking issue in pressurized water reactors, 74crack initiation, 67, 76, 79description, 66inspection methods/requirements to identify, 71–72predicting time to crack initiation, 76, 84in PWR RPV inlet/outlet nozzles, 74remedial measures, 82–84repair of RPV alloy 600 components, 80resistant materials, 84small cracks, 73susceptibility of alloys 81/182, 66

Principal (CODAP), 207Principal, for boilers, French codes, 224Probabilistic EPFM, 126Probabilistic failure mechanics (PFM), 94Probabilistic fracture mechanics (PFM) analysis, 7, 55, 79

as alternative for assessing margins in Appendix G method, 56code, VIPER, 11for inspection exemption, 6predicting PWSCC on Alloy 600/82/182 in PWRs, 76, 79

Probabilistic risk assessment analysis (PRA), 33applications, piping systems, 95to assess risk of leaks, 84background, 89capability category I, 92–93

capability category II, 92–93capability category III, 92–93codes and standards guiding, 102–103, 106–107component ranking, plant specific, 104to determine allocation of resources, 89to determine inservice activities, 89, 94, 107–108to determine risk importance, 89impact, 92Level 1, 106Level 2, 106Level 3, 106limitations, 100piping system examinations, 96–97plant-specific to determine safety significance of SSCs, 99ranking measures, 104RI-IST and, 104shutdown, 104, 106, 114for valves, 115

Probabilistic Risk Assessment (PRA) Standard. See AmericanSociety of Mechanical Engineers Probabilistic RiskAssessment Standard.

Production from a well, measurement of, 372Production weld test coupons, 248Product verification, 135–136Proof test, 141, 148

for cast iron boilers, 169pipe fittings, 169of pressure equipment, 154–155transport tanks, 366

Property damage, from pipeline incidents, 371–374PS. See Pressure, maximum allowable.PSA. See Preliminary safety analysis.PSAR. See Preliminary Safety Analysis Report.PSDAR. See Post-Shutdown Decommissioning Activities Report.PT. See Liquid penetrant examination.PT. See Penetrant testing.P-T. See Pressure-temperature.PTS. See Pressurized thermal shock.Public Law 104, 358Public Law 104–113 (National Technology Transfer and

Advancement Act), 354–355Published Document (PD) 5500 (United Kingdom), 138–139,

309–314Annex A, 317, 320, 323–324Annex B, 313Annex C, 319–325, 330Annex D, 325Annex G, 319–320, 324, 329–330Annex G.2, 320Annex G2.5, 315Annex K, 312Annex M, 313–314Annex Z, 310Appendix F, 316bolted flanged joints, 316–317design, 312, 316–317design for fatigue, 320–321Enquiry case 5500/116, 329Enquiry case 5500/122, 330Enquiry case 5500/126, 318Enquiry case 5500/128, 318Enquiry case 5500/130, 329

Pressurized water reactor (PWR) vessel internals (continued)




Enquiry case 5500/133 (Rectangular, Narrow-Faced and Full-Faced Flanges), 317

Figure 51.1 (Values of Coefficient _ for Cone/Cylinder Intersectionwithout Knuckle), 313

Figure 51.2 (Buckling Forms for Stiffener Cylindrical Shells),313–314

Figure 51.3 (Theoretical Buckling Strain e as a Function of ShellLength, Radius, and Thickness), 314–315

Figure 51.4 (Nondimensionalised Allowable External Pressure vs.Theoretical Collapse Load), 314, 316

Figure 51.5 (Jacketed Vessel Types), 318Figure 51.6 (Limpet Coil Arrangement), 318Figure 51.7 (Limpet Coil Arrangement for Use in Stiffening for

External Pressure Loading), 318Figure 51.8 (ASME-Based/Old BS 5500 Fatigue Design Curve),

321Figure 51.9 (Fatigue Design Curves from Annex C of PD 5500),

321–322Figure 51.10 (Annex A Stress Categories and Limits of Stress

Intensity-Hopper Diagram), 324Figure 51.11 (Dished End Thicknesses Compared for 2:1

Ellipsoidal Form), 328Figure 51.12 (Dished End Thicknesses Compared for 10%

Torispherical Form), 328Figure 51.13 (Dished End Thicknesses for 6% Torispherical Form

Compared), 328flat plates and covers, 317Form X, 311inspection, 319jacketed vessels, 318loads, local, 319–320materials, 311–312nozzle reinforcing, 314–316sections and appendices, 311shells under external pressure, 313–316shells under internal pressure, 312–313supports, 319Table 51.1 (Comparison of the Bases of ASME and PD 5500

Fatigue Methods), 321Table 51.2 (Fatigue Design Curves, Details of), 322–323Table 51.3 (Nominal Design Stresses), 326Table 51.4 (Testing Groups for Steel Pressure Vessels), 327testing, 319tubesheets, 324

Published Document (PD) 6439 (Stress Calculation Methods forLocal Loads and Attachments of Pressure Vessels), 313

Published Document (PD) 6497 (Stresses in Horizontal CylindricalPressure Vessels), 311, 319, 330

Published Document (PD) 6550 (Supplement to BS 5500), 311, 313,324, 330–331

PUC. See Public Utility Commission.Pumps

cavitation, 45group A, 105group B (standby), 105high-safety significant (HSS) category, 105low-safety significant (LSS) category, 104, 108OMN-Code testing program, 105Risk-informed IST application, 103seal, 50

Pump sizing, French codes, 246–247Puncture/tearing test, 336

Pure water stress corrosion cracking. See Primary water stresscorrosion cracking (PWSCC).

PVE/Pressure Vessels, 309PVE/1, Pressure Vessels (technical committee), 309PVE/1/15 Design Methods, 309PVRC. See Pressure Vessel Research Council.PVRUF reactor pressure vessel, 52PWHT. See Postweld heat treatment.PWR. See Pressurized water reactor.PWSCC. See Primary water stress corrosion cracking.Pyrophoric liquids, 340

QA. See Quality assurance.QAPP. See Quality Assurance Program Plan.QC. See Quality control specialists.QI. See Qualified Inspectors.QIO. See Qualified Inspection Organization.Qualification of welders, oil and gas pipeline systems, Canadian, 170Qualified Inspectors (QI), 366Qualified Inspection Organizations (QIO), 366Qualification of NDT for ISI, 568Quality assurance (QA), 135–136, 144

focusing in CDF-vulnerable components, 90French codes, 229, 233–234, 241pipe fittings, 169plutonium shipments, 343–345of pressure equipment, 156pressure vessels, Japanese codes, 268radioactive material packagings, 343–344, 345

Quality assurance program, 101Canadian standards, 163CANDU® nuclear power plants, 171, 173, 177–178radioactive material packaging, 349–350

Quality Assurance Program (Z series), 162Quality Assurance Requirements for Transport Packages, 1978

effective rule, 345Quality control

licensee of fissile material shipments, 338spent fuel storage containers and transportation casks, 452

Quality Control Program, Canadian standards, 168–169Quality Control Program Manufacturers of Fittings, Canadian

standards, 168Quality Management Systems (CAN/CSA-ISO-9001-00), 162Quantitative risk analysis, of pipeline failure possibility, 382Quenched and tempered non-alloy steels, for pressure vessels, French

codes, 223–224Quenched and tempered steels

for industrial piping, French codes, 224for pressure equipment, French codes, 223

Radial/shear stress, concrete containment vessels, 288Radial shrinkage, in weld repairs, 18Radiation damage, on pressurized water reactor vessel materials, 43Radiation embrittlement, 43, 54, 125Radiation exposure, 339

employees, 53Japanese codes, 292–294Off-site, 30

Radiation shielding, 335Radiation unit, 337Radioactive materials, responsibility for cleaning up spills, 340Radioactive release, risk assessment of, 90



728 • Index

Radiographic inspectionCanadian standards, 171CANDU® nuclear power plant components, 184, 185, 187French codes, 240, 249pressure vessels, Japanese codes, 264–265welds in calandria assemblies, Canadian, 175zirconium alloy components, 176–177

Radiography, 400Radiological exposure hazard models, 292Radiological impacts, fusion reactors, Japanese codes, 292–294Radiotoxicity of isotope, 436Radium, shipments of, 342RAI. See Request for additional information.Ramberg-Osgood curve, 114Ramberg-Osgood model, 121Ramberg-Osgood parameters, 119Ramberg-Osgood stress-strain equation, 121RAMSES committee, 193Random (sample) testing, joint coefficients allowed, 139Ratcheting

nuclear pressure vessels, PD 5500 (U.K.), 323prevention, French codes, 243, 251thermal, in cylindrical containment vessels, 345

Ratcheting fatigue, 299–300low-cycle, 299–300piping failure during earthquakes, Japanese codes, 295seismic shakedown, Japan, 295–296

RAW. See Risk achievement worth.RCC-C. See Design and Construction Rules for Fuel Assemblies of

Nuclear Power Plants.RCC-E. See Design and Construction Rules for Electrical Equipment

of Nuclear Islands.RCC-G. See Design and Construction Rules for Civil Works of PWR

Nuclear Islands.RCC-I. See Design and Construction Rules for Fire Protection.RCC-M. See Design and Construction Rules for Mechanical

Components of PWR Nuclear Islands.RCC-MR. See Design and Construction Rules for Mechanical

Components of FBR Nuclear Islands.RCC-P. See Design and Construction Rules for System Design,

French Codes.RCCV. See Reinforced concrete containments vessels.RCRA. See Resource Conservation and Recovery Act.RCS. See Reactor coolant system.Reactivity control units, CANDU® nuclear power plants, 175Reactor building, 165–166Reactor coolant

environmental impact on components, 34–35, 37temperature, LTOP setpoint and, 50

Reactor coolant system (RCS)aging mechanisms, 57levels of corrosion products in, 63metal fatigue, 21piping, 31pressure boundary, integrity, 30, 98–99, 104primary coolant system cracks/leaks, 63PWSCC occurrences, 68

Reactor pressure vessel (RPV), 1beltline materials, 126bottom head, 65, 71–73, 77end-of-life value, 55environmental fatigue effects, 21

failure probabilities, 55–56ferritic steels, local brittle zones, 53integrity limits, 54–56lowering head temperature, 84nozzles, 9–11, 65, 72, 77–79, 83pressure boundary, 57repair/replacement activity, 78–80, 84top head, 65, 69, 72, 84top head insulation, 72top head nozzle leak, 72–73, 80–82, 84top head PWSCC, 69, 83upper shelf energy, 15, 124vessel-to-shroud support weld cracking, 13wastage (Davis Besse) of low-alloy steel, 69weld examinations, 6–7, 69

Reactor pressure vessel internals, French codes, 247Reactor pressure vessel outlet nozzle butt welds, 69–70Reactor Safety Study, 89Redundancy principles, 58Reference fracture toughness curves, 50–55, 59Reference limit load bending stress, 118Reference load, 115Refined hydrocarbon product, types, 371Refineries, risk analysis and security of, 110Refrigeration, Japanese codes, 261Refrigeration equipment, Canadian standards, 168Refueling outage, 84

inservice examination during, 8, 71–72repair/replacement during, 13, 84

Refueling station pressure piping systems, 170Registration

Canadian standards, 168–169CANDU® nuclear power plants, 172, 174

Registration numbersCanadian, 169–170CANDU® nuclear power plants, 174regulation of pressure equipment in Spain, 563

Regulation on Pressurized Apparatus, 563Re-heaters, 156Reinforced concrete containments vessels (RCCV), Japanese codes, 287Reliability methods, first- and second-order, 126Relief valves, 45

repair guidelines, Canadian standards, 168–169Remedial measures for PWSCC test program, 75, 82Repair, 84

of boiling water reactors, 1cost of, 84of flaws, 17–19, 80–81of intergranular stress corrosion cracking in stainless

steel piping, 17pipeline systems, 375, 402–407of pressurized water reactors, 79–82, 85of primary water stress corrosion cracking, 63, 79–82weld overlay, 81–82weld replacement, 81–82

Repair/replacement activities, 97, 100–101in age management programs, 59plan document, 102of pressure boundary components, 80reactor vessel heads, 65RI-ISI programs, 95risk-informed, 98




structural integrity treatment, 101technical requirements, 101–102

Reportable incident, defined, 373Request for additional information (RAI), 33, 36–37, 39Re-rounding, 397Residual stress, 22, 24, 125, 140

from fabrication, 69French codes, 253hoop, 67reversing, 83surface, from machining or grinding, 80, 83susceptibility to PWSCC and, 66–67, 69from welds, 17–18, 51, 55, 73, 77, 81welds, PD 5500 (U.K.), 322–323

Residual stress improvement processes, as remedial measure forPWSCC, 83

Residual stress improvement program, as remedial measure forIGSCC, 83

Response analysis method, 299–300RI-ISI. See Risk-informed inservice inspection.RI-IST. See Risk-informed inservice testing.Ring forgings, 49–50Ringhal 3 nuclear power plant, 70Ringhal 4 nuclear power plant, 70Ring stiffeners, 313–314RISC. See ANS Risk-Informed Standards Committee.Rise-time-based model, for environmental fatigue effects, 22Risk achievement worth (RAW), 106Risk assessment, 89

industrial piping, French codes, 212, 219–220pipeline failure, 380–386pipeline systems, 376pressure vessels, French codes, 198, 200–201

Risk-based criteria, 159Risk categorization, 96, 101

of pipe segment risk evaluation, 95Risk-informed (RI) analysis, 90

applications, 103capability of PRA to support application, 91–94decommissioning of nuclear facilities, 425future plans for, 107–110HSS classification and, 100IST application, 103preservice, 103repair/replacement requirements, 97, 101, 108risk category and, 97safety classification, 98–100, 103, 108security applications, 89, 110standard for use of PRA, 90–93in testing mechanical equipment, 103

Risk-informed (RI) decision-making, PRA Standard application, 91Risk-informed (RI) fracture mechanics evaluations, 126Risk-informed-inservice inspection (RI-ISI), 95, 96, 100

current scope, 93future applications, 108overall process, 95of piping, 99, 100reevaluation, 96

Risk-informed inservice testing (RI-IST), 103–105Risk management, 107

definition, 381Risk neutral situation, 97

Risk studies, 90–91RMA. See Rubber Manufacturers Association.Roll expansion repair, 12–13Role of regulatory authority, 83Roll peening, to reduce potential PWSCC, 59Root cause determination, 35, 59

of component aging, 35RPV. See Reactor pressure vessel.R ratio, 5

environmentally assisted fatigue crack growth in BWRenvironment, 22

RRM, risk-informed, 109RSE-M. See Inservice Inspection Rules for Mechanical Components

of PWR Nuclear Islands.R-6 methodology, 114, 121“R” Stamp, 366–367RSTRENG, 376, 399–400RTNDT brittle to ductile transition temperature determination,

French codes, 249RTPO, 148

approval of joining procedure qualifications, 140approval of NDE examiners, 148

Rubber Manufacturers Association (RMA), 162Rubber Manufacturers Association standards, RMA IP-2, 170Rules on Design and Construction for Nuclear Power Plants, 275Rules on Fitness-for-Service for Nuclear Power Plants

(Japan, 2000), 275Rules on Thermal Power Generation Facilities, 259Rupture, 83–84

of radioactive material packaging, 352–353, 354Rupture disks, 359Russian Regulation and Codes in Nuclear Power, 601

Sacrificial cathodic systems, 169, 189Safe-end welds, 17, 19, 24Safeguard action, 144Safe operation, 145Safe shutdown, 30, 98Safety

different classes of packages of special nuclear material, 334emergency response plans of pipeline companies, 421identifying concerns using PRA, 89Japanese codes and standards, 259plant overall, 53RCC-M French codes, 228–229, 249risk from boric acid corrosion, 74

Safety accessories, 134in scope of PED, 130–131, 139, 153–154

Safety analysis, 32Safety analysis report (SAR), 418Safety classification, 100

of HSS/LSS component categories, 100, 106risk-informed, 96, 98–100, 103, 109–110safety-related (SR) vs. no safety related (NSR) classification, 100

Safety coefficients, 151–152Safety devices

French codes, 249on pressure equipment, inspection, 141, 155, 168, 249

Safety evaluation (SE), 32–33of BWR stainless steel internals, 2–3

Safety evaluation report (SER), 33–34, 37, 39Safety factors. See Structural factors.



730 • Index

Safety functionscore damage protection, 96large early release frequency, 96during shutdown, 104

Safety hazards, radioactive material accidental release, 181Safety lock, fuel-handling equipment, CANDU® nuclear power

plants, 175Safety margins, 100

French codes, 238–239, 245of nuclear reactors, 43, 49–51overall plant safety and, 53pipeline systems, 384for pressure equipment, PD 5500 (U.K.), 321of pressurized water reactor vessels, 43probability fracture mechanics used, 56

Safety measuresfusion reactors, Japanese codes, 292–294Japanese codes, 260–261

Safety objectives, 129–130, 138Safety related (SR)

definition, 98structures, systems, and components (SSC), special treatment, 98

Safety-related (SR) classification, 100Safety/relief devices, Canadian standards, 169Safety Report, French codes (RCC-M), 215, 230, 250Safety requirements, 147–151Safety reviews, for license renewal, 30–31Safety significance categories

HSS (high-safety significance) as, 104LSS (low-safety significance) as, 104

Safety systemseffects of aging on, 58low-safety significance of, 100

Safety valves, 141French codes, 249repair guidelines, Canadian standards, 168

Sample package, 336Sample size, 58Sampling, as inspection method, 58SAR. See Safety analysis report.SAR. See Synthetic aperture radar imaging.SARA. See Superfund Amendment and Reauthorization Act.Satellites-optical systems, 418

surveillance of pipeline systems, 418–419SAW. See Submerged arc welding.SBC. See Systems-Based Code.SCADA. See Supervisory Control and Data Acquisition.Scale model testing, 6SCC. See Standards Council of Canada.SCC. See Stress corrosion cracking.SCO. See Surface contaminated object.Scoping methodology, 41

for license renewal, 32review process, 32–33

SCRAM (rapid reactor shutdown), 20–21, 31, 32–33Screening methodology, 41, 59

for license renewal, 32review process, 32–33

SCV. See Steel containment vessels.SDO. See Standards-developing organizations.SDWA. See Safe Drinking Water Act.SE. See Safety evaluation.

Secondary bending stress, 118–119. See also Stresses.Secondary stresses, 124–125

in containment vessels for radioactive materials, 346membrane, French codes, 251–252nuclear power plants, Japanese codes, 294–295nuclear pressure vessels, PD 5500 (U.K.), 324

Section I (Power Boilers), 147, 169–170, 188, 258–259, 364vs. COVAP (French Boiler Code), 224vs. Japanese codes, 268–270

Section II (Materials), 267, 286Appendix 1, 287Appendix 2, 287Appendix 5, 287vs. French codes, 230vs. Japanese codes, 268–269, 286–287Part A (Materials: Ferrous Material Specifications), 169, 188PartB (Materials: Nonferrous Material Specifications), 169, 188Part C (Materials: Specification for Welding Rods, Electrodes and

Filler Materials), 169, 188Part D (Materials: Properties), 169, 188, 275, 359, 362–363Table U, 311vs. PD 5500 (U.K.), 311

Section III (Power Piping Codes), 118, 124, 193Addenda, 295allowable stresses for reactor vessel components, 67for Canadian nuclear construction standards, 159Class 1 systems, 1, 6, 346Class 2 systems, 108Class 3 systems, 108, 293, 295Code cases, 108developing reliability-based load and resistance factor design

methods for piping, 107fatigue design curves, 21Figure 42.1 (Audit of AMPs Consistent with the GALL Report),

36Figure 42.2 (Audit of Plant-Specific AMPs), 37Figure 42.3 (AMP Review Process, Consistent with GALL

Report), 38Figure 42.4 (AMR Review Process, Consistent with Precedent), 39Figure 42.5 (Interim Staff Guidance Process Flow Chart), 40–41Figure 43.8 (Charpy V-Notch Surveillance Data Showing RTNDT

Shift Due to Irradiation), 50–51Figure 44.1 (Locations with Alloys 600/82/182 Materials in PWR

Vessel), 64Figure 44.2 (Typical Control Rod Drive Mechanism (CRDM)

Nozzle), 65Figure 44.3 (Bottom-Mounted Instrument (BMI) Nozzle), 65Figure 44.4 (Typical Reactor Vessel Inlet/Outlet Nozzle), 66Figure 44.5 (Typical Core Support Lug), 66Figure 44.6 (Alloy 600 Crack Growth Rate at 338°C Plotted vs.

Hydrogen Concentration), 67–68Figure 44.7 (Effects of Hydrogen Concentration on PWSCC

Initiation and Growth), 68Figure 44.8 (Typical Small Volume of Leakage from CRDM

Nozzle), 69, 71vs. French codes, 226, 229, 236–237, 246–247intergranular stress corrosion cracking, 1vs. Japanese codes, 272–273, 275, 284joint design with American Concrete Institute, 400material fracture toughness requirements, 348nuclear requirements, 102pressure-retaining components, 20




pressurized water reactor vessel design requirements, 43service-induced degradation in BWR vessels, internals, and

pressure boundary piping, 24Table 42.1 (Consistent with GALL Report Classification), 33–34Table 42.2 (Elements of an Aging Management Program), 35Table 44.1 (Factors on Crack Initiation and Growth Time at

Typical PWR Temperatures), 67, 76Task Group, environmental fatigue effects, 21

Section III, Division 1 (Rules for Construction of Nuclear PowerPlant Components), 34, 42, 171–173, 188, 194, 259

Appendix I, 358Appendix III, 45Appendix XIII, 237Appendix XIV, 237Appendix A, 37Appendix B, 173Appendix C, 173Appendix G (Protection Against Nonductile Failure), 43–46, 60,

113, 124, 245Figure 43.9 (ASME Code’s KIR Toughness Curves), 51G-2120 (Maximum Postulated Defect), 56KIR curve, 53

vs. EN 13445, 327fatigue design procedure, 20Figure 19.2 (Fatigue Design Curve for Ni-Cr-Fe), 14maximum postulated defect, related to allowable surface

indications (Section XI), 44vs. RCC-M, 254required weld overlay thickness, 20Subsection NA, 347

Article III–2000, 347Subsection NB, 1, 172, 189, 248, 290, 347

NB-2300, 44, 60NB-2330, 44NB-2331, 50NB-3000, 18NB-3200, primary stress intensity limits, 20, 175, 238NB-3222.4, 245NB-3228.5, 274NB-3600, 239, 246NB-3650, 18NB-3661.2, 245NB-4000, 238

Subsection NC, 172, 176, 236NC-3200, 175, 237NC-3300, 175NC-3352.4(d), 175NC-3671.2, 175NC-3800, 247NC-3900, 247

Subsection ND, 172, 248, 347vs. Japanese codes, 294ND-3671.2, 175ND-3800, 247ND-3900, 247

Subsection NE, 171, 176Subsection NF, 174, 347

NF-2000, 177NF-3200, 175NF-3300, 175

Subsection NG, 1, 347Subsection NH, 275

Section III, Division 2 (Code for Concrete Reactor Vessels andContainments), 171–172, 189

Class MC, components, provisions for, 172vs. Japanese codes, 270, 287, 289Subsection NCA, 172–173

NCA-2142, 173NCA-2143, 173NCA-3250, 173NCA-3550, 173NCA-3800, 102

Section III, Division 3 (Containment Systems for TransportPackaging), 174

adoption by USNRC, 354, 353vs. Japanese codes, 268–269

Section III, Division 4, vs. Japanese codes, 291Section IV (Heating Boilers), 169Section V (Nondestructive Examination), 169, 177,

183, 188vs. Japanese codes, 270

Section VIII, 138–139, 317canister design requirements for radioactive materials,

349vs. Japanese codes, 270nuclear requirements, 102pressurized water reactor vessel design requirements, 43

Section VIII, Division 1 (Rules for Construction of Pressure Vessels),147, 169–170, 258–259, 347, 360, 364

compared to CODAP rules (French codes), 208vs. EN 13445, 328–329vs. Japanese codes, 259, 272production tests, 264Subsection A (General Requirements for All Methods of

Construction and All Materials), 151Part UD, 359Part UG, 359–360

UG-22, 361Figure UG-31, 262UG-44, 169UG-46, 179UG-90, 365UG-93, 366

Section VIII, Division 1, Subsection B (Requirements Pertaining toMethods of Fabrication of Pressure Vessels)

Part UW, 365–367UW-2, 364UW-3, 364Table UW-12, 262UW-40, 365UW-50, 366UW-51, 366UW-52, 366

Section VIII, Division 1, Subsection CPart UCS (Requirements for Pressure Vessels Constructed of

Carbon and Low-Alloy Steels)Table UCS-23, 262–263UCS-56, 365UCS-85, 262

Part UHA (Requirements for Pressure Vessels Constructed ofHigh-Alloy Steel)

Table UHA-23, 262–263Part UHX, 208

UHX-12, 208



732 • Index

Part UNF (Requirements for Pressure Vessels Constructed ofNonferrous Materials)

Table UNF-23, 262–263Section VIII, Division 1, Appendices 1–32

Appendix 1, vs. PD 5500 (U.K.), 312Appendix 13, 329Appendix 26, 208Appendix L, 363Appendix EE, 318

Section VIII, Division 2 (Alternate Rules for Pressure Vessels), 20,169,188–189, 288

class 1 vessel, 266–267compared to CODAP rules (French codes), 208

Section VIII, Division 2, Mandatory AppendicesAppendix 4, 362Appendix 5, 362

Section VIII, Division 3, 169, 188, 261Section IX (Welding and Brazing Qualifications), 169, 188

brazing and welding procedures, Canadian, 169vs. French codes, 253vs. Japanese codes, 264, 270Part QW, 248welding requirements, 148

Section X (Fiber-Reinforced Plastic Pressure Vessels), 169, 188Section XI (Inservice Inspection of Nuclear Reactor Cooling

Systems), 4, 87, 103, 127, 181, 187–188Addenda (2002 Edition), 118–119, 121allowable surface indications, related to maximum postulated

defect (Section III), 56BWR requirements, 1BWR steam dryer assembly/steam flow path, 6–7CANDU® equivalent, 159Class 1 components

fabricated with Alloys 600/82/182 materials, 71high-safety significance (HSS) items, 95piping, 98

Class 2 componentshigh-safety significance items, 95piping, 98systems, 108

Class 3 componentshigh-safety significance items, 95piping, 89, 98systems, 108

Code Cases, 94–106, 108, 109code requirements for safety relation, 95conditional consequence of failure, 93Figure 41.1 (Overview of BWR Pressure Vessel and Internal

Components), 1–2Figure 41.2 (BWR Core Shroud Weld Designations), 3Figure 41.3 (A Distributed Ligament Length Example), 3Figure 41.4 (Typical Geometry of a BWR Jet Pump), 5Figure 41.5 (Sample of Stress Time History at Cracked Location),

5, 9Figure 41.6 (Crack Lengths for Core Flow Levels), 6Figure 41.7 (BWR Steam Dryer Assembly), 7Figure 41.8 (Steam Dryer Damage), 8Figure 41.9 (Feedwater Nozzle with Cracking Location), 9Figure 41.10 (Improved Sleeve Design and Temperature

Variation), 11Figure 41.11 (Fracture Mechanics Results for BWRs), 9, 12

Figure 41.12 (BWR Feedwater Nozzle Inspection Zones), 12Figure 41.13 (BWR Set-in CRD Stub Tube Design), 12Figure 41.14 (Stub Tube Narrow Groove Welded Partial Design), 13Figure 41.15 (BWR-2 Shroud Support Geometry), 14Figure 41.16 (Calculated Values of Total K and the Polynomial

Fit), 14Figure 41.17 (Predicted Crack Growth as Function of Operating

Hours), 14Figure 41.18 (Steam Dryer Support Bracket Crack), 14–15Figure 41.19 (Temperature-Time Variations during Automatic

Blowdown Transient), 15–16Figure 41.20 (Assessment for Level C Conditions), 15–16Figure 41.21 (Weld Overlay Repair), 17Figure 41.22 (Dissimilar Metal Weld Overlay), 19Figure 41.23 (Design versus Actual Number of Transient Events),

21Figure 41.24 (Severity of Transient Actual Temperature Change

versus Percentage of Design Basis), 21Figure 41.25 (Effect of Loading Conditions on Environmentally

Assisted Fatigue Crack Growth and Comparison with ASMESection XI Curves), 22, 24

Figure 41.26 (Crack Growth Rate Prediction Model), 22–23Figure 41.27 (Comparison of BWRVIP-14 and Japan Maintenance

Code Predictions), 23Figure 41.28 (BWRVIP-60 Stress-Corrosion Cracking Deposition

Lines), 23Figure 41.29 (Crack Length versus Total Time-on-Test), 24Figure 41.30 (Predicted Crack Growth in Safe End), 24

Section XI, Division 1, Nonmandatory AppendicesFigure 43.1 (Mm Factor for Membrane Stress Intensity Factor), 46Figure 43.2 (Mt Factor vs. Thickness for Bending Stress Intensity

Factor), 46Figure 43.3 (Linearized Representation of Stresses for Surface

Flaws), 46–47Figure 43.4 (Examples of 50°F/hr. Cooldown Curves), 48Figure 43.5 (Assumed Axial Flaws in Circumferential Welds), 49Figure 43.6 (Circumferential Flaws in Girth Welds), 49Figure 43.7 (Fixed LTOP Setpoint Affects Operating Window), 50Figure 43.8 (Charpy V-Notch Surveillance Data Showing RTNDT

Shift Due to Irradiation), 50Figure 43.9 (ASME Code KIC Toughness Curves), 51Figure 43.10 (Static Fracture Toughness Data (KJC) Now

Available, Compared to KIC), 52Figure 43.11 (Original Reference Toughness Curve, with

Supporting Data), 52Figure 43.12 (KIC Reference Toughness Curve with Screened Data

in the Lower Temperature Range), 52Figure 43.13 (Original ASME KIC Data and New Variable TKIC-T),

53Figure 43.14 (Original KIC Toughness Data versus T-T0), 54Figure 43.15 (Fracture Toughness Data Normalized to 1T and

Compared to Code Case N-629 Curve), 54Figure 43.16 (Comparison of Residuals from ASTM E 900-02 and

Recent NRC Embrittlement Trend Curve Equations), 55Figure 43.17 (Estimates of Crack Initiation Compared to P-T

Limits for Normal Cooldown Transient), 55Figure 43.18 (Relationship Between Maximum Postulated Defect

and Allowable Surface Indications), 56Figure 43.19 (Framework for Implementation of Aging

Management Using Inspections and Flaw Evaluation), 59–60Figure 44.8 (Typical Small Volume of Leakage from CDRM

Nozzle), 69, 72

Section VIII, Division 1, Subsection C (continued)




Figure 44.9 (Large Volume of Wastage on Davis-Besse ReactorVessel Head), 70, 74

Figure 44.10 (Through-Wall Crack and Part-Depth CircumferentialCrack in V.C. Summer Reactor Vessel Hot-Leg OutletNozzle), 70

Figure 44.11 (Leak from South Texas 1 BMI Nozzle), 71–73Figure 44.12 (Schematic of RPV Top-Head Nozzle Geometry and

Nature of Observed Cracking), 74Figure 44.13 (Plan and Cross-Section through Corroded Part of

Davis-Besse Reactor Vessel Head), 74–75Figure 44.14 (Cross-Section through Davis-Besse Reactor Vessel

Head), 75Figure 44.15 (Distribution of Log-Mean CGR Power Law

Constant for the 26 Heats of Alloy 600 Material with Log-Normal Fit to the Data), 77

Figure 44.16 (Typical Crack Growth Predictions for aCircumferential Crack in a Steep Angle RPV Top-Head(CRDM) Nozzle), 78

Figure 44.17 (Crack Growth Predictions for a PostulatedCircumferential Crack in a Large-Diameter Nozzle ButtWeld), 78

Figure 44.18 (Effect of Inspections on Probability of NozzleFailure for Head Operating Temperature Ranges), 79

Figure 44.19 (Probability of Leakage from a Top-Head Nozzle),79–80

Figure 44.20 (Reactor Pressure Vessel (RPV) Top-Head NozzleFlaw Embedment Repair), 81

Figure 44.21 (Weld Overlay Repair Applied to RPV OutletNozzle), 81

Figure 44.22 (RPV Top-Head Nozzle Weld Replacement Repair),82

Figure 44.23 (Typical Results of Strategic Planning EconomicAnalysis for PPV Head Nozzles), 85

Figure 45.2 (Overall Risk-Informed ISI Process), 90Figure 45.3 (Potential Evolution to Nuclear Systems Code), 109Figure 46.1 (Effect of Fracture Toughness on the Governing

Failure Mechanism), 113–114Figure 46.2 (The EPRI J Estimation Scheme), 115Figure 46.3 (True-Stress True-Strain Curve for A333 Grade 6 Base

Material in NRC/BCL 4111-1 Pipe), 115Figure 46.4 (Fully Plastic J Integral for Circumferential

Through-Wall Flaws in Cylinders), 115–116Figure 46.5 (Determination of Instability J, T, and Associated Load

for Load Control EPFM Analysis), 116–117Figure 46.6 (Net-Section Collapse Load vs. Estimation Scheme

Maximum Load for Axially Loaded 304SS Pipe withThrough-Wall Circumferential Crack), 117

Figure 46.7 (Determination of J and T at Crack Instability forAustenitic SAW at 550°F), 117

Figure 46.8 (DPFAD for Failure Mode Screening Criterion), 119Figure 46.9 (Elastic-Plastic Fracture Mechanics Flow Chart for

Screening Criteria), 119–120Figure 46.10 (Ferritic Material J-T Curves used in EPFM

Evaluation), 119–120Figure 46.11 (Instability Point Determination in DPFAD Space),

121Figure 46.12 (Ductile Crack Growth Stability Evaluation), 123Figure 46.13 (DPFAD for a 1/4T Flaw), 123Figure 46.14 (J Integral-Tearing Modulus (J-T) Procedure),

123–125flaw evaluation procedures, 81, 113–128vs. French codes, 229

future inspection requirements of dissimilar metal butt welds, 72inspection requirements for alloy 600 components, 72inspection requirements for alloys 82/182 welds, 72inspection sample sizes, 182intergranular stress corrosion cracking, 1vs. Japanese codes, 276, 278–280

Section XI (Inservice Inspection of Nuclear Reactor CoolingSystems)

Main Committee, alternate inspection frequency, 11pressure testing, leakage, 108reactor pressure vessel inside surface flaws and, 51repair and replacement, 97–99RI-ISI, 97risk inform code design rules, 89risk inform code requirements for inservice testing, 89service-induced degradation in BWR vessels, internals, and

pressure boundary piping, 24special treatment requirements, 98structural margins for crack growth, 24Table 41.1 (Jet Pump FIV Stress Range vs. Cycle Data), 6Table 41.2 (Feedwater Nozzle/Sparger Inspection

Recommendations), 10Table 41.3 (BWR RPV Equivalent Margin Review Summary),

15–16Table 41.4 (Comparison of Required Thickness of Weld Overlay

Repair), 20Table 46.1 (Fully Plastic .3Integral for Circumferential

Through- Wall Flaws in Cylinders), 115, 117Table 46.2 (Safety/Structural Factors for Circumferential and Axial

Flaws), 118Table 46.3 (Default Material Properties and Z Factors for Ferritic

Piping with Circumferential Flaws), 117Table 46.4 (Z Factors for Circumferential Flaws in Ferritic Piping),

120–121Table 46.5 (Appendix K Requirements), 122Table H-4211–1 (46.3)(Material Properties for Carbon Steel Base

Metals and Weldments), 119Table H-5310-1, 119Table H-5310-2, 120Table H-6310-1 (Load Multipliers for Carbon Steel Base Metals

and Weldments), 119–120Table H-6310-2 (Load Multipliers for Carbon Steel Base Metals

and Weldments for User-Specified Data), 119–120Table H-6320, 120Task Group for Piping Flaw Evaluation, flaw evaluation in

austenitic steel piping, 127Task Group of Subgroup on Welding, 19Task Group on Risk-Based Examination, 942001 Edition, 20volumetric examination of RPV pressure-retaining shell welds, 7White Paper (Reactor Vessel Integrity Requirements for Levels A

and B Conditions), 13, 55Working Group on Flaw Evaluation, 15, 20, 118Working Group on Implementation of Risk-Based Examination, 95Working Group on Operating Plant Criteria, 49–50, 55

Section XI, Division 1 (Revision 13), 26Section XI, Division 1, Subsection IWA (General Requirements)

IWA-1400 (n), (Documentation of Quality Assurance Program),101

Table IWA-2210-1, 60IWA-3300, 125IWA-4000 (Repair/Replacement Activities), 101–102



734 • Index

IWA-4110 (Scope), 101IWA-4120 (Applicability), 101IWA-4130 (Alternate Requirements), 101IWA-4131, 101IWA-4140 (Responsibilities), 101IWA-4142, 101IWA-4150 (Repair/Replacement Program), 102IWA-4160 (Alternative Provisions [Similar to] Verification of

Acceptability), 102IWA-4170 (Inspection), 102IWA-4180 (Documentation), 102IWA-4200 (Items Used for Repair/Replacement Activities), 102IWA-4211(b), 103IWA-4220 (Code Applicability), 103IWA-4300 (Design), 102IWA-4411, 102IWA-4411(e), 102–103IWA-4411(f), 102IWA-4422, 102IWA-4460, 103IWA-4500 (Examination and Test), 103IWA-4520(c) 103IWA-4600 (Alternative Welding Methods), 103IWA-5250(a)(3), 12IWA-6210(e) (Owner’s Report for Repair/Replacement Activities),

101Section XI, Division 1, Subsection IWB (Requirements for Class 1,

2, 3, MC and CC Components and Supports)Table IWB-2500-1 (Examination Category B-D), 6,

10–11, 72Figure IWB-2500-7(a) (Surface M-N), 10Figure IWB-2500-7(b) (Surface M-N), 10Figure IWB-2500-7(c) (Surface M-N), 10Figure IWB-2500-7(d) (Surface M-N), 10IWB-3142.2, 10IWB-3142.3, 10IWB-3142.4, 10Table IWB-3510 (Allowable Flaw Indication), 56Table IWB-3510-3, 10Table IWB-3512-1, 11IWB-3514.2, 127IWB-3520.1, 60IWB-3520.2, 60IWB-3600 (Conditionally Acceptable Flaws), 1, 2, 13, 113, 116IWB-3600 plus Appendix A, 124IWB-3610(d)(2), 18IWB-3611 (Normal/Upset Conditions), 9IWB-3640, 18–20, 27, 116IWB-3740, 21

Section XI, Division 1, Mandatory AppendicesAppendix VIII (Performance Demonstration for Ultrasonic

Examination Systems), 56, 253Section XI, Division 1, Nonmandatory Appendices

Appendix A (Analysis of Flaws), 14, 22, 46–47, 113, 124A-3000, 46, 60fatigue crack growth in water environment, 9ferritic stainless steel crack growth rates, 21KIA lower bound fracture toughness curve for high-rate loading,

53Table A-3320-1, 47

Section XI, Division 1, Nonmandatory AppendicesAppendix C (Evaluation of Flaws in Austenitic Piping), 16, 21–22,

119, 125–126BWR shroud flaw evaluation guideline, 2Figure C-3210-1 (Air Fatigue Crack Growth Rate Curves for

Austenitic Stainless Steel), 5flaw evaluation guidelines, 6flaw sizes allowable (2004 Edition), 118limit load equations, 3, 4source equations, 18–192002 Addenda, 19–20

Appendix E (Evaluation of Unanticipated Operating Events), 8Appendix G (Fracture Toughness Criteria for Protection Against

Failure), 15, 43, 46–47, 60, 113, 121–122, 128determining LTOP setpoint, 50excess conservatism in, 53fracture toughness-based reference temperature, 45–49future need for probabilistic P-T limit curves, 56–57future need to reduce reference flaw size, 56G-2120 (Reference Flaw Size), 48, 56G-2215, 50heatup and cooldown limit curves, 55KIA, lower bound fracture toughness curve for high-rate

loading, 53KIC curve, 531996 Code Change, 49for pressure stresses, 47–48for thermal stress distribution, 48vessel cooldown limits, 56

Appendix H (Evaluation of Flaws in Ferritic Piping), 118–119, 127Appendix K, 15, 128, 124–125, 128Appendix L, 21

flaw tolerance evaluation, 9Appendix R (Risk-Informed Inspection Requirements for Piping),

97Section XII (Transport Tank Code), 357–367Section XII, Mandatory Appendices I-XII, 358

Appendix V, 366Appendix VI, 366Appendix IX, 366

Section XIIfabrication and inspection rules, 364–367marking certification, 358–359Modal Appendices, 364–370

Modal Appendix 1 (Cargo Tanks), 358Modal Appendix 2 (Rail Tank Cars), 358Modal Appendix 3 (Portable Tanks), 358, 367

Article 1 (Cryogenic Portable Tanks), 358, 361, 364Article 1, Table 1-5.2 (Fatigue Loads), 364Article 2 (Non Cryogenic Portable Tanks), 358

Modal Appendix 4 (Ton Tanks), 358Non-Mandatory Appendices A-G, 358Part TD (Design Requirements), 358

TD-101 (Minimum Thickness Design Requirements ofTransport Tanks), 360

TD-102 (Thickness Tolerances of Plates and Piping), 360TD-103 (Thickness Tolerances of Plates and Piping), 360TD-104 (Dimensional Symbols Representing Geometry in

Corroded Condition), 360TD-140 (Maximum and Minimum Design Temperatures), 360TD-150 (Design Pressure and Maximum Allowable Working

Pressure), 360

Section XI, Division 1, Subsection IWA (General Requirements) (continued)




TD-160 (Maximum Allowable Working Pressure), 360TD-200 (Loadings of Transport Tanks), 360–362, 364TD-210 (Maximum Allowable Stresses for Internal and External

Pressure), 362–363TD-301 (Internal Pressure Design), 362TD-312 (Design of Formed Heads under Internal Pressure),

362–364TD-312-1 (Torispherical Heads), 362TD-312-2 (Torispherical Heads), 362TD-312-3, 362Table TD-312 (M Values), 362Table TD-312.1 (K Values), 363TD-313 (Ellipsoidal Heads), 362–363TD-314 (Hemispherical Heads), 362–363TD-315 (Crown and Knuckle Radii), 363TD-400 (External Pressure Design), 363TD-500 (Flat Heads and Covers), 363Figure TD-500, 363TD-603 (Welded and Brazed Connections), 364TD-650 (Rules on Strength of Reinforcement), 363TD-680 (Minimum Nozzle Neck Thickness), 364Table TD-680 (Minimum Nozzle Neck Thickness), 364Article TD-6 (Nozzle Reinforcement Rules and Strength Path

Determination), 363Part TE (Examination Requirements), 366, 364–367

TE-120.1 (Testing Personnel Qualification and Certification),366

Article TE-1 (NDE Personnel and NDE Examination), 366Article TE-2 (Rules for the Examination and Acceptance of

Welds), 366Table TE-230.2 (Radiography of Butt Welded Joints), 366TE-250 (Acceptance Criteria), 366

Part TF (Fabrication Requirements and Repairs of Materials,Vessels, and Vessel Parts), 363, 364–365

Article TF-1 (General Requirements for Fabrication), 365Article TF-2 (Requirements for Welding Fabrication), 365Article TF-3, 365Article TF-4, 365Article TF-5, 365Article TF-6, 365Article TF-7 (Post Weld Heat Treatment), 365Article TF-8 (Requirements for Vessels Lined for

Corrosion/Erosion Control), 365Part TG (General Requirements), 358–359

TG-100 (Definitions), 358TG-102, 363TG-110.2, 365TG-130, 363Table TG-130, 364TG-320 (Manufacturer’s Responsibilities), 365TG-330 (Inspector’s Duties), 365TG-430, 366TG-440, 366Article TG-1 (Boundaries of Section XII), 358Article TG-2 (Organization of Section XII), 358Article TG-3 (Requirements on Responsibilities and Duties of

the Owner, User, and Manufacturer), 358, 365Article TG-4 (General Rules for Inspection), 358, 365–366Table 53.2 (Vessel Classification), 365–366

Part TM (Material Requirements), 358Article TM-2 (Rules on Toughness Requirements), 359TM-110 (Nonpressure Parts), 359

TM-111 (CVN Impact Test Method), 359TM-112, 359TM-113, 359TM-114, 359TM-115, 359TM-116 (Unidentified Materials), 359TM-117, 359TM-118 (Bolts and Studs), 359TM-119, 359TM-120, 359TM-121, 359TM-132, 359Table 132.1 (Carbon and Low-Alloy Steels for Transport Tanks),

359Table 132.2 (High-Alloy Steels for Transport Tanks), 359Table 132.3, 359Table 132.4, 359Table 132.5, 359Table 132.6, 359Table 132.7, 359TM-212 (Impact Test Specimens), 359TM-221 (CVN Acceptance Values), 359Figure TM-221, 359TM-222 (Rules on Lateral Expansion Requirements), 359TM-241 (CVN Exemption Rules for Carbon and Low-Alloy

Steel), 359Figure TM-241 (Allowable MDMT for a Given Material and

Thickness), 359–360Figure TM-241.2, 359TM-243 (Allowable Temperature Reduction in Design

Temperature), 359TM-244 (Impact Test Exemption Guidelines for Carbon Steels),

359–360TM-250 (Toughness Rules on High Alloy Steels), 360TM-260 (Ferritic Steels for Transport Tanks), 360TM-262, 360

Part TP (Requirements for Repair, Alteration, Testing and Inspectionfor (Continued Service), 358, 366–367

TP-100, 367TP-200, 366Article TP-1 (General Requirements and Responsibilities), 366Article TP-2 (Use of National Board Inspection Code), 366Article TP-3 (Rules for When Vessels Inspected), 366Article TP-4 (Inspections and Tests for Transport Tanks), 366–367Article TP-5 (Acceptance Criteria for Tests and Inspections), 367Article TP-6 (Reports and Records from Inspections and Tests),

367Part TR (Rules for Pressure Relief Devices), 358Article TR-1 (Regulations on Set Points and Capacity), 358Article TR-2 (“UV” Valves as Alternative to “TV” Valves), 358Article TR-3 (Nonreclosing Pressure Relief Devices), 358Article TR-4 (Capacity Certification), 359Article TR-5 (Marking and Certification), 359

Part TS (Stamping and Certification Requirements, Manufacturer’sData Reports and Other Records), 358–359

Article TS-1 (Content and Method of Stamping), 359Article TS-2 (Obtaining and Applying Code Symbol Stamps), 359Article TS-3 (Data Reports), 359Article TS-4 (Special Requirements), 359

Part TT (Testing Requirements), 358, 366Article TT-1, 366Article TT-2 (Pneumatic and Hydrostatic Testing), 366



736 • Index

Article TT-3 (Proof Testing for Maximum Allowable WorkingPressure), 366

Article TT-4 (Spark Testing on Vessels with Elastomeric Lining),366

Part TW (Welded Construction Requirements), 358, 364–365TW-100.1 (Requirements for Specific Fluid Service), 364–365Figure TW-100.1, 364TW-130.3 (Weld Joint Categories), 364TW-130.4 (Weld Joint Efficiencies), 364Table TW-130.4, 364TW-130.5 (Rules on Weld Details, Shells and Flat Plates), 364Figure TW-130.5, 364Figure TW-130.5-1, 367Figure TW-130.5-2, 364Figure TW-130.5-3, 364TW-130.7 (Nozzle Attachment Rules), 364Table TW-134, 362TW-140 (Nozzle Attachment Rules), 364Article TW-1 (General Requirements for Tanks Fabricated by

Welding), 357–359pressure relief devices, 357–359reports and records, 357–359requirements, 357–359rules on design requirements, 360–364rules on materials requirements, 359–360stamping, 357, 359Table 53.1 (Design Load Factors for Normal Operations in

Specified Transportation Modes), 361Security seal, 340Seismic block, 14Seismic Design Review Guidelines, Japanese codes, 290–298Seismic design, 67, 104

reliability-based load and resistance factor design, 108Seismic Ordinance 515 (Seismic Design Standards for High-Pressure

Gas Equipment), 264–265Self-assessment, 107Semi-ferritic stainless steels

for industrial piping, French codes, 221, 223for pressure equipment, French codes, 235for pressure vessels, French codes, 203–204

Sensitization, postweld heat treatment of Alloy 600, 63SEP. See Sound engineering practice.SER. See Safety evaluation report.Service level(s), limits, deformation, and plastic instability, 241Service Level A, 45–46, 55, 60

evaluation procedures, 15–16French codes, piping, 245–246French pressure equipment, 243inclusion criteria for HSS snubbers, 106Service Limits, normal conditions, 346static loading, fracture toughness conditions, 51stress allowable limit, Japanese codes, 273–274

Service Level A/B, 14, 124–125structural factor, 118

Service Level B (Upset Condition), 45–46, 55, 60evaluation procedures, 15–16exclusion criteria for low-safety significant snubbers, 106required weld overlay thickness values from acceptance criterion,

20static loading, fracture toughness, 51stress allowable limit, Japanese codes, 273–274

Service Level C (Emergency Conditions)analysis procedures, 15–16inservice inspection of nuclear boilers and pressure vessels, 181vs. Japanese seismic operation states, 294static loading, fracture toughness and, 51weld overlay thickness criterion, 20

Service Level C/D, 124, 126structural factor, 118

Service Level Danalysis procedures, 15French pressure equipment, 242inservice inspection of nuclear boilers and pressure vessels, 181vs. Japanese seismic operation states, 294Service Limits, faulted conditions, 346static loading, fracture toughness and, 51stress allowable limit, Japanese codes, 273–274Service load, of operating PWR plant, 59

Set-in stub tube design, 12–13Severe Accident Management Guidelines, 100Severe Accident Safety, 89SF. See Structural factor.SGHWR. See Steam-generating heavy-water reactor.Shakedown, 124

of containment vessels for radioactive materials, 346 Shakedown factors, 316, 320Shakedown limit, 320, 323Shakedown loads, 320Shear stress, welds, PD 5500 (U.K.), 323Shell welds, reactor pressure vessel, 6-8Shielded metal arc welding (SMAW), 15, 17, 65, 117

upper shelf energy evaluation, 15–16Z factor value, 117–118Shielding, loss of, 336, 338

Shipper, 340–341Shipping containers

ferritic steel containment vessels, fracture toughness, 348–350for radioactive materials, fabrication criteria, 346–347welding criteria for fabrication, for radioactive materials, 356–357

Shipping paper, for radioactive materials, 340–341Ship Safety Law, 260Shot peening, 82Shroud, boiling water reactor internals, 13

core weld designations, 3flaw evaluation guideline, 1–4repair/replacement, 4support structure, 1–2, 24

SHT. See Solution heat treatment.Shutdown, 33, 63, 69, 100

PRA analysis, 104PRA Standards and, 107–108probabilistic risk assessment (PRA) and, 104, 106reducing dose rates, 82safe shutdown condition, 30, 99safety functions during, 104

Shutdown system, CANDU® 6 reactor, 165SI. See International System of units.Siemens, discontinuation of use of alloy 600, 64Simple pressure vessels, New Approach Directive, 145Simplified elastic-plastic analysis (Notification 501), 272–273SIN-TAP, 121SKI Report TR 89:20 (Research Project 87116), 121, 128Slenderness ratio, 262–267

Part TT (Testing Requirements) (continued)




Small bore piping, inservice inspection, 40Small-scale yielding (SSY) condition, 113, 119, 120–121

zone, 113, 120SMAW. See Shielded metal arc welding. SNCT. See Syndicat

National de la Chaudronnerie Tôlerie et Tuyauterie.SNM. See Special nuclear material.Snubbers

component classification, 100, 105evaluating piping, 19inservice testing, 103, 105supporting requirements, 106

Socket welds, 95–96Solution heat treatment (SHT), of stainless steel piping, 17Sound engineering practice (SEP), in PED, 310South African Koeberg 900 MWe plant, 193South Texas Project Nuclear Operating Company, 99South Texas Project (STP) Unit 1, 71–73, 83, 100Sparger design, 9–10Spark test, 365–366Special form, 340, 342, 344

capsule requirement, 344of licensed material, definition, 335of radionuclides, Type A package limits, 346

Special Metals Corporation, 63Special Permit No. 5000, for 6M package, 340Special Permit No. 5300 for 7A package, 340Special Permit No. 5400, 340Special Permit No. 5417, transportation of radioactively

contaminated items, 340Special treatment, of LSS items, 100–101Specification 2R, 341Specification 6L, 341Specification 6M, 340Specification 7 A, 340Specification 55, 340Specification packages, 334–335, 337, 342, 340–341Spent nuclear fuel, 433, 350

storage, risk analysis and security of, 110Spot radiography, pressure vessels, Japanese codes, 262Spring back, 397SR. See Safety related.SR. See Supporting requirements.SRM. See Staff Requirements Memorandum.SRP-LR. See United States Nuclear Regulatory Commission

Standard Review Plan for review of License RenewalApplications for nuclear power plants.

SSC. See Standards Steering Committee.SSC. See System, structure, or component supports.SSY. See Small-scale yielding.Staff Requirements Memorandum (SRM)

00-0117, 349–350SECY-98-300, 98–99Option 1 (Risk-Informed Changes on a Case-by-Case Basis), 98Option 2 (Risk-Informed Regulation Initiative), 98Option 3 (Direct Risk-Inform the Technical Requirements in

10CFR50), 98SECY-99-200, 350

Stainless steelsboiling water reactor piping cracks, 16butt welding, 65–66cladding, 12, 73, 83crack arrest and, 69

crack growth rate, 2, 23crack propagation due to fatigue, 69fracture toughness of irradiated, 3intergranular stress corrosion cracking, 1, 17L-Grade, SCC, 25piping, 1, 16–17, 116–117pressure equipment, EN 13445, 327for pressure equipment, Japanese codes, 263–265, 286for pressure equipment, PD 5500 (U.K.), 311–312,

327, 329for transport tanks, 361, 365weldability, French codes, 233–234

Stainless steels, specific types304, crack growth rate monitoring, 24304, fast breeder reactor material, 1–2, 12, 25, 27, 250304, intergranular stress corrosion cracking, 16304, nuclear grade, 17304, pipe, predicted instability loads, 116–117304, pressure equipment, PD 5500 (U.K.), 311304L, fast breeder reactor material, 1–2, 16, 250304LN, 1304LN, piping, 16316, fast breeder reactor material, 1, 250, 251316, nuclear grade, 17316, piping, 16–17316, pressure equipment, PD 5500 (U.K.), 311316L, fast breeder reactor material, 1, 4, 16, 28, 250316L(N), fast breeder reactor material, 1, 16, 250321, pressure equipment, PD 5500 (U.K.), 311347, pressure equipment, PD 5500 (U.K.), 311

Stakeholderdefinition, 380roll in decommissioning a nuclear facility, 380–382

Standards Council of Canada (SCC), 160–162Standards-developing organizations (SDOs), 109, 110

risk-informed safety classification, 161supporting risk-informed approach, 109

Standards Steering Committee (SSC), 160Startup/shutdown events, 9, 20–21Startup testing, of BWR jet pump, 5“State Plan” states, 425Static cycling test, bend of pipe, 297–298Static fracture toughness, 52–54Static loading, 51Station blackout, 30, 32Steady state secondary stress, 18Steam, as group 2 fluid, 133Steam boilers, 156

French codes, 191, 217–218. See also COVAP.Steam dryer

boiling water reactor internals, 1, 6–8, 24support bracket cracking, 14–15

Steam flow path, BWR, 6–7Steam-generating heavy-water reactor (SGHWR), fracture resistance

assessment, 121Steam generators, 82, 156

piping, maintenance program, 83Steam generator tubing, 39, 183

PWSCC cracks, 68, 73use of nickel alloys in, 63–64

Steel bolting, for pressure vessels, French codes, 201–202Steel containment vessel (SCV), Japanese codes, 288



738 • Index

Steel plates, ferrous and nonferrous material clad, for pressurevessels, French codes, 201–202, 206

Steels, for pressure vessels, Japanese codes, 262, 264Stiffeners, 313–314

light, 318Stiffener tripping, 314Stirrers, 141Stoomwezen, 329Stopple(tm) bypass repair method, pipeline systems, 405Storage tanks

French codes, 246Strain cycling, pressure vessels, PD 5500 (U.K.), 320Strain hardening, 114Strain-hardening exponent, 114Strain-rate, of local brittle zones, 53Strategic planning

economic analysis, 81–82for RPV head nozzle PWSCC, 81

Stray current corrosion, as pipeline failure mode, 407–408Stress

and alloy 600 susceptibility to PWSCC, 68alloy 600 resistance to, 68bending, 46, 116, 117–118, 121, 239cladding, 14, 24, 51, 55Code allowable, 113compressive, 67, 83–84, 246critical, 113flow, 113, 116–117, 399formula, 47French codes, 242, 246–247, 248–249hoop, 67, 315, 403improvement remedies for cracks, 17limits, 124, 273membrane, 45–49, 288, 316, 312–313, 319operating stresses above thresholds, 59on pipeline systems, 115, 117, 395–396pressure, 14, 47, 55pressure equipment, EN 13445, 327pressure equipment, French codes, 245pressure equipment, Japanese codes, 263–264, 269, 271–272, 330pressure vessels, PD 5500 (U.K.), 321–324primary, in weld overlay repair, 18primary water stress corrosion cracking, 68, 76residual, 14, 22, 24, 51, 84shrinkage, due to weld overlay, 18of supports, PD 5500 (U.K.), 319sustained, 22–23thermal, 9, 14, 23, 45–47, 55, 75, 124–125, 245, 287through wall, 47Stress analysis, 314industrial piping, French codes, 213, 224–225local loads on cylindrical shells, PD 5500 (U.K.), 320Stress concentration factorof containment vessels for radioactive materials, 346–347nozzle in a cylindrical vessel, 315–316PD 5500 (U.K.), 312pressure equipment, PD 5500 (U.K.), 321pressure vessels, PD 5500 (U.K.), 321–323spherical shells, PD 5500, 314–315

Stress corrosion cracking (SCC), 140as aging mechanism, 57, 59alloy 182 welds, 13

of austenitic stainless steel, in BWR plants, 63–64boiling water reactor, and fitness-for-service (Japanese), 277–279in boiling water reactor jet pumps, 5–6

Spanish Regulation in the Nonnuclear Industry, 563Spanish NDE Qualification Methodology, 569caustic, 63

chloride-induced, 63–64concrete casks, Japanese codes, 290controlled by hydro testing assessment, 386–387crack growth rate relationship, 2–4, 20, 22–24, 281crack initiation/growth, 24, 33environmentally-assisted, 24fatigue protection, nuclear power plants, 34fluoride-induced, 64intergranular, of piping, 66, 85, 233, 249, 499, 573, 684as pipeline failure mode, 374

Stress Corrosion Cracking in Pipelines (1996, Canada), Stress function method, 45Stress intensity, 20, 35

of containment vessels for radioactive materials, 345, 348limit, 273nuclear pressure vessels, PD 5500 (U.K.), 44, 127, 159

Stress intensity factor (KI), 4, 44, 76, 118calculating, 45, 46in circumferential crack growth predictions, 76, 77Code guidance, 18due to pressure, 45, 47industrial piping, French codes, 191, 553, 554maximum value, 51for membrane, 45pressurized water reactors, 45thermal, 45–49

Stress intensity parameter, nonlinear, 114Stress ratio, 118Stress reduction factor, 317Stress relaxation, 57, 59, 510

irradiation enhanced, 57, 59Stress risers, vibration stress at crack, 5Stress strain, 114, 119Strip yield model, 114Structural factor (SF) (safety factor), 118

boilers, French codes, 253CODAP (French codes), 208circular cylinder, 314, 490fatigue stress intensity factors, 6, 96, 125, 529, 597industrial piping, French codes, 142, 191, 553, 554PD 5500 (U.K.), 312pressure vessels, French codes, 224pressure vessel, PD 5500 (U.K.), 319

Structural failure probabilities, for piping systems ranking, 95Structural integrity treatment requirements, LSS safety-related items

for, 101Structural reliability model, risk evaluation, 95Structural reliability theory, 126Structural specifications, Canadian standards, 163Structural steel specifications, Canadian standards, 163Stub tube, 12, 13, 24Submerged arc welding (SAW), 15

failure mechanism in welds, 117Z factor value in welds, 117

Sulfur, 67Superheated water boilers, French codes, 191, 217. See also COVAP.




Super-heaters, 156Supervisory Control and Data Acquisition (SCADA), 393, 417, 420,

421Supporting requirements (SR), 92, 94

defining, 94PRA capability requirements for, 89

SupportsCANDU® nuclear power plants, 172CODAP future specifications, 207EN 13445 standard, 330French codes, 246Japanese codes, 258for pressure equipment, PD 5500 (U.K.), 319

Surface corrosion profiling, of pipeline systems, 325Surface examination, 72, 98

for detecting aging effects, 41, 58of reactor pressure vessel welds, 51

Surface treatmentFrench codes, 191As remedial measure for PWSCC, 63–68

Surry Power Station Unit 1, 100Surveillance program, French codes, 229Surveillance sample coupons, 43Sustained stress intensity factor, 23Swedish SKI Report TR 89:20 (Research Project 87116), 121Syndicat national de la Chaudronnerie, Tôlerie et Tuyauterie (SNCT)

(French organization of pressure vessel and pipingmanaufacturers association), 191–193, 195, 198, 209, 217,254

description and purpose, 191headquarters address, 193Web site, 193

Synthetic aperture radar (SAR) imaging, 418System, structure or component (SSC), 92, 162System-based code (SBC), using risk insights, 109System classification list, CANDU® nuclear power plant, 174System designer, CANDU® nuclear power plants, 172

Tadotsu Technical Test Center of NUPEC, 297Tangential shear stress, concrete containment vessels, 288Taylor Forge method, 317TCs. See Technical committees.Tearing instability, 121, 647Tearing moduli, 116Technical committees (TCs), 168

French codes, 191, 192Technical documentation, of pressure equipment, 137Technical Guidelines for Seismic Design of Nuclear Power Plant,

294, 295, 298Technical Standards and Safety Authority (TSSA), 168Technical Standards on Thermal Facilities for Electricity Generation,

270TEMA Standard, 262, 263Temperature

corrosion rate of hot concentrated aerated boric acid on hot low-alloy steel surface, 74

effect on PWSCC in hydrogen concentration variables, 67ferritic steel fracture toughness and, 53fluctuations in, 21lowering RPV head, 79lower range, fracture toughness and, 51potential for age-related degradation of internals, 57, 59

pressurized water reactor operation, 45of radioactive materials, packaging, 339, 348, 352of radioactive materials, restrictions, 337–338rate of PWSCC initiation and growth, 67reduction, as remedial measure for PWSCC, 82snubber service life and, 105–106stress intensity factor of safety, 45transients, effect on fatigue life, 35upper-shelf, 113

Temperature-dependent material properties, 15, 53Temperature indexing, 44Temperature monitoring devices, in pressure equipment, 154Temper-bead welding, ambient temperature, 19TENPES. See Thermal and Nuclear Power Engineering Society.Tensile strength, PED limit, 310–311Tensile stresses, 66–67, 288Tensile testing, 249

piping, Japanese codes, 297of pressure equipment, 157

10th International Conference on Nuclear Engineering (ICONE 10-22733), 98, 112

10-year in-service inspection outages, 72Terrorism, acts of, risk-informed analysis to counter, 110Testing, 90

boilers, French codes, 223–224, 236–237French codes, 249of motor-operated valves (MOVs), 105PD 5500 (U.K.), 319pressure equipment, EN 13445, 330pressure vessels, French codes, 206, 240pressure vessels, Japanese codes, 263–264

Testing programs, inservice testing (IST) program guidance, 104Test interval, 105Test pressure, of pressure equipment, PD 5500 (U.K.), 319T0 fracture toughness reference temperature, 43, 53, 54TGL Standards (Germany), 328Theoretical buckling pressure, 314Thermal activation energy model, 83Thermal and Nuclear Power Engineering Society (TENPES)

Committee for Environmental Fatigue Evaluation Guidelines, 21

Thermal and Nuclear Power Engineering Society (TENPES)Guideline, 275, 258

Thermal and Nuclear Power Engineering Society (TENPES) (Japan),273–274, 276–277, 291

Thermal embrittlement, as aging mechanism, 57, 59Thermal environments, of light-water reactors (LWR), 45Thermal expansion, 55, 63, 116Thermal fatigue, piping failure, 96Thermal fluid boilers, French codes, 181, 226Thermal fluid heaters, 169Thermal loading, 15, 122Thermal ratchetting, 346Thermal shock, fracture toughness requirements for protection

against pressurized, 30, 32Thermal sleeve bypass leakage detection system, 9Thermal stress(es), 14, 43, 45–47, 122–123, 245. See also Stresses.

of containment vessels for radioactive materials, 346, 347feedwater nozzle, 9Japanese codes, 287use in predicting crack growth rate, 76from vessel heatup/cooldown, 55



740 • Index

Thermal stress intensity factor (KIT), 47–49Thermal stress ratchet rule, 243Thermal test, 336Thermocouple nozzle, 63, 65Thermo-mechanically treated steels

for industrial piping, French codes, 202for pressure vessels, French codes, 203–205

Thiosulfate ion content, radiographic film processing qualityevaluation, 249

Third party damage (TPD) index, 382Thorium-231, 6513-D influence coefficients, 48Three-layer polyolefin coatings, for pipeline systems, 409–414Three Mile Island (TMI), 89Three Mile Island Unit 1 nuclear power plant, 813 Sm rule, 243Threshold level, distribution of times to occurrence at, 76Through-wall circumferential crack

calculation, 115–116stainless steel pipe, predicted instability loads, 116–118

Through-wall cracks, 3–4, 6, 9, 76–77, 117, 396pressure equipment, Japanese codes, 293

Through-wall stress, 47Through-wall temperature gradient, 17–18TI. See Transport Index.TIG. See Tungsten inert gas welding. Time-history analysis, stress analysis at crack location, 5–6Time-limited aging analysis (TLAA), 30–32, 34–35, 38, 41Titanium

for pressure equipment, PD 5500 (U.K.), 311for pressure vessels, Japanese codes, 265, 286

Titanium alloysCODAP future specifications, 208for pressure equipment, Japanese codes, 286

TLAA. See Time-limited aging analysis.TMI. See Three Mile Island. TOFD, 254Tokamak-type D-T facility (ITER) for fusion reaction, 291–293Tolerance specifications and pressure boundary standards, Canadian

standards, 163Ton Tanks, 358Torqued bolts, 59Torsional stress, piping, Japanese codes, 300Toughness

at crack extension onset, 118pressure equipment conformance, 147, 169

Toughness tests, of pressurized water reactor (PWR) vessels, 44Total applied stress intensity factor, 14Total displacement, 115–116“Towards the Performance-Based Technical Code of Nuclear

Facilities and Utilization of Voluntary Standards”, 259Tpc has launched a power, 685TPD. See Third party damage index.Traceability, 140

of pressure equipment components, 154Transient

accidents caused by, 89critical pressure, effect on fatigue life, 35critical temperature, effect on fatigue life, 35heatup/cooldown, 48operating, 35reduction of risk from anticipated, 31, 32

requirements for reduction of risk from, 33selection for evaluation, 15

Transient conditions, 14steam-dryer-support-bracket, 14

Transient events, 20–21, 51, 648Transient monitoring, effect on critical locations, 35Transient operation, 8, 9, 614Transient temperature, 53Transmission line, 170, 372, 373Transportable cylinders, Japanese codes, 261Transportation of Explosives and Other Dangerous Articles Act, 335,

337, 341Transportation Security Administration (TSA), 420, 421Transport Canada, regulating transportation of dangerous goods, 168Transport Groups, radioactive materials, 342Transport Index (TI), 351, 352Transport tanks, 357. See also Section XII (Transport Tank Code).Transport unit, 337Transverse flux, for pipeline system assessment, 391Transverse upper-shelf energy (USE), 15, 16, 121, 124, 318, 388To reference fracture toughness transition temperature, 44Trend curve prediction, for shifts in nil-ductility reference

temperature, 43Trending, 58Tresca yield criterion, 316Trigger-point temperature, 125Triple thermal sleeve design, 9Tripping, stiffener, 314, 471Tritium, total package limit, 172, 428, 437Trunk line, 372TRVP. See Trojan Reactor Vessel Package.TSA. See Transportation Security Administration.TSCA. See Toxic Substances Control Act.TS-R-1. See International Atomic Energy Agency, TS-R-1.TSSA. See Technical Standards and Safety Authority.“T” Stamp, 367Tsuruga-1 nuclear power plant, 13

stress-corrosion cracks in alloy 182 welds, 13–14Tsuruga-2 nuclear power plant, 70Tubes and tubing

Alloy 600, crack initiation in, 68Japanese codes, 44, 257–259steam generator, 63, 64

TubesheetEN 13445 standard, 325pressure equipment, PD 5500 (U.K.), 325

Tubesheet heat exchangers, French code design rules, 208Tuboscope, 387Tungsten inert gas (TIG) welding, 65“TV” mark, 359Two-parameter method, fracture evaluation of piping, Japanese

codes, 2812004 ICONE-12 conference, 84Type A(F) radioactive materials, Type A radioactive materials, quantities allowed in packaging, 334,

340, 342, 343Type A-Type B quantity provisions of IAEA regulations, 340Type B containers, 342, 354Type B(DP) dual-purpose packages, radioactive material, 339Type B fissile shipping containers, 339Type B(F) radioactive materials, Type B(M)F radioactive materials, 346




Type B(M) (Multilateral) packaging standards, 346Type B quantity, definition, 342Type B radioactive materials, quantities allowed in packaging, 334,

340, 342, 345, 355Type B(U)F radioactive materials, Type B(U) (Unilateral) packaging standards, 346, 348Type C packages, 350, 352, 353

U Certificate of Authorization, 264ULC. See Underwriters’ Laboratories of Canada.Ultimate strength, of pressure equipment, 157Ultimate stress, piping, 118Ultimate tensile strength (UTS)

French codes, PED limit, 310and plastic collapse of pipe, 387of pressure equipment, PD 5500 (U.K.), 311–312

Ultrasonic test (UT) (examination), 59, 72of BWR jet pump cracked weld, 6of BWR shrouds, 3calandria assembly, CANDU® nuclear power plants, 175Canadian standards, 170CANDU® nuclear power plant components, 183, 184, 188as condition monitoring program, 59and crack growth rate monitoring, 24of dissimilar metal weld overlays, 19feedwater nozzles, 9French codes, 191, 193, 196interval factor, 10in Japanese fitness-for-service code, 276–278of pipeline systems, 388–389, 390, 395, 400pressure vessels, Japanese codes, 262–263, 290of RPV head penetration nozzle, 72of RPV-to-shroud support plate weld, 13transport tanks, 366zirconium alloy components, 176–177

Uncracked ligament length, 115Underground cable, testing, 39Underwriters’ Laboratories of Canada (ULC), 160, 162Unfired vessels, 133–134Uniform Building Code, 461Unified Procedure,Uniform dose basis, 351United Kingdom (U.K.) See also British standards, specific types

unfired pressure vessel rules, 309–330United Kingdom Pipeline Regulator, 375United Kingdom’s Pipeline Safety Regulations, 375United Nations, 365

labeling systems for radioactive materials, 364United Nations Hazard Classifications

2.1 (flammable gas), 365–3662.3 (toxic gas), 365–3666.1 (toxic materials), 365–366

United Nations Recommendations on the Transport of DangerousGoods, Model Regulations, 357

United Nations Sub-Committee of Experts on the Transport ofDangerous Goods, 357

United States Atomic Energy Act of 1954, 341, 343United States Atomic Energy Commission (AEC), 339

Directorate of Licensing, 344Directorate of Regulatory Operations, 344Division of Materials Licensing, 342

manual, 342Reactor Safety Study, 89

United States Bureau of Statistics, pipeline incidents and propertydamage, 372

United States Coast Guard, 339United States Code, sections 552 and 553, 341, 343United States Competent Authority, 340, 350, 357for transport tanks, 357United States Department of Defense (DOD), 89United States Department of Energy (DOE), 425, 443, 447, 449, 450

decommissioning plan to remove radioactive material, 661nuclear waste disposal, 685

United States Department of Energy, Office of Civilian RadioactiveWaste Management (DOE/OCRWM), 349

United States Department of Health, Education and Welfare, 260United States Department of Homeland Security, 110United States Department of Labor, 260United States Department of Transportation, 357, 358

Hazardous Materials Regulations, 339–341, 357hazardous (including radioactive) material transportation, 350National Response Centerpipeline system assessment requirements, 390special permit, 340–343transport tank code, 357, 359

United States Department of Transportation Office of Pipeline Safety,property damage from oil pipeline incidents, 371

United States Department of Transportation/Pipeline and HazardousMaterial Safety Administration (USDOT/PHMSA), 357

United States Department of Transportations, Research and SpecialPrograms Administration (USDOT/RSPA), 357, 358

United States Hazardous Materials Regulations (HMR), 357United States National Environmental Policy Act of 1969, 30United States National Pipeline Mapping System initiative, 378United States Navy, refuel and defuel U.S. nuclear powered warships,

43, 483United States Nuclear Regulatory Commission (NRC) (USNRC), 3,

306, 440, 441, 444, 446, 447, 448, 450, 512, 514acceptable long-term repair, 19, 81acceptance of weld overlay repairs, 17, 19adoption of ASME Boiler and Pressure Vessel Code, 357Advanced Notice of Public Rulemaking (2000), 98aging management program (AMP), 21, 32allowable crack depth, 14AMP/AMR audits, 34approval of IST pilot programs, 103approval of weld overlay repair application, 19boiling water reactor flaw evaluation, 23boiling water reactor inspection, repair methods, 1, 3boiling water reactor RPV equivalent margin review summary, 16bounding crack growth rates for flaw evaluation, 22defining decommissioning, 471–485draft radiation embrittlement trend equations, 54evaluating crack length direction, 2evaluation of existing plant AMPs, 58Generic Aging Lessons Learned (GALL) Report, 21, 41inspection plans for PWSCC of alloy 600 base materials, 63, 68,

69, 83inspection program to manage effects of fatigue, 38inspection requirements for reactor pressure vessel (RPV) top head

nozzles, 72letter to STP Nuclear Operating Company, 100license renewal guideline updates, 41



742 • Index

license renewal process, 29, 30, 38Metrication Policy, 351nuclear industry risk-informed activities, 107pressurized thermal shock (PTS) reevaluation study, 56Proposed Rule for Public Comment (2003), 98regulatory process, 30risks from nuclear power, 89safety evaluation report (SER), 33special inspections, 662, 663standard review plans, 40Web site: www.nrc.gov, 42

United States Nuclear Regulatory Commission (NRC) (USNRC)Bulletins, 3, 441, 444

2003-02 (Leakage from Reactor Pressure Vessel Lower HeadPenetrations and Reactor Coolant Pressure BoundaryIntegrity), 73

2003-02, Temporary Instruction 2515/152 (Reactor PressureVessel Lower Head Penetration Nozzles), 73

United States Nuclear Regulatory Commission (NRC) (USNRC)Code of Federal Regulations (CFR)

10CFR, 489, 502, 682–68410CFR30 (30FR15748), 337, 33810CFR40 (Licensing of Source Material), 33510CFR40(28FR2111), 335

risk-inform plans, 107Appendix B, 98, 101, 683Appendix G (Fracture Toughness Requirements for Nuclear

Power Reactors), 44Appendix H (Reactor Vessel Surveillance Program

Requirements), 4410CFR50.48 (Fire Protection Regulations), 3110CFR50.49 (Environmental Qualifications), 3110CFR50.54(f) (Individual Plant Examination (IPE) for Severe

Accident Vulnerabilities), 3310CFR50.55 (Augmented Examination of Reactor Vessel), 710CFR50.61 (Pressurized Thermal Shock), 3110CFR50.62 (Anticipated Transients Without Scram), 3110CFR50.63 (Station Blackout), 3110CFR50.65 (Maintenance Rule), 30, 10210CFR50.69 (Risk-Informed Categorization and Treatment of

Structures, Systems, and Components for Nuclear PowerPlants), 98, 102, 107, 108, 110

10CFR51 (Environmental Protection Regulations), 3110CFR54 (Renewal of Operating Licenses for Nuclear Power

Plants), 58910CFR54.3, 3210CFR54.4, 3210CFR54.21, 35, 38, 5910CFR71 (Packaging of Radioactive Material for Transport), 343–34510CFR71 (1983 Final Rule), 34610CFR71 (1988 Proposed Rule, Major Changes),347, 34810CFR71 (1995 Final Rule), 34810CFR71 (2002 Proposed Rule), 349, 35010CFR71 (2004 Final Rule), 350, 355

Appendix B (Hypothetical Accident Conditions), 336, 341, 345,348, 354

exemptions, 341, 349fabrication criteria for shipping containers, 347Issue 15 (Change Authorization Issue), 354quality assurance requirements, 343–345

Subpart B, 336Subpart C, 336Subpart D, 336, 353Subpart H, 352–354Subpart I, 354

10CFR71 (31FR9941) (Packaging), 34110CFR71 (33FR750), 339, 341, 10CFR71 (33FR14918), 34110CFR71 (36FR22184), 34210CFR71 (38FR20482), 34410CFR71 (48FR35600) (Packing and Transportation of

Radioactive Material), 35010CFR71 (60FR50248), 35010CFR71 (61FR31169), 35110CFR71 (62FR5907) (Fissile Material Shipments), 341, 34910CFR71 (63FR8362), 34910CFR71 (64FR72633), 35010CFR71 (65FR44360), 349, 35010CFR71 (67FR21395-21396), 35110CFR71.31, 337, 33810CFR71.53, 338, 35410CFR71.61, 35210CFR71.63, 354, 35510CFR71.88, 35110CFR72 (Protection Against Radiation in the Shipment of

Irradiated Fuel Elements), 334, 33710CFR72 (26FR8982), 335, 33710CFR72 (28FR2142), 335, 33714CFR103 (Hazardous Materials Regulations Applicable to

14CFR103 (33FR750), 339, 34149CFR, 334, 335, 339–341, 351, 361, 366, 369, 390–39349CFR (60FR50291), 35049CFR78, 33549CFR170 (Rules of Procedure for the Hazardous Materials

Regulations Board), 338, 33949CFR170 (33FR750), 339, 34149CFR171 (33FR750), 339, 34149CFR172 (33FR750), 339, 34149CFR173, 33449CFR173 (33FR750), 339, 34149CFR173.393, 339, 34049CFR173.417, 35149CFR174, 33449CFR174 (33FR750), 339, 34149CFR175, 33449CFR175 (33FR750), 339, 34149CFR176, 33449CFR176 (33FR750), 339, 34149CFR177, 33449CFR177 (33FR750), 339, 34149CFR178, 33449CFR178 (33FR750), 339, 34149CFR178 (Revisions), 33949CFR179, 36149CFR179 (33FR750), 339, 34149CFR179.400-13, 36149CFR180 (33CFR750), 339, 34149CFR181-185 (33FR750), 339, 34149CFR186-190 (33FR750), 339, 34149CFR192 (Subpart O) (Pipeline Integrity Management), 37549CFR195 (Hazardous Liquid Pipeline Operators), 37549CFR195.428 (Pipeline System Assessment, SCADA), 393

United States Nuclear Regulatory Commission (NRC) (USNRC) (continued)




United States Nuclear Regulatory Commission (USNRC) DocumentControl Desk, 594

United States Nuclear Regulatory Commission (USNRC) DraftRegulatory Analysis (Draft RA), 441

United States Nuclear Regulatory Commission (USNRC) DraftRegulatory Guide, DG-1.121, 594

United States Nuclear Regulatory Commission (USNRC) FederalRegister (FR), 350

revisions of 10CFR71, 450United States Nuclear Regulatory Commission (USNRC) Generic

Letters (GL), 546GL 81-11, 9GL 88-20 (Individual Plant Examination (IPE) for Severe Accident

Vulnerabilities), 33GL 92-01, 15

United States Nuclear Regulatory Commission (USNRC)Information Notices, 30

United States Nuclear Regulatory Commission (USNRC) Issuance ofOrder, EA-03-009 (Establishing Interim InspectionRequirements for Reactor Pressure Vessel Heads atPressurized Water Reactors (PWR)), 72, 73

United States Nuclear Regulatory Commission (USNRC)Maintenance Rule, 94

plant expert panel, 96United States Nuclear Regulatory Commission (USNRC) Metrication

Policy, 351United States Nuclear Regulatory Commission (USNRC) Operations

Center, 355United States Nuclear Regulatory Commission (USNRC) Regulatory

Guides (NUREGs), 333, 355, 537NEI guidance document endorsement, 31NUREG draft (results of PTS reevaluation study), 56NUREG-XXXX (Fracture Analysis of Vessels—Oak Ridge

NUREG 1.124, 27NUREG 1.130, 247NUREG-0224 (Reactor Vessel Pressure Transient Protection for

Pressurized Water Reactors), 45NUREG-0313, 3, 17, 22, 684NUREG-0313, Revision 2, 684NUREG-0360 (Qualification Criteria to Certify a Package for Air

NUREG-0619, 9Generic Letter 81-11, 9Table 41.2 (Feedwater nozzle/sparger inspection

recommendations), 10NUREG-0744, 15, 121NUREG-1150, 90NUREG-1800 (Standard Review Plan for Review of License

Renewal Applications for Nuclear Power Plants), 57NUREG-1801 (Generic Aging Lessons Learned [GALL] Report),

58NUREG/CRs, 333, 346, 355NUREG/CR-1815 (Protection Against Brittle Fracture Failure in

Ferritic Steel Shipping Containers up to Four Inches Thick),346, 348

Table 1 (Radioactivity Limits for Each of Three Categories), 348NUREG/CR-3019 (Welding Criteria for Use in the Fabrication of

Shipping Containers for Radioactive Materials), 346–348NUREG/CR-3854 (Fabrication Criteria for Shipping Containers),

346–348NUREG/CR-5704 (Austenitic Stainless Steel, Environmental

Life), 21, 34, 37NUREG/CR-5999 (Interim Fatigue Curves), 34

NUREG/CR-6260 (Component Fatigue Life), 21, 34, 35, 38NUREG/CR-6583 (Carbon and Low Alloy Steel Environmental

Life), 21, 34, 37United States Nuclear Regulatory Commission (USNRC) Regulatory

Guides, 441risk-informed ISI and IST implementation, 90

RG 1.26, 99RG1.99, 124RG 1.147, 106, 594, 629RG 1.150, 52RG 1.161, 15, 124RG 1.174 (Using PR A in Risk-Informed Decisions on Plant-

Specific Basis), 8RG 1.175 (Plant-Specific Risk-Informed Decisionmaking:

Inservice Testing), 104RG 1.178 (Plant-Specific Risk-Informed Decisionmaking: ISI of

Piping), 94, 106, 571RG 1.192 (Operation and Maintenance Code Case Acceptability),

106, 594RG 1.200, 107

Appendix (Endorsement of NEI 00-02 Peer Review Process andSelf-Assessment Plans), 107

RG 7.4 (Leakage Tests on Packages for Shipment of Radioactivematerials), 335

RG 7.6, 345RG 7.6, Revision 1, 345RG 7.11 (Fracture Toughness of Base Material for Ferritic Steel

Shipping Cask Containment Vessel, Maximum 4 InchesThick), 348

RG 7.12 (Fracture Toughness Criteria), 348United States Nuclear Regulatory Commission (USNRC) Safety

Goals, 8United States Nuclear Regulatory Commission (USNRC) Special

United States Nuclear Regulatory Commission (USNRC)Standard Review Plan for Review of License Renewal (SRP-LR) applications, 41

Appendix A, 33Renewal Applications for Nuclear Power Plants, 57Section 2.1.3, 34Section 5.2.2, 45

United States Nuclear Regulatory Commission (USNRC) TACnumbers

TAC M89871/TAC M89493,United States Nuclear Regulatory Commission (USNRC)

Technology-Neutral Framework, 109United States Office of Pipeline Safety, 371, 374, 374Unresolved Safety Issue (USI) A-11 (USIA-11), 121

low-temperature overpressure (LTOP), 8, 43, 45, 49, 50USI-A26, 45Upper-shelf Charpy energy, 121Upper-shelf energy (USE), 15

end of life values, 55Upper-shelf operation, 121–124Upper-shelf temperature, 113, 124–126Upper-shelf trigger temperature, 125Uranium

enriched, 340natural, 163, 436, 437, 635natural and depleted, 437unirradiated, 441

Uranium-233, 438Uranium-235, 351, 437, 438, 440, 474



744 • Index

plus plutonium, 337, 343–345Uranium fuel pellets, 478Uranium hexafluoride, 350, 351Uranium isotopes, 438Uranium metal, 351Uranyl nitrate solutions, 436URN 99/1147 (DTI Guidance Booklet on PED Requirements), 309Usage factor, French codes, 245USDOT/PHMSA. See United States Department of

Transportation/Pipeline and Hazardous Material SafetyAdministration.

USDOT/RSPA. See United States Department of Transportation,Research and Special Programs Administration.

USE. See Upper-shelf energy.USI. See Unresolved Safety Issue.“U” Stamp, 367UT. See Ultrasonic test (UT) (examination).UTS. See Ultimate tensile strength.U-tubes, Japanese codes, 263“UV” mark, 359V. See Volume, internal, of chamber.Vacuum box testing, CANDU® nuclear power plant components, 181Vaccum vessel (VV), Japanese codes, 292Valve(s)

inservice testing (IST), 99pressure-temperature rating, 246probabilistic methods in qualification standards, 109probabilistic risk assessment for, 519risk-informed IST application, 104in scope of PED, 129

Valve design rules, French codes, 246V.C. Summer nuclear power plant, 69

alloy 82/182 butt weld axial crack leakage, 70, 71VDEs. See Vertical displacement events.Vertical displacement events (VDEs), 292, 293Vessel attachment, weld cracking, 13–15Vessel design life, 15, 45Vessel-to-shroud support, weld cracking, 13Vibration

of snubbers, 106startup, 5stress range, steps in calculating, 5

Vibration tests, piping, seismic influences in Japan, 295, 296Visible spectrum and in the very near infrared (VNIR), 418Visual examination/testing (Examination Level: VT-1, VT-2, VT-3),

72. See also VT-1 examination; VT-2 examination; VT-3examination

alloys 82/182 butt weld leakage, 70bare metal, 71–73bare metal for PWSCC, 71, 75bare metal of BMI nozzles, 71bare metal of RPV head surface, 71of BWR shrouds, 3CANDU® nuclear power plant components, 163of crack repair, 15to detect aging effects, 58enhanced, as aging management strategy, 60French codes, 250joint coefficients allowed, 139of low-safety-significant (LSS) pipe segments, 90, 96, 100NRC requirements, 11as Section XI provision, 103

of sparger, 9, 10steam generator tubes, of vessel-to-shroud support weld cracking, 13zirconium alloy components, 176–177

VNIR. See Visible spectrum and in the very near infrared. Voidswelling, irradiation-induced

Volumetric examination, 10of BMI nozzles, 71to detect aging effects, 58inservice use and, 56of partial penetration nozzles, 72of permanent joints, 140of reactor pressure vessel (RPV) shell welds, 7of reactor pressure vessel welds, risk category and, 97

VR stamp, 359VT-1 examination

as aging management strategy, 59–60of BWR steam dryer, 6character recognition height, 60enhanced, 59–60of inner radii surface of nozzles, 10maximum direct examination distance, 60of welds in beltline region, 72

VT-2 examinationas aging management strategy, 59–60of low-safety-significant (LSS) piping segments, 96of reactor vessel pressure-retaining boundary during the system

leak test, 72VT-3 examination

as aging management strategy, 59-60enhanced, 60maximum direct examination distance, 60of welds outside the beltline region, 72

VV. See Vacuum vessel.WASH-1400 study, 89WASRD. See Waste Acceptance System Requirements Document.Waste Acceptance System Requirements Document (WASRD),Waste disposal containers, Waste-heat boilers, 156Waste incineration boilers, 156Waste Isolation Pilot Plant Land Withdrawal Act, Water chemistry

changes affecting PWSCC rate, 82to mitigate piping internal corrosion, 59

Water environment, 2–3alloy 600 corrosion resistance in high temperature, 63austenitic stainless steels fatigue crack growth rate, 21–22effects on reduction of fatigue life of light-water reactor

components, 21ferritic steels fatigue crack growth rate, 21high-temperature primary, alloy 600 SCC in, 64high-temperature pure, alloy 600 SCC in, 64intergranular stress corrosion cracking of stub tube, 12

Water gauges, visibility, 169Water hammer, and piping failure, 96Water heaters, Canadian standards, 169–170Water-jet conditioning, 84Water-moderated reactors, long-term operation safety aspects, 42Watertube boilers, French codes, 216–217. See also COVAP.WBS. See Work Breakdown Structure.Weibull statistical distribution, 53, 79




in predicting initiation of PWSCC in pressurized water reactors(PWRs), 74–76

Welds(s). See also Weld metals, specific types.attachment, 72axial, 8base metal, 52BMI, 72boric acid leakage, 69–71butt, alloy (82/182), 65butt, outlet nozzle, through wall axial crack, 69–70Canadian standards, 170circumferential, 7, 8, 48CRDM, 74crevice, 17defects in pipeline systems, 400dissimilar metal, 19, 72, 83dissimilar metal, flaw evaluation, 277–279fitness-for-service flaw evaluation, Japanese, 276–278flat ends and covers, PD 550 (U.K.), 317–318full penetration, 52, 66full penetration double bevel, generic J-integral fracture resistance curve equation constants, 124girth, 17, 49on high-fatigue lines, 38hot-leg, 69impact testing, 144inservice inspection of HSS segments, inspection in pressure coils exposed to direct radiant heat,

Canadian standards, 168J-groove, 65, 69joint coefficients, 139joint coefficients, industrial piping, French codes, 218joint efficiency, pressure vessels, French codes, 201joint factors, U.K. rules, 309joints, 139–140joints, PD 5500 (U.K.), 318–319joints, pressure equipment, Japanese codes, 263, 264, 282, 290laser repair, 83local brittle zone, 52–53longitudinal, 48–49nozzle-to-safe end socket, 72permissible joints in pressure vessels, French codes, 210–211pipe-to-nozzle, 17pipe-to-pipe, 17postweld heat treatment requirements, 19preparations, 65pressure boundary, 55in pressure equipment, 157pressurizer, 73repair, 13, 17–18, 45, 74repair, cracking of, 78–79residual stress, 16–17, 51RI-ISI of, 94safe-end, 17, 24shell, 6–8socket, 96stresses, PD 5500 (U.K.), 322–323stub tube, 24thickness measurement, 49toughness levels, 50vertical, 8vessel attachment cracking, 13

vessel-to-shroud support cracking, 13–14water-jet conditioning of, 82

WeldabilityFrench codes, stainless steels, 233

Weld-deposited hardfacing, French NF M 64-100 standard, 248Welded joint(s)

defects during construction, 102efficiency, pressure vessels, French codes, 203fracture evaluation, Japanese codes, 281French codes, 252In pressure vessels, design of, 65Transport tanks, 364

Welded joint coefficientEN 13445 vs. PD 5500, 309, 310, 312French codes, 252industrial piping, French codes, 212

Welded structures, fracture analysis, 114Weld efficiency factor, French codes, 246Welding, 17–20. See also Gas tungsten arc welding; Shielded metal

arc welding; Tungsten inert gas welding.ambient temperature temper-bead, 19code compliance, 80cold temperatures and, 17criteria for fabrication of shipping containers for radioactive

materials, 347distortion avoidance, 65electroslag, 15field, 17filler metals, 14French codes, 252gas tungsten arc welding, 18oil and gas pipeline systems, Canadian, 170–171postweld heat treating, 19preheating, 19of pressure equipment, PD 5500 (U.K.), 312pressure vessels and piping, Japanese codes, 262, 268, 281repair/replacement, 110residual stresses, 22, 77shielded metal arc welding, 65submerged arc welding, 15temper-bead, 19transport tanks, 358, 359, 365, 366, 368of weld overlay repairs, 17–20, 81

Welding consumables, for pressure equipment, 143, 144, 201Welding Data Package, French codes, 247Welding Procedure Qualification Test (WPQT), 262, 263, 264Welding procedures, registration, Canadian, 172Welding Research Council (WRC), 308

Bulletin 175, 45Bulletin 404, Bulletin 413, 15, 122, 123

Welding specifications, Canadian standards, 162, 163Welding Specifications, W series, 162, 163Weld joint efficiency, 201, 203, 208, 212, 222, 290, 363

boilers, French codes, 212, 234pressure vessels, Japanese codes, 262

Weldments, piping, carbon steel, circumferential flaws, 119Weld metal

cladding with duplex, 17crack growth data, 21, 76in dissimilar metal weld overlay, 19requirements for weld overlay repairs, 18



746 • Index

Weld metals, specific typesalloy 52, 81, 82

in PWSCC-resistant repairs, 82resistance to PWSCC, 81

alloy 82, 63, 65, 66, 67, 70–74, 76, 77, bare metal visual inspection of butt welds, 71butt weld leak, 69chromium concentration, 66composition, 65crack growth behavior in PWSCC, 77crack growth rate testing, 76crack initiation behavior in PWSCC, 79inspection methods/requirements, 71–72location in PWR Vessel, 64primary water stress corrosion cracking of, 63PWSCC cracks in CRDM nozzles, 70PWSCC cracks in inlet/outlet nozzle butt welds, 69–70PWSCC in, 66, 69uses, 64–66weld overlay repair, 81

alloy 152, 65resistance to PWSCC, 67

alloy 182, 17, 65, 66, 69, 76, 77bare metal visual inspection of butt welds, 73butt weld leakage, 69chromium concentration, 66composition, 65crack growth behavior in PWSCC, 77crack growth rate testing, 76crack initiation behavior in PWSCC, 79inspection methods/requirements, 71–72location in PWR Vessel, 64primary water stress corrosion cracking (PWSCC) of, 63PWSCC cracks in CRDM nozzles, 70PWSCC cracks in inlet/outlet nozzle butt joints, 69–70PWSCC in, 66, 69uses, 64–66visual inspection, 71weld overlay repair, 81

Weld overlay repair (WOR), 17–20, 25, 81Weld replacement, as a PWSCC repair, 81–82Weld shrinkage, 67Westinghouse, 193Westinghouse designed PWR power plants, 456

bottom-mounted instrument (BMI) nozzle, 65CRDM nozzles in, 65

use of alloy 82/182 butt welds, 65Westinghouse Owners Group (WOG), 68

WCAP-14572 Rev. 1-NP-A (Topical Report Applying Risk-Informed Methods to Piping ISI), 97, 572

Method A application, 97WGM. See Working Group Materials.WGP. See Working Group Pressure.WIPP. See Waste Isolation Pilot Plant.WOG. See Westinghouse Owner’s Group.WOL. See Weld overlay.WOR. See Weld overlay repairs.Working Group Materials (WGM), 142Working Group Pressure (WGP), 131, 143, 144

Guideline 7/17, 143Guideline 7/24, 143

World Health Organization, 365WPQT. See Welding Procedure Qualification Test.WRC. See Welding Research Council.WTO/TBT Agreement, 257, 259, 260Yield (plastic collapse), as pipeline failure mode, 374Yield strength

French codes, 252and irradiated stainless steel fracture toughness, 3of nuclear pressure vessel steels, 3of nuclear reactor vessels, 66PED limit, 310of pressure equipment, PD 5500 (U.K.), 311–312PWSCC susceptibility and, 67, 68of steels for pressure vessels, French codes, 202

Yield stress, piping, 118Young’s modulus, 115, 123, 139, 304, 314, 448Z factors, 117–120

ferritic piping, 118–120for shielded metal arc welds, 65, 117for submerged arc welds, 117

Zick method, 330Zinc

addition to primary coolant, 84, 118addition to reactor coolant, 104, 118

Zircaloy, for liquid-injection shutdown system nozzles, 175Zircaloy-2, for calandria tubes, 636, 642Zirconium alloys

for CANDU® components, 172, 177for fuel channel pressure tubes, 174, 177, 182, 183for pressure tubes, 164, 183, 652



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