ML20033E005

Technical Evaluation Rept First Interval Inservice Insp Program Sequoyah Nuclear Station,Unit 2

ML20033E005
Person / Time
Site:	Sequoyah
Issue date:	07/31/1989
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20033E006	List: ML20033E005 ML20034B772 ML20034B773
References
CON-FIN-D-2109, CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-89-1473, NUDOCS 8907210033
Download: ML20033E005 (86)

Text

{{#Wiki_filter:-, ..SAIC-89/1473 = TECHNICAL EVALUATION REPORT-FIRST INTERVAL INSERVICE INSPECTION PROGRAM SEQUOYAH NUCLEAR STATION UNIT 2-Submitted to U.S. Nuclear Regulatory Commission. Contract No. 03-87-029 Submitted by Science. Applications International Corporation' Idaho Falls, Idaho 83402 July 1989 T 6Hp7210$33k 97f \\ Yk.k, Q _/ M Etnp6ovee Owned Company ' ^

F CONTENTSi ?y

3. -INTRODUCTION..'...................

3-2. EVALUATION OF INSERVICE INSPECTION PLAN.... /....... '3 t -2,1 Introduction..........:.....:........ 3 r 3 } 2.2 Documents' Evaluated I 2.3 Summary of Requirements .........-...J... .3 ' 4 2.3.1 l Code Requirements................. 4' l 2.3.1.1 _ Class;l Requirements............. 4 2.3.1.2 Class 2 Requirements- _........... 5-t 2.3.1.3 Class 3 Requirements......._.... 5' 2.3.2 Preservice Inspection. Commitments............. 5, 2.4 Compli ance with Requirements. -..........._. _... 16 2.4.1 Applicable Code Edition............... 6 2.4.2 Code Requirements..........,...... 6 2.4.3 License Conditions 7 2.5 Conclusions and Recommendations ..........x. 7 3. EVALUATION OF RELIEF REQUESTS 8 3.1 Class 1 Components.............. _ 9 3.1.1' Reactor Vessel 9 3.1.1.1 Relief Request-ISI-5,-Reactor Vessel Bottom Head Circumferential Weld, Category B-A, Item Bl.21........... 9-3.1.1.2 Relief Request.ISI-10, Reactor Vessel. Flange-to Upper Shell-Weld, Category B-A, Item Bl.30 12 3.1.2 Pressurizer (no relief requests) 1 3.1.3 Steam Generators and Heat Exchangers 14 3.1.3.1 Relief Request ISI-6, Steam Generator Nozzle Inside Radius Section, Category B-0, Item B3.140......... 14 i . _,,,. - _ _ + s

n 3.1.4 Piping Pressure Boundary...... "....... #. 17 3.1.4.1 Relief Request '151-3, - Pressure-Retaining.- l Dissimilar Metal Welds'~ 1n Piping,, Category B.F, Item B5.50 =.... 37 3.'I. 4. 2 Relief Request 151-7,- Reactor Coolant Loop Piping Welds, Category B-J, Item B9.10...:... a. ...a 19 3.1.4.3 Relief Request 151-3,-- Pressure-Retaining Welds in Piping, Category'B J,. Items B9.10, B9.20, and B9.30..'..... 21 1 3.1.5 Pump Pressure Boundary 24. 3.1.5.1 Relief Request 151-8, Pressure-~ Retaining Welds on Pump Casings, Category B-L-1, Item B12.10.....-... -24 3.1.5.2 Relief Request ISI-1, Pump Internal Pressure Boundary-Surface, l Category B-L 2,-Item B12.20-........-. 26 3.1.6 Valve Pressure Boundary...... -. _........ 29 i 3.1.6.1 Relief Request ISI-2, Valve. Internal Pressure Boundary Surface, Category B M-2,' Item B12.40........ 29 3.2 Class 2 Components.................... 32 3.2.1 Pressure Vessels and Heat Exchangers 32 3.2.1.1 Relief Request ISI-4, Steam Generator 4 Class 2 Circumferential Shell-Welds, ) Category C-A, Item C1.10 32 i-3.2.1.2 Relief Request ISI-12, Excess Letdown Heat Exchanger and Regenerative Heat Exchanger Shell Welds and Integrally Welded Attachments, Categories C-A, C-C, and C-E, Items C1.10,.C1'.20,- C1.30, and C3;10 34 3.2.1.3 Relief Request ISI-13, Residual Heat Removal Heat _ Exchanger Nozzle-to-Vessel Welds, Category C-B, Item C2.20...... 38 i i b ii 'i 1

i ~ 3.2.2 Piping Pressure Boundary (no relief requests); 3.2.3 Pump Pressure Boundary' (no relief requests) 3.2.4 : -Valve Pressure' Boundary (no-relief requests)^- i

i 3.3 ^ Class 3 Components- (no relief requests);

3.4 Pressure Tests _(no relief requests)' 3.5 General .e. 41. 3.5.1 Examination Method <....,,............ 41..- 3.5.1.1. Relief Request ISI-9, Ultrasonic Calibration. Block.for~Ferritic-Material Less-than Two. Inches Thick'.....-

41' 3.5.2 Examination Scheduling 43.

3.5.1.2' Relief Request 151-14, Scheduling-Requirements of' Tables IWB-2412-1 and IWC-2412 1 for. Examination of Class 1 and 2-Piping Welds, Piping-Supports, and Major Component. Supports 43 4. REFERENCES................,,............ 45 APPENDIX A:

SUMMARY

OF-REQUIREMENTS OF ASME BOILER AND PRESSURE = VESSEL CODE, 1977 EDITION WITH ADDENDA THROUGH l SUMMER 1978 i 4 l -i iii a

-l TECHNICAL EVALUATION REPORT j FIRST INTERVAL INSERVICE' INSPECTION PROGRAM' j i Seouoyah Nuclear-Station Unit 2 j l 1. INTRODUCTION Section 50.55a of 10 CFR Part 50 defines the requirements for the Inservice Inspection (ISI) Program for light-water cooled nuclear. power-facilities. Incorporated by reference in this regulation'is Section XI of the Boiler and Pressure Vessel Code published ~ by the American Society of Mechanical Engineers (ASME), which provides the basis for' implementing inservice inspection.* - Two types of 1nspections are required: (1) a preservice' inspection conducted before commercial operation to establish a baseline and _(2) peri-odic inservice inspections conducted during 10-year inspection intervals that normally start from the date of commercial operation. Separate plans for' completing preservice-inspection and-each 10-year inservice inspection must be. formulated and submitted to the Nuclear Regulatory Commission (NRC). The plan for each 10-year interval must be submitted at least six months before-the start of the interval. .During the initial 10-year, interval, inservice inspection examinations must comply with the requirements in the latest edition and-addenda of Section XI incorporated in the regulation.on the date 12 months before the date of-issuance of the operating license. The first interval program.for Sequoyah Unit 2 has been written to the 1977 Edition with addenda through Summer 1978. Sequoyah Unit 2 began commercial operation June 1, 1982. ~0n August 21, 1985, i Sequoyah 2 went off-line and stayed off-line until May 13,-1988.- In accor-1 dance with IWA-2400(c), the first interval has-been extended by_996 days. The l first interval therefore runs from June 1, 1982, through February 21,.1995. 4 Section 2 of this report evaluates the first interval.ISI Plan. developed i by the licensee, Tennessee Valley Authority (TVA), for Sequoyah Unit' 2 for (a) compliance with this edition of Section XI, (b) compliance with ISI-related commitments identified during the NRC's review before granting an i Operating License, (c) acceptability of examination sample, and,(d) exclusion criteria. j Based on the date Sequoyah's construction = permit was issued'(May 27, i 1970), the plant's components (including supports).1 hall meet the required inservice examinations and tests except design and access-provisions and preservice examination requirements, to the extent practical, set forth in Section XI of editions adopted by referer.ce.in Paragraph (b) of,10 CFR 50.55a. 4

• Specific inservice test programs for pumps and valves (IST programs) are being evaluated in other reports.

o I n:

~ ~ L. 1 Paragraph 30 CFR 50.55a(g) recognizes that-some-requirement's of the

current edition and addenda of Section XI may not be practical to implement because of limitations of design, geometry, and materials of construction of components and systems-that were designed to the older Code. -The regulation therefore permits exceptions to impractical examination or testing require-1 ments of the current Code. to be requested.

Relief from these requirements .may be granted, provided the health-and safety tf the public are not : endangered, giving due consideration to the burden placed on the licensee if the reciuirements were imposed. Section 3.of this report evaluates requests 1 for relief dealing with inservice examinations of components and with system i pressure tests. The regulation also provides that ISI' programs may meet the requirements i of subsequent Section XI editions and addenda, incorporated by-reference in. the Regulation, subject to approval by the NRC. - Portions of such editions or- ~ addenda may be used, provided.all related requirements: of the respective editions or addenda are met. These instances are addressed in Sections 2.4^ and 3 of this report. Likewise,.Section XI provides that certain components and systems may be exempted from volumetric and surface requirements. In some instances, however, these exemptions.are not acceptable to the-NRC or are. acceptable only with restrictions. The Preservice Inspection-(PSI) Program for _Sequoyah Unit 2, a. 4-loop Westinghouse pressurized water' reactor (PWR)., was not required by the Code of Federal Regulations, based on a construction permit date of May 27, 1970. However, TVA performed 1 preservice inspection,.as stated in the ISI. program, based on the 1974 Edition of the Code with. addenda through ummer 1975. Letters concerning PSI and ISI were issyqq June 24, 1977,5 ) July 8, 1977,(2). 3 December 8, 1978,l ) February S,(7) April 24,1980 d May 5,1980,14,-1979 )01 April 19, yL / y 31 1980 t ])(6). 9, May 22, 19 December 11, May27;982,gL 1 October (Lg) July 301982,(12) DecegkSeptember23,1982,4) December) 1983,lI4,' April 12

L984,

,- 19 Octob r ;5, 1984,(18? Novemp )14, 1986,5 g) July 30, 1986,]40)1 January 30, 1987, 21/ and May 11, 1988. Staff evaluations' of initial PSI and ISI submittals were made in Safety Evaluat Suppl 9rtsts1and2,datedMarch1979,jgg) Reports,(SERs),(g8EG-00llwith: l M* February 1980, ) and August 1980.l' 1982,gs1and2ISIpro 10,1986,dE;.983,gsweresubmittej2gSeptember. August 2) March.7

L984, 4,

18,-1985,(33) July 30,1987,dl21,1989,*gg) November 9,1988 December E November and March the NRC requested additional informationrequiredtocompleteitsreview9{5the ISI prograrp inforrpgi and in later revisions of the ISI plan.1May6,1986,d on was furnished on March 12, 1986,t May 5, L 1989,l t The thirteen outstanding relief requests are' evaluated in Section 3 of'this report. ~ l 2 1

.= H c 2. EVALUATION OF INSERVICE INSPECTION PLAN ) 2.1 Introduction \\ l The approach being taken in this evaluation-is to review the applicable program documents to determine the adequacy of their response to Code-requirements and any license conditions pertinent to ISI activities. The' rest of this section describes the submittals reviewed, the basic requirements: of the effective Code, and the appropriate license conditions. The results of the review are then described. Finally,- conclusions and recommendations are given. 2.2 Dpeuments Evaluated l A chronology of documents on Sequoyah Unit 2 PSI.and ISI is given-l in'Section 1 of this report. Those documents that impact this--ISI program l evaluation are (1) the latest revisions of the ynits 1 and 2 ISI Programs 9, 1988,92 iggter, (2) portions,of-l which were~ attached to the November f theSequoyahFinalSafetyAnalysis(geport(FSAR),W/(3)portionsof Supplement 2 to the SER.(SSER #2), DJ (4) the relief requests,.and (5) to a lesser extent, submittals on-PSI. l l_ 2.3 Summary of Reduirements 1 l The requirements on which this review'is focused include the following: (1) Comoliance with ADolicable Code Editions..The Inservice Inspection Program shall be based on the Code editions defined in 10 CFR-50.55a(g)(4) and 10 CFR 50.55a(b). The 1icensee for_Sequoyah. Units 1 and 2 has written the first interval program to the 1977 Edition with addenda through Summer 1978.- These Code requirements are summarized in Section 2.3.1 and detailed Code requirements are given in Appendix A. The 1974 Edition' Summer 1975 Addenda.is-l required for selecting Class 2 welds in systems providing the 1 functions of residual heat removal,' emergency core cooling, and containment heat removal. This.is a requirement of 10 CFR-50.55a(b)(2)(iv)(A). The 1974, Summer 1975 Code may also be i used to-select Class 1 and 2 piping welds as allowed by 10 CFR 50.55a(b)(2)(ii) and 10 CFR 50.55a(b)(2)(iv)(B). (2) Accentability of the Examination Samole. Inservice volumetric, i surface, and visual examinations shall be performed on' ASME Code Class 1 and 2 components and their supports using sampling schedules. l described in Section XI of the ASME Code and 10 CFR 50.55a(b). Sample sizg designations are identified as part-of'the Code requirements given in Appendix A. g (3) Exclusi6n Criteria. The criteria used to exclude components from examination shall be consistent with IWB-1220, IWC-1220, and i 10 CFR 50.55a(b). 3 ~ " -

l l (4) PSI Commitments. The'3nservice Inspection' Program should address? q all license conditions, qualified acceptance conditions, or other. 1 ISI-related commitments described in-_the-Safety Evaluation Report-and its supplements for. the preservice examination.- j i 2.3.1~ Code Requirements I . The following requirements are summarized;from the 1977 Edition o'f Section XI with addenda through Summer 1978. Many requirements call forI the examination of all areas, while other requirements specify more limited examinations based on criteria such as representative percentage, components-examined under other categories, material. thickness.. location relative to other welds or discontinuities, and component: function and construction.-Those l -components with examination requirements based on the 1974-Edition of Section XI with addenda through Summer 1975 are marked below with an asterisk. For detailed requirements, see Appendix A of this report-or the Code itself, t 2.3.1.1 Class 1 Reauirements.. The following Class 1-components are to be examined in the first interval in accordance~with' Table IWB-2500-1: (1) Pressure Retaining Welds in Reactor Vessel ~ (2) Pressure Retaining Welds-in Vessels Other than Reactor Vessels-(3) Full Penetration Welds of Nozzles in Vessels-(4) Pressure Retaining Partial Penetration Welds in Vessels (5) Pressure Retaining Dissimilar Metal Welds (6) Pressure Retaining ~ Bolting, Larger. than 2 in, in Diameter - (7) Pressure Retaining Bolting, 2-in. and Smaller in Diameter (8) Vessel Supports- + (9) Pressure Retaining Welds in Piping * (10) Integral Support Members for Piping, Pumps, and Valves (11) Component Supports for Piping, Pumps, and Valves (12) Pump Casings and Valve Bodies, including Pressure Retaining Welds (13) Interior of Reactor Vessel,-including Core Support Structures, Interior Attachments, and Removable Core Support' Structures ~ (14) Pressure Retaining Welds in Control Rod Housings (15) All Pressure Retaining Components - Pressure Tests - (16) Steam Generator Tubing i i 4 4 L- ~ -.

] 2.3.1.2 Class 2 Raouirements. - The following Class 1 components are to l c be examined in the first interva' in accordance with Table IWC-2500-1*' 1 .(1) Pressure Retaining. Welds:in Pressure ' Vessels t (2) Pressure, Retaining Nozzle Welds in Vessels j (3) Support Members P (4) Pressure Retaining Bolting Ex:eeding' 2 in.- in Diameter (5) Pressure Retaining Welds in Piping * -t (6) Pressure Retaining Welds in Pumps and Valves-i -(7) All Pressure' Retaining Components - Pressure Tests-2,3,1.3. Class 3 Raouirements. For systems or portions thereof required i to operate in support of the below described safety functions.,the following. components are to be examined in accordance with Table IWD-2500-1: i q (1) To support normal plant safety functions: (a) Pressure retaining components (b) Component supports.and restraints- '(c) Mechanical and. hydraulic snubbers for components greater than 4 in, nominal pipe size i-(2) To support post-accident safety functions: (a) Pressure retaining components (b) Component supports and restraints i. (c) Mechanical-and hydraulic. snubbers for components greater than i 4 in, nominal pipe size-I i (3) To support residual heat removal from spent fuel storage pool: f 1 (a) Pressure retaining components (b) Component supports and restraints (c) Mechanical. and hydraulic snubbers for components greater than 4 in, nominal pipe size 1 4 2.3.2 Preservice Inspection Commitments No preservice, inspection commitments were identified for Unit-2. i + i 5 a w .n.e --*-d--- 4-e-w a-s w--- --. ------- As

j .s 3 2.4.Como11ance'with Reouirements-2.4.1 Applicable Code Edition 5 i Tte initial irservice inspection interval examination program must - comply (10 CFR 50.55a(g){4)(i)) with the requirements of the latest edition-1 and addenda:of Section Xr incorporated into 10 CFR 50.55a on the date 12 c. months before the date of issuance of the operating license.. Unit 2. received. its-operating license September 15, 1981, and thus, the applicable Code-is the 1977 Edition with Addenda through Summer'1978, which became effective = November 1,_1979. As allowed by'10 CFR Part'50.55a i chosen to select Class 1 and 2 piping welds-for exam (b)(2), the itcensee has < ination in'accordance with the 1974. Edition with addenda through Summer,1975'(Examination Categories B-J,' C-F,-andC-G). The licensee receiv NRC on January 25,1988.g9gpproval for the use'of Code' Case N-356 from' the / 4 2.4.2 Code Requirements 1 The First Interval ISI Program.of record-(exclusive of testing) is_ contained in TVA procedure SI-114.2 for Unit 2.( ytpp and valve 'i This i procedure is intended to supersede all previous ISI program procedures. The submitted ISI program was reviewed, and the following observations-were noted. The Inse'rvice Inspection Program for Sequoyah Unit 2. identifies approp-riate Code classes for.each component of the power plant. The design of the Code Class-I components of_ the reactor coolant pressure boundary in Sequoyah Unit 2 incorporates provisions for access-for inservice examination in accordance with Section XI of the.ASME. Code. Examination instructions and procedures, including diagrams or-system drawings identifying the extent of areas of components subject to examina-tion, have been prepared. They are listed in the ISI program component i tables, cross-referenced to weld.and' hanger isometrics and component identification drawings, and marked on pipe-and: instrument drawings-(P& ids). Examinations and tests are to be performed and evaluated and the results recorded providing a basis for evaluation and comparison with the results of subsequent examinations as required by Code. Visual, surface, and volumetric examinations are defined as specified by Code. Exemptions from examination are to meet 10 CFR 50.55a(b) and Code specifications IWB-1220, IWC-1220, and IWD-1220. Replacements are performed to IWA-7000. 6

i I Examination requirements, methods, acceptante standards, inspection intervals, deferrals, the selection of items toube examined, the number of items to be examined, and the examination-fraction of each weld inspected meet the requirements of Tables IWB-2500-1, IWC-25001, JWD-2500-1, support sample i size, and 10 CFR 50.55a(b)(2). l 2.4.3 1.icense Conditions 3 In a Safety Evaluation Report ' dated November 14,1986,(19) the NRC addressed concerns related to the quality of the. field welding program at Sequoyah. As a result of this review, the NRC required that the Sequoyah plant's first 10-year inservice inspection program be augmented'and accel-erated; i.e., that TVA be required to complete inspections of 100% of ASME Classes 1 and 2 piping and pipe support field welds and major components L (reactor vessel, steam generators, pressurizers,-and reactor coolant pumps) support welds made in the field that are already included-in the first '10-year program in two consecutive refueling outages following.the restart. The licensee) committed to this inspection program in a letter dated-January 30, 1987, W.and has revised the.ISI program accordingly. A relief. request has been submitted from the Code-scheduling requirements-that cannot be met due to the accelerated inspection; plan. This relief request is evaluated-in-Section 3 of this report. 1988(jg) Safety Evaluation Reports ' dated March 31,1988(40) and May 11, the NRC reviewed TVA's program for controlling microbiologically 1 induced corrosion (MIC) in essential raw cooling water piping. The-NRC concluded that a-semiannual weld-by-weld visual inspection program be performed to identify leaking welds along with'a radiographic examination to determine the extent of MIC degradation in the weld volume. subsequent to-l leakage detection. - The' following additional items were also required:- i l (1) at the same time leakage is detected, an assessment ~ should be made i to ensure the acceptance. criterion is not exceeded prior to the i scheduled outage or repair, (2) the frequency of direct visual inspection of weld-specific leaks I should be increased to monthly, and-(3) the leakage from leaking welds should not exceed 0.5 gpm and the i L total leakage from all welds should not exceed 2 gpm.: g-The licensee is performing this inspection under a separate Technical Specification (TI-109) and has not included it in the ISI plan. o 2.5 Conclusions and Recommendations i Based on the evaluation in the preceding sections, it is concluded that the Sequoyah Unit 2 first-interval program meet the requirements of (1) the Code to which it is written, and (2) the NRC regulations (10 CFR 50.55a(b)(2) regarding examination sample and exclusion criteria. Specific requests for relief are addressed in the following section. l 7 L

~ q l 3. EVALUATION OF RELIEF REQUESTS The TVA ISI program for Sequoyah Unit 2,; Procedure SP-114.2, included thirteen (eight Class 3, three Class 2, and-two general) relief requests. The following sections evaluate the thirteen pending relief requests. The material included in the paragraphs titled Code Relief Reouest, Pronosed Alternative ~ Examination, and Licensee's Basis for Reauestina Relief is quoted directly from the relief request except for minor editorial changes such as removing references to figures and tables not' included in this report.- -i dated May 5,1989,]L IF that there are more welds than those specifically.ef RequI Concerning Re listed that may be included under Relief Request ISI-3. Only those welds. specifically listed under Relief Request'151-3 are evaluated in this report. Any additional welds should-be identified at the time the examination is - . determined to be impractical. I e E r ee 5 4 4 t 4 8

3.1. CLASS 1 COMPDNENTS Subsections IWA and 'IWB of the Code govern thE examination of Class II piping and components. Specific requirements are given in Table IWB-25001-l 3.1.1 Reactor Vessel '3 1 1 1 Relief Ranuest 151-5. Reactor Vessel Bottom Hea'd Circumferettigi Weld. Catacorv B A. Item B1.21 Qgig Raouirement All pressure-retaining circumferential and meridional head 3 welds in the reactor vessel head shall be= volumetrically examined in accordance with Figure IWB-2500-3 over the accessible' portion up to 100% of the weld length during the first inspection interval; The bottom head welds may be examined at or.near the end of the ~ interval. j Code Relief Reaues.t Relief is requested from performing a 100% volumetric examination of the lower head dollar weld. Proposed Alternative Examination A remote ultrasonic examination will be conducted from the vessel inside diameter on all accessible areas of the weld. i s Licensee's Basis for Reauestina Relief TVA will employ automated remote inspection devices to examine most of the reactor vessel welds. These examinations will be conducted from the vessel inside diameter. However, the: lower _ head weld on each reactor pressure vessel is part'ially inaccessible for examination from the vessel inside diameter due to instrumentation tubes which penetrate the lower head (weld no. W01-02; see attached i drawings). Portions of the' weld can be examined from one. side (as permitted by T-441.4, Article 4 of Section XI) and will include:100% 2 of the examination volume in accordance with'IWB-3511.1 of Section V. These portions of the weld will be reexamined during the inservice intervals in accordance with the examination Category B-A of Table IWB-2500-1. 9 . 1

4 . j Evaluation The RPV UT. examinations-are' performed from the. vessel inside-surfaces using mechanized positioning equipment. - Parallel and transverse scans of the bottom head cap to-spherical. ring weld are limited due to interference with in-core, instrumentation tubes. - Estimated from.a drawing >rovided in Reference 36, approximately two-thirds to three-fourtis.of the upper side and all of the lower side of the cap-to-spherical ring weld could not be reached.by ultra - l . sonic equipment.during preservice inspections.. The limitations for-the inservice inspection may be reduced due to improvements in j ultrasonic-technology since preservice examination-- Item Bl.21 requires examination-of the accessible length of, circumferential head welds. -Since-in this case the weld:is 100% accessible from the outside surface, that is_ the length of weld thatL 3 the Code requires.. - The licensee performed a manual examination'of 100% of the lower head circumferential weld from the outside surface during preservice examinations.t However, the-licensee has estimated-l that performing this examination for inservice inspection would-require approximately 164 man-hours.and result in at;1 east 16 man-rem of incurred radiation dosage. In Reference-36,'the~1icensee states that his limited inside 4 diameter examination meets the requirements of the Code. However, no estimate 'of the percentage of weld length and volume to be examined; during. inservice inspections has been'provided. Paragraph T-441 of Article 4 should not be construed to mitigate the: requirements of.- Section XI. According to Paragraph T-441 of Article 4, the. scanning. of the examination volume shall-be carried out from both sides of-the-weld on the,same surface wherever feasible. Where configuration or adjacent parts of the component are such that scanning from both" i - sides is not feasible,' this fact shall' be included in the report of the examination. However, 100% of:the.Section XI-required length and volume must still be examined. -Based on ALARA constderations and the fact that the weld'was 100% volumetrically examined during preservice: inspection with no indications found, it would be reasonable to grant relief for this interval, provided examination to the extent practical-is performed - i from the.inside diameter. The licensee should notify the NRC by the end of the ten-year interval clearly what percentage of the weld was inspected. General visual examinations per _IWA-5240 should be' made during each system pressure test for evidence of leakage in the areas of the lower head. Conclusions'and Recommendations Based on the evaluation,,it is. concluded that for the weld l discussed above, the Code' requirements are impractical. It is further concluded that the proposed examinations will provide l E 10 i ..e- .w ,__,c.-p ,y-e -. ins 9 v. 9, y ,q_, yeev-

+ + - t 3 necessary assurance of structural reliability durin's this interval, i Therefore, relief lis recommended as requested, provided the weld is volumetrically examined to; the extent practicalf from the inside. surface and the NRC-is notified of the percentage examined.= -i i References References 26, 27, 28, 29, 30, 31, 32, and 36. 3 P t ? i ( l l l F i-E i I i 11 -.... i

3.3.1.2 Relief Reauest 151-10.~ Reactor Vessel F1anae-to-Unner Shel1~ Weld. { Cateoorv B.A. Item Bl.30 . E. Code Reauirement 5 Essentially 100% of the length of shell-to-flange welds shall:- be volumetrically examined in accordance with Figur.e IWB-2500-4 .during the:first inspection interval.. Examinations shall be. l scheduled according to Table IWB-2412-1. i Code Relief Raouest . Relief-is requested to defer volumetric" examination o'f the shell-to-flange weld to near the end'of the inspection interval rather than the required examination frequency as _given. in Table IWB-2412-1. i Pronosed Alternative Examination A remote ultrasonic examination of the: subject weld wills be l t conducted from the vessel.inside diameter near~the-end:of the r l inspection interval. I Licensee's Basis-for Reauestina Relief-The reactor vessel flange-to-upper shell-weld is located behind i the core barrel making it inaccessible until the core barrel is-removed. l l The distance from the-vessel' flange-to-upper shell. weld,- l 41.9 in., is a significant distance from the' flange face. The usage-i factor at this distance is very low-(0.00662), whichtindicates.the-i weld loading is similar to the loadings of the vessel.circumferential L shell welds rather than a flange weld. Deferral'of the~ flange weld, as allowed for shell welds, is therefore appropriate. p Evaluation Vendor-supplied drawings indicate that the core support barrel is supported by the core support ledge. The vessel-to-. flange-weld is located 29.32 inches below the core support ledge. Therefore, to gain access to the weld for inspection,.the core barrel must be removed; an operation normally performed only once per 10-year interval. 12 1

.o -i 1 Sequoyah's' construction permit was11ssued May 27, 1970, thus design for. inservice ~ inspection access is not required by 10 CFR - 50.55a(g). Further, it is agreed that consideration of-the upper a shell-to-flange weld as a shell-weld is appropriata and would not. result in' a significant decrease in plant safety.- Conclusions and Recommendations b Based on the evaluation, it 'is: concludedfthat' for the weld discussed:above,= adherence to the Code requirements is impractical. It is further concluded that the proposed examinations will provide-necessary assurance of structural reliability during this interval. Therefore, relief.is recommended as requested, provided the licensee performs the-Code-required examination when:the-core barrel is removed near the end of the interval.. References References 26, 27, 28, 29,:30, 31, 32,~and'35. n 3.1.2 Pressurizer No relief requests. ] ~! j I l i l 13 L

?! i 3.1.3? SteamGenerators_and'HeatExchangers;-- 3.1.3.1, Relief Recuest ISI-6. Steam Generator Nozzle inside? Radius Section. Cateoory'B-D. Item B3.140 Code Reauirement i Volumetric examination is required of all-primary steam i generator nozzle inside radius sections' covering the volune: described in Figure IWB-2500-7_ during each inspection interval, a

~

The nozzle-to-vessel' weld and adjacent areas of-the nozzle and. _.. vessel-are included. At least 25% but not more than 50%'(credited): =; of the nozzles shall be examined by.the end of the first inspection _ period and the remainder by the end_ of the inspection interval.. 7 i 3 ) Code Relief Reauest i Relief is requested from' volumetric examination of the; steam s generator-integrally cast nozzle inside radius _ sectionLfor this i L

interval, 4

Procosed Alternative Examination TVA will initiate examination of the steam generator nozzle; inner radius sections in accordance with the applicable Section XI Code during the second ten-year-inspection interval. This will-provide for examination of all the' steam generator' nozzle inner i l radii by the period ending at half the plant life. Licensee's Basis for Reauestina Relief Each steam generator. consists of two integrally cast nozzles and two integrally cast manways. Relief =from the inspection requirement above is based on EPRI Report NP-4242, "Long-Term Inspection. Require-ments for Nuclear Power Plants," dated. March 1986; : This = report presents a linear fracture mechanics analysfs which-predicts that~ cracks the size of 0.25 inch. reference flaw size) will prop (which is greater than the allowable-agate to only slightly greater than one-sixteenth of the nozzle wall thickness during the entire life:of the plant. 'The report proposes that the nozzle _ inner radius tut examined.at least at half the plant life, and subsequently, at' the regular Code inspection intervals, j i l 14 ,m

q The primary chamber radiation exposure dose rate is generally _ i on the ' order of 30 renVhr. As. a result,- individual " stay-time" in. the' chamber would be limited to a degree where meaningful results ~ from alternative surface and visual examinations could not be ( achieved. In addition, the presence of the interior surface: ~; austenitic stainless steel cladding, which has a-higher ductility' than the base ASME-SA-216 grade WCC casting material, raises the possibility of under cladding cracking which would not be visible with'a surface examination. Evaluation I l The licensee is basing relief request ISI-6 on EPRI ReportL l NP-4242,(41) "Long-Term Inspection Requirements for Nuclear Power Plants."

This report provides an; assessment of the failure modes -

j -of steam generator primary side nozzles. The report concludes that "... the primary side nozzle of the steam-generator is in-4 no danger of leaks during,the entire life of the plant ifl the nozzle is integrally cast with the chamber... The con-servative stress' distribution across.the nozzle' thickness-predicts that cracks the size of 0.25 in. (roughly four times the allowable reference flaw size) will propagate: i to only slightly_ greater than one-sixteenth of the nozzle-thickness during the entire life of the-plant.

However, the effect of low amplitude high-cycle variations of the stresses are neglected in'these calculations.

To be more conservative, these nozzles should be~ inspected at least at half the plant life, and subsequently, the Code interyd s may be followed." Relief is warranted for this interval, based on the fact that: current ultrasonic examination methods generally arovide=unsatis-l factory results on integrally cast nozzles,; and t1at performance i i of a visual or surface: examination would result in high radiation-dosages with no assurance of a meaningful' examination; -If it is 4 necessary to enter the steam generator _ inlet and outlet plenums. for maintenance or other reasons, a' visual. inspection _ of theLnozzle inner radii should be performed; Visual examination during the required system pressure;and hydrostatic tests will provide initial evidence-of through-wall-i leakage for this interval, i L Conclusions and Recommendations 1 Based on the above evaluation, it is concluded that;for the nozzle examination discussed above, the Code requirements are impractical. It is further concluded that the alternative exami-f nation discussed above will provide-adequateLassurance of structural reliability. Therefore, relief is recommended as requested, provided: j 15 .I

u =' e i (a) the licensee should visually inspect the. nozzle inner radii'-if~ it..is necessary to enter the steam generator inlet and outlet - l plenums, for maintenance or..other inspection activities. l l' 1 References ~ References 26, 27, 28, 29. 30, 31, 32, 35,'and 41. q -t '. { ? + i f ll-i ~ I 4 4 y l. 16 f .4.,-->

3.1.4 Piping Pressure Boundary 3.2.4.1 Relief Raouest ^151-3. P*eiture-MAinino Dissimilar Metal Va16s in pinina. Catanerv B F. Item B5.5Q Code Reauirement All dissimilar metal safe end welds in piping shall be surface and volumetrically examined in accordance with Figure IWB 2500 8 during the first inspection interval. The examinations may be performed coincident with the vessel nozzle emainktions required by Examination Category B-D. Dissimilar metal welas between combinations of (a)lloy steel to high nickel alloys,gh alley steel, carbon or low alloy steels to< hi (b) carbon or low a and (c) high alloy steel to high nickel alloys are included. Code Relief Reauest Relief is requested from performing a 100% volumetric examina', ion of dissimilar metal piping welds with interferences. g Prooosed Alternative Examination A "best effort" ultrasonic examination and a surface exami. nation will be performed on accessible areas of the welds, Licensee's Basis for Recuestino Relief Some piping welds will be impractical to examine from both sides due to non removable hanger interferences or valve and pump casings joining the welds. Evaluation The two Category B-F welds for which relief is requested are listed in a table attached to relief request 151-3, along with details of the scan limitations encountered for each weld. Weld RCW-24 SE is limited to 40% of the Code-required scan due to pressurizer safe end geometry and nozzle curvature. Weld RCW 29 is limited to 90% of the Code-required scan due to physical obstructions and pressurizer nozzle geometry. The proposed alternative examination of an ultrasonic examination to the 17 A l llllE ll l

E neximum extent practical and a surface examination, along with the Code required system pressure and hydrostatic tests, will provide adequate assurance of structural reliability. 4 Conclusions and Recommendations Based on.the above evaluation, it is concluded that for the piping dissimilar metal welds discussed above. the Code requirements are impractical. It is further concluded that the proposed altern-i native examination will provide adequate assurance of structural reliability. Therefore, relief is recommended as requested. i i References References 26, 27, 28, 29, 30, 31, and 32. t t g D 1 1 4 l l I 18-I l l-

3.1.4.2 Relief Renuest 151-7. Reactor Coolant - Loco Pioino Walds. s Catenorv B-J. Item B9.10 Code Raouirement The licensee has optionally (as allowed by 10 CFR 50.55a(b) (2)(11)) chosen to determine the extent of-examination of Clas;-1 piping welds using the IC74 Edition, Summer 1975 Addenda. This Code requires that examinations be performed on all the area of 25% of the circumferential joints (including the adjoining 1 ft sections of longitudinal welds) each interval. - A different 25% sample is required in successive intervals. The 1977 Summer 1978 Code requires-that for circumferential and longitudinal welds in pipe of. nominal size 4 in, and greater, surface plus volumetric examinations be performed in accordance with Figure IWB 2500 8. Code Relief Reauest Relief is requested from performing a volumetric examination.of' 1 two reactor coolant loop piping circumferential welds'(RC-2351 and: RC-3151). Proposed Alternative Examination i None. Licensee's Basis for Reauestino Relief The two subject welds are located inside the reactor vessel shield wall and are thus inaccessible. l l Evaluation Sequoyah's inservice inspection plan has opted, as: allowed by 10 CFR 50.55a(b)(ii), to select Class 1 piping welds for examina-tion according to the rules of the 1974 Edition with addenda through Summer 1975. This edition of the Code allows a 25% sample of piping welds to be, examined during the first inspection interval, with a different 25% sample for each succeeding inspection interval. Therefore, the two inaccessible welds would have to be included in the examination sample sometime over the 40 year life of the plant. However, relief from Code requirements can be delayed 19 .c

until the fourth interval. For the first inspection interval, welds other than the two inaccessible welds should be picked for examination. l Conclusions and Recommendations .j Based on the above evaluation,~ it is concluded that relief is not necessary at this time and should not be granted. Relief should not be requested until the ISI plan for the fourth inspection interval is made. ] References References 26, 27, 28, 29, 30, 31, and 32. i i I l ) l ) 1 20 m

3.3.4.3 Relief Reauest ISI-3. Pressure-Retainina Welds in Pinina. Cateoorv B-J. Items 89.10. 89.20. and 89.30 (Items 4.5. 4.6. and 4.7 in 1974 Summer 1975 Code) i Code Reauirement As permitted by 10 CFR 50.55a(b)(2)(ii), the. licensee has elected to determine the extent of examination for Class 1 piping by the requirements of Tables IWB-2500 and IWB 2600, Category B-J of Section XI of the ASME Code 1974 Edition with Addenda through Summer. 1975. This Code version requires examination of longitudinal and' circumferential welds and base metal for one wall thickness beyond the edge of the weld. Longitudinal welds.shall be examined for at least 1 ft from the intersection with the edge of the. circumferential weld selected for examination. For connections, the areas shall include the weld metal, pipe branch the base metal for one pipe wall thickness beyond the edge of the weld on the main pipe run, and at least 2 in of the base metal along the branch run, i t The examinations performed during each inspection interval shall cover all of the area of 25% of the circumferential joints including the adjoining 1-ft sections of longitudinal joints and 25% of the pipe branch connection joints. A different 25% is to be examined each interval. l The 1977 Summer 1978 Code requires a volumetric and surface examination in accordance with Figure-IWB 2500 8 of longitudinal and circumferential-welds in pipes 4 in, or larger. For pipes smaller i than 4 in., only a surface examination is: required. For branch connection welds in pipe 4 in, or larger, a surface and volumetric examination is required in accordance with Figures IWB 2500 9, -10, i and 11. For pipes smaller than 4 in., only a surface examination is ~ required. Code Relief Raouest Relief is requested from performing a 100% volumetric 1 examination of piping welds with~ interferences. fronosed Alternative Examination i A "best-effort" ultrasonic examination and a surface examination will be herformed on accessible areas of the welds, f s D 21

N f Licensee's Basis for Reauestina Relief \\ Some piping welds will be impractical to' examine from both sides ( due to non-removable hanger interference or valve and pump casings joining the welds 9 Evaluation The five Category B-J welds for which relief is requested are listed in a table attached to relief request 151-3, along-with details of the scan limitations encountered for each. weld. Of these welds, three are in the Reactor Coolant System (Welds RC 3, RC-152, and RCF-23), one is in the Safety Injection System (Weld SIF-128), and one is in the Residual Heat Removal System-(Weld RHRH-125). The major reason for requesting relief is geometric configuration of the weld. Four of the five welds for which relief is requested will' i receive between 50% and 98% of the Co6-required scan. For these welds, the proposed alternative of a surface examination and the Code-required system pressure and hydrostatic. tests will provide adequate. assurance of structural reliability, t For Weld SIF-128, geometric configuration obstructs 100% of ] Code required ultrasonic scans. - The licensee-has opted per 10 CFR - 50.55a(b)(2)(ii) to use the 1974 Edition, Summer 1975 Addenda for-selection of Class 1 piping welds for examination. ' Under this system, a different 25% sample of welds it selected each 10-year interval, so that at the.end of the 40 year life of the plant, each Category B J piping weld will have been inspected once. For Weld i SIF-128, it is recommended that ISI examinations be put off until + the third or fourth 10-year interval to allow developing ultrasonic examination technology to be utilized in the inspection. For this interval, a different weld should be chosen that will allow a greater percentage of the weld to be examined using currently available ultrasonic equipment. The NRC staff should be notified which weld. has been substituted and the percentage of the Code-required examinations that is practical on the weld. l l l Conclusions and Recommendations I Based on the above evaluation, it is concluded that for the e piping welds discussed above, the Code requirements are impractical, i It is further concluded that for four of the five welds discussed-above (Welds RC-3, RC-1 52, RCF-23, and RHRH 125), the proposed r alternative. examination will provide adequate assurance of structural' reliability. Therefore, relief is recommended as requested for these ~ four welds. l I l t L 22 l

For Weld 51F-128, it is recommended that examination be. delayed to allow developing ultrasonic technology to be utilized in the examination. A different weld should be chosen for inspection this i interval that allows a good ultrasonic scan using currently available equipment. The NRC staff should be notified concerning the i substituted weld. References i References 26, 27, 28, 29, 30, 31, and 32. i b 9 k i i b i s k i 4 h s 23 d

1 n 3.1.5 Pump Pressure Boundary-1 3.1.5.1 Relief Raouest ISI-B. Pressure-Retainino Welds on Pumo'tasinos. Cateoory B-L-1. Item B12.10 Code Reauirement Essentially 100% of the pressure retaining welds in at least one pump in each group of pumps performing similar functions.in the system (e.g., recirculating coolant pumps) shall be surface and volumetrically examined in accordance with Figure IWB 250016 during q each inspection interval. The examinations may be performed at or n6ar the end of the inspection interval. i Code Relief Raouest Relief is requested from performing a volumetric examination of. 3 l the reactor coolant pump casing weld. h prooosed Alternative Examination ~ One reactor coolant pump casing weld will be surface examined during each inspection interval. Licensee's Basis for Reauestino Relief Each reactor coolant pump casing consists of a-two piece welded Type 304 SST casting. The present capability of ultrasonic testing is not sufficient to examine cast material of this thickness and i achieve meaningful results. [ Evaluation The present NDE volumetric techniques available do' not provide meaningful results for welds in cast pump casings. Surface exami-nations in conjunction with hydrostatic testing and pump performance data, as required by Subsection IWP of Section XI, should provide adequate assurance of the pump case weld integrity. -. Examination technology is currently being actively developed. The licensee should cMnmit to evaluating new NDE technigs'es as they become available, and consider their application to the volumetric examination of the pump casing welds. l r t p 1 24

d Conclusions and kacommendations Based on the evaluation, it is concluded that for the. weld discussed above, adherence to the Code requirement is impractical. It is further concluded that the proposed examinations will provide necessary assurance of structural reliability during this interval. j Therefore, relief from volumetric examination is recommended as requested, provided 1 i (a) one reactor coolant pump casing weld is surface examined 4 during each inspection. interval-2 (b) visual examination of the pump casing for leakage is conducted in conjunction with system leakage and j hydrostatic tests under Oategory B P l (c) periodic inservice testing of the pumps.is conducted in accordance with IWP i i \\ It is further recommended that the licensee keep abreast of improvements in state-of the-art NDE techniques. If a means becomes available for examining the Code-required pump casing welds, these welds should be examined in lieu of the examinations recommended above. References l References 26, 27, 28, 29, 30, 31, and 32. l l l l l 4 L l i i 25

3.1.5.2 Relief Recuest 151-1. Pumn Internal Pressure Boundary Surface. Cateacry B-L-2. Item B12.20 j Code Renuirement The internal surfaces of at least one pump in each group of' l pumps performing similar functions in the system (e.g., recircu-lating coolant pugs) shall be visually examined (VT-1) during each l inspection interval. The examination may be performed on the same pump selected for volumetric examination of welds. The examinations may be performed at or near the and of the inspection interval. Code Relief Reauest i Relief is requested from the requirement of performing a visual examination of reactor coolant pump internal surfaces unless a pump is disassembled for maintenance. Further, if one pump from either unit is disassembled and examined, it is considered that this is representative of all reactor coolant pumps and no further examina-i tions will be performed to satisfy Examination Category B L-2. Proposed Alternative Examination The internal surface of the reactor' coolant pump casing will be visually examined whenever the surfaces are made accessible when a pump is disassembled for maintenance purposes. If during the i 10 year interval, a pump from either unit is not disassembled for maintenance, a pump from one unit shall be examined from the exterior. This shall be accomplished by ultrasonic thickness [ measurements of the pump casing. Liceasee's_ Basis for Reauestino Relief In the absence of required maintenance, disassembly of a reactor coolant pump solely to perform a visual examination of internal sur-faces is impractical. This would represent an. unnecessary employee exposure to high radiation and contaminatior areas and an excessive expense to TVA. In addition, the two units at Sequoyah Nuclear Plant will-operate unde'r similar conditions. Therefore, if a pump from one-of the units is disassembled for maintenance during a 10 year e 26 i ~

interval, the visual examination petformed will be representative of the pump condition for each unit. This would avoid unnecessary employee exposure to high radiation dose rates. Evaluation Disassembly of the reactor coolant pumps to the extent required for examination under Category B-L 2 is a major maintenance effort. Inspection of a pump on a similar plant required nearly 6000 manhours and resulted in nearly 50 manrem exposure. In addition, the pump inspection activity generated a large volume of radioactive waste. The visual inspection of the pump internal surfaces is intended to detect erosion, corrosion,'and surface cracking. The structural reliability of reactor coolant pump casings in service has been good, and no significant degradation har been detected. Accordingly, the-large expenditure of effort to disassemble and inspect edch pump is not practical in view of the reliability of primary coolant pump casings. Visual examination of each pump internal surface would not produce an increase in plant safety commensurate with the cost of conducting the examinations. Considering the examination of any one reactor coolant pump from either unit as representative of all pumps for each unit is reasonable considering their similar operating conditions. The licensee's commitment to perform ultrasonic measurements of the thickness of the pump casing, if a visual examination is not performed, coupled with visual examination for leakage during system pressure tests under Category B-P and IWP test data, should provide adequate assurance of pump integrity. Conclusions and Recommendations Based on the above evaluation, it is concluded that visual examination of pump casing internals is impractical. It is further concluded that the proposed examination will provide necessary assurance of pump reliability. Therefore, relief is recommended as requested provided (a) the required visual examinations are conducted under Category B-t-2 if a reactor coolant pump from one unit is disassembled for maintenance (b) if,' during the ten-year interval, a pump from either unit-is not disassembled for maintenance, a pump from one unit 'shall be examined from the exterior by ultrasonic thickness measurements 27 (

j ,(c) visual examination of the pump casing for leakage is conducted in conjunction with system leakage and hydro-i static tests under Category B P (d) periodic inservice testing of the pumps is conducted in f accordance with IWP. I References References. 26, 27, 28, 29, 30, 31, and 32. 7 h 5 R R 4 r 28

1 i 3.1.6 Valve Pressure Boundary ) 3.1.6.1 Relief Raouest 151-2. Valve Internal Pressure Roundary Surf ace. Cateaorv B M-2. Item B12.40 1 l Code Reouirement Visual examination (VT-1) of valve internal surfaces of at. least one valve within each group of valves that are of the same. l constructional design, manufacturing' method, and that are performing i similar functions in the system shall be performed during the first j inspection interval. The examination may be performed on the same t valve selected for volumetric examination of welds.- The examinations may ba performed at or near the end of the inspection interval.- Code Relief Recuest Examination of one valve in each group of valves of the same constructional design and manufacturer performing similar functions, from either unit during the 10-year interval will satisfy the examination requirements of both units. Proocied Alternative Examination Examination of valves will be performed as they are dis-i assembled. For valves that remain to be examined as the end of the inspection interval nears, a case-by-case study,will be made to determine the practicality of disassembling a valve from one of the units solely for visual examination. If necessary, a request for relief will be made at the time it is determined that disassembly of a particular valve is impractical. 1 Licensee's Basis for Reauestina-Relief l 1 During routine maintenance, visual examinations of valve body I internal pressure boundary surfaces are performed and documented-under existing plant administrative procedures. Most Class A valves, f I particularly containment isolation valves, are disassembled fre-quently for maintenance. In addition, the two units ~at Sequoyah' Nuclear Plant will operate under similar conditions. If a valve-from one of the units is disassembled for maintenance within a. l 10 year i'nterval due to the similar operating characteristics of both units, we feel that the visual examination performed would be 29 1 2

v representative of both units and would be sufficient to satisfy the examination requirements for both units for that particular valve classification as defined in Examination Category B M-2. ,l l - l! 1 Evaluation Disassembly of large valves to the degree necessary to examine l the internal pressure-retaining surfaces is a major effort, which may involve large personnel exposures. To do this disassembly solely to j perform a visual examination of the internal body is impractical. The licensee has committed to the concept of visual examination if a similar valve from either unit is disassembled for maintenance. The visual examination specified is to determine'whether anticipated r severe degradation of the body is occurring due to phenomena such as erosion or corrosion, i The request to treat the examination of a particular~ valve group I from.one unit as satisfying examination requirements for both units is reasonable, considering the fact of similar operating conditions for both units and the undue hardship created by requiring dis-The Code requires that only one valve assembly of a working valve. in a group of valves of the same construction, performing similar functions, need be examined. It is evident, from the fact that the t examination requirement is for only one valve and not a sample percentage of valves, that the intent of the Code is. essentially met I by examination of-one of similar type and manufacturer between each If, however, a certain type of valve is not examined at either unit. t unit and the examination is deemed -impractical, the licensee should request relief on a case-by-case basis near the end of the_ interval. These examinations, coupled with the valve testing requirements of IWV and hydrostatic inservice inspection requirements, should provide adequate assurance of valve integrity. - P Conclusions and_J g g ndations-Based on the above evaluation, it is concluded that visual examination of valve casing internals is impractical. It is further concluded that relief should be granted provided: (a) one comparable valve in either unit is examined. If a certain type of valve is not examined at either unit, reli f should be requested on a case by-case basis near e j the end of the interval, 4 i 30

l l (b) periodic inservice testing of the valves is conducted in t accordance with IW, and i I (c)- visual examination of the valves for leakage is conducted in conjunction with system leakage and hydrostatic tests under Category B P. p I I l \\ l References l References 26, 27, 28, 29, 30, 31, and 32. .I l l I ) i l e \\ t l I l p B I s d 4 e 31 I

) 3.2 CLASS 2 CDMPDNENTS-4 j. 4 Subsections IWA and IWC.of the Code govern the examination of ) I i Class 2 piping and components. Specific requirements are given in Table i ~ JWC-2500-1. I t fj 3.2.1 Pressure Vessels and Heat Exchangers i t 3.2.1.2 Relief Reauest 151-4. Steam Generator Class 2 Circumferential 4 i 1 Shell Welds. Cateaory C.A. Item C1.10 + 3 i i Code Reauirement Essentially 100% of the shell circumferential welds at gross i [ structural discontinuities shall be volumetrically examined in accordance with Figure IWC 2520-1 during each inspection interval, g A gross structural discontinuity is defined in NB 3213.2. For i multiple vessels with similar design.-size, and service (such.as ) steam generators and heat exchangers), the required examinations ? j may be limited to one vessel or distributed among the vessels. 1 Code Relief Reauest Relief is requested from volumetrically examining one cirs cumferential shell weld on each generator (Welds SGW DI, -D2, 03, and -D4). Procosed Alternative Examination Ultrasonic examination to the extent practical. Licensee's Basis for Recuestina Relief i Steam generator circumferential shell weld numbers SGW-DI, -02, D3, and D4 are inaccessible due to the location of the upper steam l. generator support brackets. e 32

4 O Evaluation Steam generator welds SGW DI. -D2, -D3, and D4 are partially inaccessible for ultrasonic examination because of a permanent support ring located at the edge of the weld. The steam generator weld allows access for ultrasonic examination from the conc side only. The support ring is held in place by wedging (1/2"x6"x12") steel pads between the support ring and the steam generator. These pads are spaced at approximately 13 inch intervals 360 degrees around the. support ring. Approximately 42% of the weld circumference is j obstructed by the pads and the remaining 58% of the weld is, obstructed by the support ring itself. Even where the steel pads. rest against the weld, part of the weld and heat-affected zone can be examined using 45-and 60 degree scan angles from one side. A greater percentage, though still not 100% of the required volume, is accessible where only the support ring provides interference. The licensee has estimated that for the 1 entire weld at least 68% of the Code required volume can be examined. The licensee performed this "best effort" examination on SGW-01 during the Unit 1, Cycle 3 outage. Sequoyah's construction permit was issued May 27, 1970. Thus. 4 design for inservice-inspection access is not required by 10 CFR 50.55a(g). A best effort examination to the extent practical within the limitations of design,~ geometry, and materials of construction will meet the intent of 10 CFR 50.55a(g)(4). Conclusions and Recommendations Based on the above evaluation,'it is concluded that relief 1 from the Code requirements of a volumetric examination of the above mentioned steam generator circumferential shell welds should be granted, provided that the welds are ultrasonically examined to the extent practical. References References 26, 27, 28, 29, 5, 31, 32, and 36. I i 33 l ll

c l ~ l [ 3.2.1.2 Relief Renuest 151-12. Excess Letdown Heat Exchancer and Reaenerative Heat Exchancer Shell Welds and Intearally Welded 1 Attachments.Cateaories C-A. C-C. and C.E. Items C1.10. C1.20. C1.30. and C3.10 3 Code Reauirements Essentially 100% of the shell circumferential welds at ross structuru discontinuities (Item Cl.10) shall be volumetrica ly d examia.ed in accordance with Figure IWC 2520-1 during each inspection is.ierval. A gross structural discontinuity is defined in NB 3213.2. I Examples are junctions between shells of different thicknesses, cylindrical shell to conical shell junctions, and shell- (or head) to flange welds, and head to-shell welds. Essentially 100% of the circumferential head to shell weld (Item C1.20) and the j tubesheet-to.shell weld examinedinaccordancew(thFiguresIWC25201andIWC25202, Item C1.30) shall a i respectively. For multiple vessels'with similar design, size, and 4 i service (such as steam generators and heat exchangers), the required examinations may be limited to ene vessel or distributed among the- } vessels. In accordance with Item C3.10, the surfaces of_100% of each integrally welded support attachment in pressure vessels shall be surface examined in accordance with Figure IWC 2520-5 during each inspection interval. Examination is limited to plate-or shell type and linear-type supports whose base material design thickness exceeds 1/2 in. For multiple vessels of similar design and service, the required examinations may be conducted on only one vessel. Where multiple vessels are provided with a number of similar supporting elements, the examination of the support elements may be distributed among the vessels, 9 J [pde Relief Reauest Relief is requested from performing 100% volumetric examination f of shell welds and 100% surface examination of integrally welded i attachments in the excess letdown heat exchanger and regenerative heat exchanger. i Prooosed Alternative Examination h The ELHX w' eld was examined ultrasonically from one side for almost the entire length, except where the head's flanged nozzle-inlet and outlet welds interfered with the examination. This allowed for a UT examination of 90% of the weld from the one side. From the j i 34 I

l opposite side, the head taper and head closure studs and nuts severely restricted access to the weld area and as a result, no meaningful results could be obtained. The UT was performed on a best-effort basis and supplemented with a liquid-penetrant test (PT) { over 100% of the weld area. In the case of the RHX when examinations were performed, the g state-of-the-art UT examination capabilities did not allow the i achievement of meaningful results because of the component fabrication processes involved. The RHX is a centrifugally cast stainless steel, SA-351, CF8

vessel, in addition, the examinations were hindered by supports which were essentially nonremovable from the area of two of the four j

welds because of the location in the heat exchanger room and their position on the vessel. As a result, a PT was performed in place of the required UT. On the two restricted welds examined, only 75% of the weld lengths were surface examined because of supports covering part of the weld areas. No alternative is proposed for surface examination of integrally g welded attachments of the subject heat exchangers. I Licensee's Basis for Reauestina Relief 5 In accordance with 10 CFR 50.2(v) of the Code of Federal l Regulations, the RHX and ELHX are defined to be within the reactor coolant pressure boundary (RCPB) (ASME Class 1). Secondly, 10 CFR 50.55.2(c) further states: l "(2) Components which are connected to the reactcr coolant system and are part of the reactor coolant pressure boundary l defined in 50.2(v) need not meet these requirements (Class I requirements),provided: (ii) The component is or can be isolated from the reactor coolant system by two valves (both closed, both open, or one closed and the other open). Each open valve must be capable of automatic actuation and, assuming the other valvo is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by-the reactor coolant makeup system only " As a result of this paragraph, the RHX, ELHX, and the connecting piping to the Reactor Coolant System (RCS) and piping out to the containment isolation valves would be exempt from ASME Class 1 examination requirements. 35

Sequoyah Nuclear Plant Units 1 and 2151 programs, 31-114.1 and $1-114.2, classify components for examination in accordance with Regulatory Guide 1,26 Revision 3. The guidelines of Regulatory. Guide 1.26, paragraph C.), classify components which are exempted by 10 CFR 50.55a(c)(2)i1 from ASME Class 1 examination requirements (such as the RHX, ELHX, and associated piping) as ASME Class 2 equivalent. The Sequoyah finti Safety Analysis Report (FSAR), Section 9.3.4.1.7, further describes the RHX, ELHX, and associated piping as *not required to function during a loss-of coolant accident." Additional system design considerations are the following: -(1) The ELHX is not used during normal plant operations and is isolated from the RCS by three valves, two normally closed gate valves, and a rhock valve. (2) The AHX is isolated frem the RCS by two check valves. leakage from the RCS is prevented by the two check valves in series, which were leak-tested during the preoperational test program. Leakage also can be detected by monitoring for signs of system incomin leakage and by monitoring the RCS for signs of outgoing 1 akage (FSAR 67.3-26). (3) There are no leakage problems into the CVCS from the RCS because of the higher pressure at which the CVCS is generally maintained (FSAR 6.3-27). In addition, the inlet and outlet piping (one inlet and one outlet) associated with the ELHX is 1-in, nominal pipe size (NPS)'and 3-in. NPS for the RHX. Both of these are less than the 4-in. NPS which, in cccordance with IWC 1220(c), is exempt from the examination requirements in Table IWC-2500-1 (Class 2) of ASME Section XI. Also note that the Winter 1980 Addenda, paragraph IWB-1220(b)(2) exempts fram examination components and their connections in piping of 1-in. NPS and smaller. (Having one inlet and one nutlet @ipe, each 1-NPS or 31saller.) TWA also feels shst the.ALARA tensiderations t -,,e the liquid penetrant (lPT). examinations on the ELHX.and RHK are excessive for the benefitscrecetwed from.the.. inspections.. The actual radiation exposums to perforn the first inspection period examinations alune asem 2,200 ftREM and 5,4001etal, respectieely. Based on the component classifications as defined in 10 CFR 50.55, the nonsafety related design of the heat exchangers and their supports preventing 190% examination coverage, and the ALARA con-siderations in view of the.small benefit realized, TVA requests relief from the ASME Section XI examination requirements as shown in Table JWC 2500-3. Examination Categorias C-A, C-I and C-E. J

- sqr 4 i Evaluat ion 1 The ELHX and RHX are properly classified as ASME Codo Class 2 in accordance with Regulatory Guide 1.26, Section C.). The licensee is correct in asserting that the ELHX and RHX are not required to meet Class 2 requirements in accordance with lo CFR 50.55, Footnote 2. i The licensee has stated that the inlet and outlet piping i associated with the ELHX is 1 in, nominal pipe size (NPS) and l 3 in, NPS for the RHX. IWC 1220(c) allows exemption for component 4 connections (including nozzles in vessels and pumps), piping And 1 associated valves and vessels (and their suonorts) that are 4 in. NPS and smaller. Therefore, the LHX and RHX may be exempted from the examination requirements of IWC, and relief is not required. -l l Conclusions and Recommendations Based on the above evaluation, it is concluded that the heat i exchangers discussed above are exempt from the examination require-ments of IWC-2500, and therefore relief'is not required. References Reference 32. 5 i 1 F i l 4 a l l l L 37 1

c 3.2.1.3 Eglief Raouest 151 13. Residual Heat Removal Heat Exchanner Nozzle to Vessel Welds. ' Cateaory C B: Item C2.20 i i Code Reauieement All nozzles in vessels over 1/2 in. in nominal' thickness at terminal ends of. piping runs shall. be surface and volumetrically examined in accordance with Figure IWC-2520 4 during each inspection j interval. Terminal ends are the extremities of piping runs that connect to vessels. Only those piping runs selected for examination I under Examination Category C F are included. l Code Relief Renuest 1 Relief is requested from volumetric examination of nozzle-to-J vessel welds in the Residual Heat Removal Heat Exchanger. 1 i Pronosed Alternative Examination I The nozzle to vessel welds will be examined utilizing a surface only examination. I Licensee's Basis for Recuestino Relief i Each heat exchanger consists of an inlet-outlet head chamber with one inlet and one outlet nozzle and two integrally attached support brackets. Volumetric examination of the nozzle to vessel l welds and the nozzle inner radii is inhibited by the shell-to-J tubesheet weld, shell to head weld, and the integral supports. I Proximity of the welds and inner radius sections do not allow for o the achievement of meaningful results from a volumetric' examination. Evaluation [ Sketches have been provided with Relief Request 151-13 showing the limitations to ultrasonic examination. It is agreed that these l l limitations could restrict the performance of the required examina-i tions. Integrally welded attachments are within 11/8 in on two sides of the weld, and shell welds are within 1/2 in. on.the other i two sides, rendering the weld essentially inaccessible to any meaningful ultrasonic examination. The proposed surface examination, along with the Code-required system pressure and hydrostatic tests.- 1 will provide adequate assurance of structural reliability. J 38 l 1

t Conclusions and Recommendations j Based on the above evaluation, it is concluded that for the nozzle to-vessel welds discussed above, the Code requirements are impractical. It is further concluded that the proposed alternative examination will provide adequate assurance of structural reli-ability. Therefo m, relief is recommended as requested.- a 3 i References Reference 32. 1 r ? a i b f I l 4 39 a

I 3.2.2 Piping. Pressure Boundary i No relief requests. i 3.2.3 Pump Pressure Boundary No relief requests, j i 3.2.4 Valve Pressure Boundary No relief requests. 3.3 CLASS 3 COMPONENTS No relief requests. 3.4 PRESSURE TESTS No relief requests. 2 t L F 0 4 -r 40 t 4 =

1 H 3.5 GENERAL i 3.5.1 Examination Method 3.5.1.1 Relief Renuest 151-9. Ultrasonic Calibration Block for Ferritic-Material tess than Two Inches Thick 1 i I Code Recuirement Paragraph IWA-2232(c) requires ultrasonic examination of welds in vessels less than 2 in. in thickness be in accordance with Article 5 of Section V. . Paragraph T-533.2(a). Article 5 of Section V requires that the basic calibration block include a basic calibration hole drilled parallel to the contact surface. Paragraph T-533.2(b) j permits the use of other calibration reflectors provided equivalent responses to that from the basic calibration hole are demonstrated. Code Relief Recuest i Relief is requested from the requirement that the basic calibration block include a calibration hole drilled parallel to the contact surface. I Erecosed Alternative Examination The basic calibration block will include a five percent notch located on the I.D. and 0.0. surfaces (nominal depth 5% of thickness). Licensee's Basis for Raouestina Relief The licensee currently uses five-percent notches.in lieu of side-drilled holes. Although the use of the five-percent notch 1 cannot be shown to be equivalent in all cases to the applicable side drilled holes, the licensee considers that examinations are technically acceptable based on the calibration requirements of Paragraph 111-3430 of Appendix !!! to the 1977 Edition, Summer 1978 Addenda of ASME Section-XI. The calibration notches for ferritic material are 10%t when t is less than 0.312 in, and 0.104t -0.009t2 q for material 0.312 in, to 6 in, thick. l l 41 r

4 Evaluation The ~1977 Edition, ' Summer 1978 Addenda allows ten-percent notcheg for material less than 0.312 in. thick and notches of 0.104t -0.009t - for 1 material 0.312 in. to 6 in, thick. - The licensee's commitment' to use five-percent notches is considered equivalent to or exceeding the 1977 Edition, Summer 1978 Code examination techniques. An important feature of the overall ISI program is that past - inspections serve as ll baseline by which inservice examination - results.are evaluated. Accordingly, it is appropriate:to use methods-during inservice inspection which are consistent with those used j previously provided the previous examination methods were technically acceptable. Conclusions and Recommendations Based on the above evaluation, it is concluded that the proposed calibration technique will provide the necessary accuracy to enable the proper performance of the required ultrasonic examinations. Therefore, relief is recommended as requested. References J References 26, 27, 28, 29, 20, 31, and 32. k 0 4 42

i l 3.5.2 Examination Scheduling i 3.5.2.1 Relief Reauest 1S1-14. Schedulina Raouirements of Tables ) IWB-2412-1 and 1WC-2412-1 for Examination of Class 1 and 2 Pioina Welds. Pinina Suenort.s. and Maior Comeonent Suenorts f code Reauirement With the exception of the examinations that may be deferred until the end of an inspection interval as specified in Table IWB-2500-1, and for steam generator tubing, the required examinations shall have been completed during each successive inspection interval in accordance with Tables IWB 2412-1 and IWC-2412-1. The i inspection period specified above may be decreased or extended by as much as one year to enable an inspection to coincide with a plant outage, within the limitations of IW8 2400(c). j Code Relief Reauest Relief is requested from the~ scheduling requirements of Tables IWB-2412-1 and IWC 2412 1 for the examination of Class 1 and'2 piping welds. Prooosed Alternative Examination The percentages required by. Code will not be met on the second and third period of the first inspection interval. The percentages t for the first period will be higher than allowed by Code. Licensee's Basis for Reauestina Relief l In a Safety Evaluation Report dated November 14,1986,(19) the NRC staff required an accelerated examination program for Class 1 and 2 piping and pipe support field welds and major component (reactor vessel, steam generatorst pressurizer, and reactor coolant pumps) support welds made in'the field. The licensee is required to complete inspection of 100% of the above welds that tre scheduled in the first ISI interval during two consecutive refueling outages following restart. In brder to comply with the requirements of the NRC's I augmented and accelerated field weld program, TVA requests relief from the Code percentage requirements. 43 i t - - - - - - ~ - - N

4' Evaluation an augmented and accelerated program of) examination of field welds Pursuant to 10 CFR 50.55a(g)(6)(ii, the WRC has required i in Class 1 and 2 piping, their supports, and major component sup-ports. This augmented ISI program was required based on concerns i expressed by employees in the quality of the field welding program at Sequoyah. The licensee's compliance with the accelerated ISI plan results in a skewing of the examinations towards the first part of the ten year interval so that any defective welds will be identified early in the interval. Granting relief will not result i in less welds being inspected than requiredt only that scheduling requirements of Tables IWB 2412-1 and IWC-2412-1 will not be met. In this case, it is more prudent to follow the augmented plan required by the NRC staff than the Code. .i Conclusions and Recommendations Based on the above evaluation, it is concluded that for the i f examinations discussed above, the examination plan required by the NRC staff should be followed instead of the ASME Code examination. Therefore, relief is recommer.ded as requested. References References 19 and 32. I l f 1 6 I 4 44 i i . ~.

.1 REFERENCES 4. l.- S. A. Varga (NRC) to G. Williams, Jr. (TVA), June 24, 1977; inservice-Inspection Requirements. ~ 2. J. E. Gilleland (TVA) to S. A. Varga (NRC), July 8,1977;! intent to ' submit 151 plan. ~ i .3. S. A. Varga (NRC) to N. B. Hughes (TVA), December 8,.1978; request for - additional information, preservice and. inservice inspection plans.. k 4. J. E. Gilleland (TVA)_to S. A. Varga (NRC), February 14, 1979; response to request for-additional information. 5. J. E. G111 eland (TVA) to S. A. Varga (NRC), April 19, 1979; Preservice inspection Program, Revision 5. 6. J.E.Gilleland(TVA)toS.A.-Varga(NRC),May 22, 1979; Preservice 1 , Inspection Program, Revision 7. ) 7. L. S. Rubenstein (NRC) to H. G. Parris (TVA), December _11, 1979; review of two PSI issues. 8. L. M. Mills (TVA) to L. S. Rubenstein (NRC), April 24, IS30; Preservice Inspection Program, Revision 9. 4 9. L. M. Mills (TVA) to L. S. Rubenstein (NRC), May 5, 1980; two addi-tional requests for relief from preservice and inservice inspection requirements. 10. L. M. Mills (TVA) to A. Schwencer (NRC), July 31,.1980; request:for 1 elief from Section V calibration block requirements. 11. NRC to TVA,_May 27, 1982; Safety Evaluation of request for relief from hydrostatic testing requirements. 12. L. M. Mills'(TVA) to E. Adensam (NRC), October 13, 1982; request for a deferral of hydrostatic testing of replacement welds. 13. T. M. Novak-(NRC) to H. G. Parris-(TVA), December _ 23, 1982; safety. evaluation of hydrostatic testing of replacement welds. 14. L. M. Mills (TVA) to E. Adensam (NRC), December. 21, 1983; request for j deferral of hydrostatic testing of replacement welds. i 15. T. M. Novak (NRC) to H. G. Parris (TVA), April 12, 1984; safety evaluation of hydrostatic testing of replacement welds, i 16. L. M. Mills (TVA) to E.'Adensam (NRC), July 30L 1984;-request for j deferral of tydrostatic testing of valve replacement welds. i i 45 d ~

l' 17. L. M. Mills (TVA)'to E. Adensan (NRC), September 18, 1984;' revised < ~ -request-for deferral-of hydrostatic testing of valve replacement' welds. 4 18. T. M.'Novak-(NRC) to H. G.-Parris (TVA), October 35,;1984; safety i evaluation of hydrostatic testing of valve replacement welds.- L

19. NRC to TVA, November 14, 1986; Safety Evaluation of TVA's field welding' l

program. 20. R. Gridley (TVA)-to B. J. Youngblood (NRC), July 30, 1986;- commitment for_ submitting ISI progra:a. 21. R. L. Gridley (TVA) to B. J.-.Youngblood (NRC), January 30,.1987;- response to.NRC Safety Evaluation Report dated November 14, 1986. ~ 22. R. A. Hermann (NRC):to S. A. White-(TVA),,May 11, 1988;' Safety evaluation of relief request resulting from microbiologically induced - 4 corrosion._(MIC) in the essential ' raw cooling water (ERCW) system.- 23. Safety Evaluation Report, NUREG-00ll,. March 1979; evaluation of the preservice inspection program. [ 24. Safety ~ Evaluation Report, NUREG-0011, Supplement No. _1, February 1980;: evaluation of.the preservice inspection program. 25. Safety Evaluation Report, NUREG-0011, Supplement Nob 2, August 1980;- augmented examinations. 26. L. M. Mills (TVA) to E.:Adensam (NRC),-' November 2, 1982; Sequoyah Unit 1 inservice inspection program. y 27. D. S. Krammer (TVA) to E. Adensam (NRC),_- August 23, 1983; Sequoyah Unit. 2 inservice inspection program. 7 28. L. M. Mills (TVA) to E.'Adensam-(NRC), 7, 1984;7Revisiont1 of-i Sequoyah Unit 1 inservice inspection pr. March ogram. 29. L. M. Mills (TVA)~to E. Adensam (NRC), September 13, 1984; Revisions 2 through 5 to Sequoyah Unit 1 inservice inspection program, and i Revisions 1 through 5 to Sequoyah Unit 2. inservice inspection'progra,m. t 30. R. Gridley (TVA) to 8. Youngblood (NRC), December 10, 1986; Units 1 and 2 ISI Program, Revision 7. o L 31. R. Gridley (TVA) to NRC, July 30, 1987; Unit 2 ISI Program, Revision 10. 32. R.Gridley(TVh)toNRC, November 9,1988; Unit 2ISIProgram, i Revision 13. 33. E. Adensam (NRC) to H. G. Parris (TVA), November 18, 1985; request for additional information on the inservice inspection program. a 46 1

34. NRC to TVA, March 21.1989;-request for additicnal information on the ISI Program.

35..R. L._ Gridley (TVA) to B. Youngblood (NRC), March 12, 1986; response to request for additional 11nformation.

36. R. Gridley (TVA) to B. Youngblood'(NRC), May 6,1985; supplemental response to request for additional information. 37. C. H. Fox (TVA) to-NRC, May 5,1989; response to request for addit &onal! information dated March 21, 1989.

38. Sequoyah Nuclear. Station Units 1 and 2 Final-Safety Analysis Report,

1974.,

39. NRC to TVA, January'25, 1988; approval for use of Code Cases N-341 ~and N 356.

40. G. G. Zech (NRC) to S. A. White (TVA), March'31, 1988;, Safety: Evalu-ation of Microbiologically Induced Corrosion (MIC) program.

41. Lgna Term Insoection Reauirements for Nuclear Power Plants, EPRI-Report

~ NP-4242, March 1986. -{ i 'l 'I i i 1 1 47 i L

4 g 4 0 4 + y APPEIGIX A 1 REQUIREMENTS OF SECTION XIi 0F THE AMERICAN SOCIETY OF MECHANICAL' ENGINEERS BOILER AND PRESSURE VESSEL CODE 1977 EDITION WITH ADDENDA THROUGH SUMMER 1978) s i 4 a An Emswayee-Ownee Company.

14 = t . APPENDIX A-Requirements of Section XI-of the American Society of Mechanical Engineers Boiler and Pressure Code, 1977 Edition with Addenda through Summer.1978 I A.1 CLASS 1 REQUIREMENTS A.I.1 CATEGORY B-A, PRESSURE-RETAINING WELDS IN REACTOR YESSEL A.1.1.1 Shell Welds, Item B1.10 A.1.1.1.1 Circumferential and Long' tudinal Welds, Items Bl.11;and B1.'12 i ~ All pressure-mtaining circumferential; and longitudinal she11' welds:in. the. reactor _ vessel shall be volumetrically examined in accordance with Figures IWB-2500-1 and -2 over essentially-100% of:their lengths;during;the first inspection interval. Examinations may be performed at or near the end of the. interval. A.1.1. 2 Head Welds, Item Bl.20 A.1.1.2.1. Circumferential and Meridional Head Welds, Items 'B1.21. and Bl.22 : All pressure-retaining circumferential'and meridional' head welds in the reactor vessel head shall be-volumetrically examined'in accordance with' Figure IWB-2500-3 over the accessible portion up to 100% of the weld length during the first inspection interval. The bottom head weids may:be examined at or near the end of'the interval. - A.1.1. 3 Shell-to-Flange Weld,. Item B1'.30 Essentially 100% of the length of'the shell-to-flange weld shall be volumetrically examined in accordance with Figure IWB-2500-4 during-the - i first inspection ~ interval. If the examinations are conducted from the flange face, the remaining examination required to be conducted from the i vessel wall may be performed at.or near the end of each inspection interval. o A.1.1. 4 Head-to-Flange Weld, Item Bl.40 1 Essentially 100% of the length of the head-to-flange weld shall-be volumetrically examined in accordance with Figure IWB-2500-5 'during the first inspection i flange face, the, interval. If the examinations are conducted from the remaining examination required:to be conducted from the vessel wall may be performed at or near the end of each inspection interval. 9 A-1 L.u 1

j i A.1.1. 5 Repair Welds, ' Item B1.50 A.1.1.5.1 Repair. Welds in= the Beltlin'e Region. Item Bl.51: All weld repair areas'in the beltline region shall' be volumetrically, examined in accordance with Figures IWB-2500-1 and -2 during the first inspection interval.- Examinations may be performed at:or near the end of the interval. g b CATEGORY B-8,. PRESSURE-RETAINING WELDS IN VESSELS OTHER THAN) A.1. 2 ' REACTOR YESSELS l A.1.2.1 Shell-to-Head Welds in Pressurizer Vessels, Item B2.10 A.1. 2.1.1 Circumferential' Shell-to-Head Welds,: Item B2.11 All circumferential shell-to-head welds in. the-pressurizer'shall be volumetrically examined ~ in accordance with: Figure IWB-2500-1 over essentially 100% of their let.gth during the.first inspection interval. o A.1.2.1.2 Longitudinal ' Shell -Weld, Item B2.12 One foot of the selected longitudinal shell weld in the pressurizer intersecting the examined circumferential shell-to-head weld shall be t volumetrically ' examined in accordance with Figure-IWB-2500-2 during;the 4 first inspection interval. A.1. 2. 2 Head Welds in Pressurizer Vessels Item B2.20 A.1. 2. 2.1 Circumferential and Meridional Head-Welds, Items B2.21 and B2.22 - One circumferential and one meridional' head weld in the pressurizer shall be volumetrically examined in accordance with Figure IWB-2500-3 over essentially 100% of their lengths during the first inspection interval'. A.1. 2. 3 Head Welds in the Primary Side of the Steam Generators, Item B2.30 l A.1. 2. 3.1 Circumferential and Meridional Head Welds, Items-82.31 and.B2.32 .i All circumferential and meridional head welds in the primary side of F the steam generators shall be volumetrically examined in accordance with-Figure IWB-2500-3 o,ver essentially 100% of their length during the first inspection interval. A-2 ? ~ - - - - - - -. -. - - .--n--.

A.1.2.4 Tubesheet-to-Head Weld. Item B2.'40. l The tubesheet-to-head'welo'in the primary side' of the steam generators shall be volumetrically examined in accordance with Figure IWB-2500-6 over-- essentially 100% of its length during the first inspection interval. A.1. 2. 5 Shell (or Head) Welds-in the Primary Side of'the Heat Exchangers, i Item B2.50 A.1.2.5.1. Circumferential Welds, Item B2.51 ' All circumferential, shell' (or head) welds in the primary: side of the. i I heat exchangers'shall be volumetrically examined in accordance with Figures IWB-2500-1 and -3 over essentially 1005 of their length during the first E inspection interval. A.1. 2.5.2 Longitudinal (or Meridional) Welds, Item'B2.52 i t All longitudinal (or meridional) welds in the primary side of the-hea't: exchangers shall'be volumetrically examined in accordance with Figures. IWB-2500-2 and -3 over essentially 100%.of their length during-the first. inspection interval. . A.1. 2. 6 Tubesheet-to-Shell (or Head)- Welds, Item B2.60 The.tubesheet-to-shell-(or head) welds shall be volumetrically examined: [ in accordance with Figure IWB-2500-6 over essentially 100% of its length during the first-interval. A.1. 3 CATEGORY B-0, FULL PENETRATION WELDS OF N0ZZLES IN VESSELS (INSPECTION PROGRAM B) l b A.1.3.1 Reactor Vessel Nozzle-to-Vessel Welds, Items B3.90 and B3.100 All nozzle-to-vessel welds and inside r'adius sections in-the reactor [ vessel shall be volumetrically examined in accordance with Figure IWB-t 2500-7 during the first interval of operation. The nozzle-to-vessel-weld and adjacent areas of the nozzle and vessel are included. At'least 251 but not more than 50% (credited) of the nozzles shall be examined by the end of the first. inspection period and the remainder by. the end of the inspection interval. If exami. nations are conducted from inside the component.and the nozzle weld is examined by straight beam ultrasonic method from the nozzle bore, the remaining examinations required to be conducted-from the shell' 1 inside diameter may be performed at or near the end of each inspection' interval. a b A-3 4 4- . m m --w _, _ + -,i . ' - - = ~ ,r

A.1. 3. 2 Pressurizer Nozzle-to-Vessel-Welds, Items 83.110 and 83.120-All nozzle-to-vessel. welds and inside radius sections in the pressur-'- izer shall be volumetrically examined in accordance with Figure IWB-2500-7 during the first interval of operation. The nozzle-to-vessel weld and - adjacent areas of the nozzle and vessel are included. At least 25% but not-more than 50% (credited) of the nozzles shall be examined by the end of the - first inspection ' period' and' the remainder by the end of th( interval. inspection A.1.3.3 Steam Generator Nozzle-to-Vessel Welds; Items B3.130 and B3.140 All nozzle-to-vessel welds and inside radius' sections in the primary side of the steam generator shall, be volumetrically examined in accordance with Figure IWB-2500 during the first interval of operation. The nozzle-to-vessel weld and adjacent areas of the nozzle and vessel are included. At least'25% but not' more than 50%' (credited) of the nozzles shall be - examined by the end of the first inspection period.and the remainder byJ the-end of the inspection interval. A.1.3.4 Heat Exchanger Nozzle-to-Vessel ' Welds, Items B3.150 and B3.160_ .i All nozzle-u-vessel welds an'd inside radius sections in the' primary side of the heat exchanger shall be volumetrically examined in accordance - with Figure IWB-2500-7 during the-first interval of operation.' The nozzle-to-vessel: weld and adjacent areas of the nozzle and vessel are included. At least 25% but not more than 50% (credited) of-the nozzles a shall be examined by the end of the first inspection period and-the; } remainder by the end of the inspection interval. i i } A.1. 4 CATEGORY B-E, PRESSURE-RETAINING' PARTIAL PENETRATION WELDS IN VESSELS A.1.4.1 Reactor Vessel Partial Penetration Welds, Item B4.10 L i A.1.4.1.1 Yessel Nozzles, Item B4.12 l The external surfaces of partial penetration welds on 25% of reactor. l i vessel nozzles shall be visually examined (VT-2)- during the first inspec-tion interval. The examinations shall cumulatively cover the specified percentage among each group of penetrations.of comparable size and function. i A-4 j i

i A.1.4.1.2 - Control' Rod Drivo-NozzlescItem 84.13 L L The external surfaces of partial penetration. welds' on'255 of the control rod drive nozzles shall-be visually examined (VT-2),during the. a first inspection interval. The examinations shall, cumulatively cover the: L specified percentage among each group of penetrations,of comparable size j and function. s A.1. 4.1. 3 Instrumentation Nozzles, Item B4.14 ' l i The external surfaces of partial penetration welds on 25% of the L . instrumentation nozzles shall be visually examiled (VT-2) during the first' L . inspection. interval. - The examinations shall curulatively cover the specified percentage among each group of penetrations of. comparable. size. i and function. } l A.1.4.2' Heater Penetration Welds' on the Pressurizer; Item B4.20-The external' surfaces of 25% of the heater p:.netration welds on the pressurizer shall be. visually examined (VT-2) dLring the first inspection; inte rval. The examinations. shall cumulatively cover the specified per-1 centage among each group of penetrations of = comparable; size--and function.. v A.1. 5 CATEGORY B-F, PRESSURE-RETAINING DISSIMILAR METAL WELDS. ~ A.1.5.1 Reactor Vessel Nozzle-to-Safe End Welds, Item B5.10' All dissimilar metal nozzle-to-safe end welds in the reactor vessel shall be surface and volumetrically-examined.in accordance with Figure IWB-2500-8 during the first inspection interval. ;The examinations may'be l-performed coincident with the vessel nozzle examinations. required by Examination Category B-D. Dissimilar metal welds between combinations.of-(a) carbon or -low alloy steels to high alloy steels, (b) carbon or low l-alloy steels to high nickel alloys, and (c) high alloy steel: to'high.' nickel alloys are included. i. l i 1 d A.1.5.2 Pressurizer Nozzle-to-Safe End Welds,- Item 85.20 i All dissimilar metal nozzle-to-safe end welds in.the pressurizer shall be surface and volumetrically examined in accordance with Figure IWB-2500-8 during the first inspection interval. The examinations-may be performed : s coincident with the vessel nozzle-examinations required by Examination Category B-D. Dissimilar metal welds between combinations of (a) carbon or low alloy steels to high alloy steel, (b) carbon or low alloy steel to high nickel alloys, and (c) high alloy steel to high nickel alloys are included. ~ A-5 L J r-,-- w .-r.-

~. l-I -A.1.5.3 Steam Generator Nozzle-to-Safe End Welds, item B5.30 All dissimilar metal nozzle-to-safe end welds is 1:he steam generator shall be surface and_volunstrically examined in' accordance with Figure-IWB-2500-8 during the first inspection interval.-- The examinations may be-performed coincident with the vessel nozzle examinations requined by Examination Category B-D. ' Dissimilar 1 metal welds between combinations'of-(a) carbon or low alloy steels to hi steel to high nickel alloys, and _(c)gh alloy steel, (b) carbon or low alloy . ~ _ high alloy steel to high nickel-al,loys . are included. A.1.5.4 Heat Exchanger Nozzle-to-Safe Ends Welds, Item 85.40 All dissimilar metal nozzle-to-safe end welds in the heat exchangers shall be surface and volumetrically examined in accordance with Figure-IWB-2500-8 during the first -inspection interval. : The examinations may be : performed coincident with the, vessel nozzle' examinations required by = Examination Category B-D. Dissimilar metal welds between combinations-of (a)-carbon or low alloy steels to hi steel-to high nickel alloys, and'(c)gh alloy steel,-(b) carbon or low alloy high alloy steel:to high nickel alloys. are included. A.1.5.5 Piping Safe End Welds, Item B5.50 All dissimilar metal safe end welds in piping shallibe surface and volumetrically examined in accordance with Figure IWB-2500-8 during the first inspection interval. The examinations may be performed coincident J with the vessel nozzle examinations required by Examination Category B-D. 1 Dissimilar metal welds between combinations of (a) carbon.or. low alloy 1 steels to high alloy steel, _(b) carbon or low alloy. steel L to high nickel alloys,.and (c) high alloy steel to high nickel alloys are included. 1 A.1. 6 CATEGORY B-G-1, PRESSURE-RETAINING BOLTING LARGER THAN 2 INCHES. IN DIAMETER y j A.1.6.1, Reactor Closure Head Nuts, Item B6.10 The surfaces of all reactor closure head nuts larger than 2 in. in i diameter shall be examined during the first-ins may be examined (a) in place under tension, (b)pection interval. Bol ting when the, connection;is i disassembled, or (c) when the bolting is removed. Examinations may be performed at or near the end of the inspection interval. 4 I A-6

~ A.1.6.2 Reactor Closure Studs. Items 86.20'and B6.30 ~ ~ All closure studs in' the reactor vessel larger than 2 in. in diameter shall be volumetrically examined in accordance with Figure !WB-2500 duringtthe first inspection interval. A surface examination is also-required when the studs are removed. Examinations may be. performed ~.at or near:the end of the inspection interval.- A.1.6.3 Ligaments Between Stud Holes in the Reactor' Vessel, item B6.40 All ligaments between stud. holes in the reactor vesselishall bei. volumetrically examined in accordance with Figure IWB-2500-4 during the -. first inspection interval. Examination includes threads in base metal and is required only when the connection is disassembled. Examinations _ may be - performed at or near the end of the inspection interval. 1 A.1.6.4 Reactor Closure Washers and Bushings, Item B6.50 The surfaces of all closure washers-and bushings.on bolting larger than q 2 in.- in diameter in the reactor vessel shall:be visually examined (VT-1) during the first inspection interval. _ The examinations are required only-ci when the connection is disassembled and may be performed at or near the end-j of the inspection interval.- l I 1 A.1.6.5 Pressurizer Bolts-and Studs, Items B6.60 and B6.70 l All bolts and studs larger than 2 in.' in diameter in' the pressurizer, i l shall be volumetrically examined in accordance with Figure IWB-2500-12 1 during the first inspection -interval. A surface' examination is' also .i required when the bolts and studs are removed. _ Examinations 'may be performed at or near the end of the inspection interval, q l A.1.6.6 Pressurizer Bolting, Item B6.80 1 The surfaces of all nuts, bushings,' washers, and threads on bolting larger than 2 in. in diameter in base material and flange ligaments between threaded stud holes in_ the pressurizer shall be visually _ examined (VT-1) 1 -during the first inspection interval. ' Bolting may be examined (a) in place. 1 under tension, (b) when the connection is disassembled, or1(c) when the. j bolting is removed. Bushings, threads, and ligaments in base material of g flanges are required to be examined only when the connections are dis-assembled. Bushings may be inspected in place. Examinations 'may be i performed at or near the end of the inspection interval. l [ 1 A.1.6.7 Bolts and Studs, in Steam Generators, Items B6.90 and, B6.100 All bolts and studs larger than 2 in. in-diameter in base materials.of steam generators shall be volumetrically examined in accordance with Figure IWB-2500-12 during the first inspection interval. A surface examination is I. A-7 i v.,--e e v w -we-+, --a c +~. - - - - ---w--~-w ~~-~

also required when the bolts and studs; are removed., Examinations may be

)

performed at or near the end ofithe inspection. interval. - A.1 6.8. Bolting in' Steam Generators, Item B6.110 The surfaces of ~ allinuts, bushings, washers, and1 threads:in bolting _ larger % n 2 in. in diameter in base material and flange ligaments between threadei stud holes -in steam generators shall1be visually examined (VT-1)- 3 during the first inspection interval.' Bolting may be examined (a)-in place. under tension, (b) when-the connection'is disassembled, or-(c) when-the 3 bolting is removed. ' Bushings, threads, and ligaments in base materials-of flanges are required to be examined only when the. connections are dis-assembled. Bushings may be: inspected in place.. Examinations may be. i performed at or near the end of the inspectionLinterval.- A.1.6.9 Bolts and Studs,-in Heat Exchangers, Items B6.120 an'd 86.130 , All bolts _ and studs larger than'2. in. ~in-diameter in base materials of ? heat exchangers shall be volumetrically examined in accordance:with Figure IWB-2500-12 during. the first inspection interval. _ A surfaceLexamination is - also required when the bolts and studs are removed. Examinations may be performed at or near:the end of the inspection interval.. Examinations are 1 limited to-bolts and studs on components selected for examination. under. Examination Categories B-B, B-J, B-L-1, and B-M-1, as applicable. t l A.1.6~.10 Bolting in Heat Exchangers, Item B6.140 The surfaces of all nuts, bushings, and threads:on; bolting largerithan-l 2 in, in diameter in base material and flange ligaments-between threaded ~ stud holes in heat exchangers shall-be visually examined (VT-1) during the first inspection interval. Bolting may be examined (a) in place under l tension, (b) when the connection is disassembled..or.(c) when the bolting is removed. Examinations may be performed at or near the end of,the in-spection interval. Bushings, threads, and ligaments in base materials; of - flanges are required to be examined only when the connections are disassembled. Bushings may be' inspected in place. Examinations-are limited to bolts and studs on components selected for examination under + Examination Categories B-B, B-J, B-L-1, and-B-M-l', as applicable. I A.1.6.11 Bolts and Studs in Piping, Items 86.150 and B6.160 All bolts.and studs larger than 2 in. in diameter :in piping shall be voltanetrically examined in accordance with Figure IWB-2500-12 during the. first inspection interval. A surface examination is also required when the bolts and studs are removed. Examinations may be performed at or near the~ end of the insRection interval. I A-8

I i l'. A.1.6.12 Bolting in Piping,'Itea 86.170~ .The surfaces of all nuts, bushings','.washersi and threads on boltir.g' larger than 2 in. -in diameter in base material and, flange ligaments between threaded stud-holes in1 piping shall;be visually examined-(YT-1).during ther first inspection ' interval.. Solting:may be examined (a) in place under i tension (b) when the connection 1s-disassembled, or (c) when the bolting are required to be examined only when the' connections are disassembled. ' '[ is removed.. Bushings, threads, and ligaments in base > materials of flanges Bushings may be inspected in place. Examinations nay.be performed at or 1 near the end'of the inspection interval. j A.1.6.13 Bolts and ' Studs in Pumps, Items B6.180 and B6.190' All bolts and studs larger than 2 in. in diameter in pumps shall be 'l ? volumetrically examined in accordance with Figure IWB-2500-12 during the.. 4 first inspection interval.t. A surface examination is also. required when the bolts and studs are removed. Examinations may,be performed at orinear the endc of the inspection interval.. Examinations are ' limited to bolts and studs on components selected'for examination under Examination Categories. B-8,. B-J, B-L-1,. and B-M-1, as appli cable. A.1.6.14 Bolting in Pumps, Item B6.200

t The surfaces of all nuts, bushings, washers, and threads in bolting

~ L larger than 2 in in diameter in base material and-flange ligaments between threaded stud holes in pumps'shall be visually examined (VT-1) L(a)< in' place under tension, (b) when the connection isL disassembled, or (c) when the 1 bolting is removed. Examinations may be performed at or near the end-:of-the inspection interval. Bushings, threads, and ligaments in base materials of flanges are required to be examined only when the connections ~ are disassembled. Bushings may be inspected in-place. Examinations are 4 limited to bolts and studs on components selected for-examination' under Examination Categories B-B, B-J, B-L-1, 'and B-M-1,- as applicable.: i l A.1.6.15 Bolts and Studs in Valves, Items B6.210 and B6.220 i All bolts and studs larger than 2 in. in diameter in valves shall be i volumetrically examined in accordance with Figure IWB-2500-12:during the first inspection interval. - A surface examination lis also required when the bolts and studs-are removed.-- Examinations may be performed at or near the 1 end of the inspection interval. Examinations are limited to bolts and-1 I studs on components selected for examination under Examination Categories B-B, B-J, B-L-1, and B-M-1, as applicable. .4 A.I.6.16 Bolting in Valves, Item B6.230 The surfaces of all nuts, bushings, washers, and threads in bolting larger than 2 in. in diameter in base materials and flange ligaments between threaded stud holes in valves shall be visually examined (VT-1) l A-9 i. .\\ i

y . Bolting may be' exa;ained (a) in place: j - during the first; inspection. interval ~. under tension, (b) when the connection 11s' disassen61ed, or (c)Lwhen the'. . bolting is removed. Examinations may be perfomed at or near the end.of i the inspection' interval. Examinations are limited to bolts and studs. on. 1 components selected for examination under Esamination Categories.B-B,. B-J.- 1 B-L-1,- and B-M-1, - as: applicable. d .A.1.7 CATEGORY B-G-2, PRESSURE-RETAINING ' BOLTING 2 INCHES AND SMALLER 1 l~ IN DIAMETER j i A.1. 7.1 - Bolts. Studs, and Nuts in Reactor Vessel, Item B7.10 The_ surfaces of all bolts, studs ~and nuts 2 'in. or less -in-diameter in1 the reactor vessel shall be visually. examined (VT-1) during the first 9 inspection ' interval. Bolting may be examined (a) in place. under tension, (b) when the connection is disassembled, or-(c) when the bolting is. removed. A.1 ; 7. 2 Bolts Studs, and Nuts. in Pressurizer, Item 37.20 The surfaces of all bolts' studs, and' nuts-2 in.; or less:in~ diameter in the pressurizer shall be visually examined (VT-1);during the:first-in-' l spection' in terval. Bolting may be examined (a) in place'under tension, (b) when the connection is-disassembled, or (c) when the bolting is removed. .i l-A.1.7.3 Bolts, Studs, and Nuts in Steam Generators, Item B7.30 .The surfaces of all bolts, studs, and nuts 2 in.'or less-in diameter in the steam generators shall be visually examined (VT-1)' during.the first. p inspection interval. Bolting may be examined (a) in place under tension, (b) when the connection is' disassembled,!or (c) when the bolting _is removed. l A.1.7.4 Bolts, Studs, and Nuts in Heat ~ Exchangers, Item B7.40 1-The surfaces of all bolts, studs, and nuts 2 in. or less. in diameter in the heat exchangers shall' be -visually examined (VT-1) during-the first inspection interval. Bolting may be examined (a) in place under tension, (b) when the connection is disassembled, or (c) when the bolting-is. removed. A.1.7.5 Bolts, Studs, and' Nuts in piping, Item B7.50 The surfaces of all bolts, studs, and nuts 2 in. or less in diameter' r in piping shall be visually examined (VT-1) during the first inspection i interval. Boltinginay be examined (a) in place under tension, (b) when the connection is disassembled, or (c) when the bolting is removed. i A-10

The surfaces of a11' bolts, ^ studs end tiuts 2 in, or less in = diameter in pumps shall be visually examined (,VT-1)- during the i:rst inspection

interval, Bolting may be examined. (a) in p1 ace under tension, (b') when-

.the connection is disassembled,:or (c) when the bolting is removed. j A.1,7. 7 Bolts Studs', and Nuts in Valves. Item B7.70-3 The: surfaces of-~all bolts,* studs, and nuts.2 in. or lessiin diameter in valves shall be visually examined (VT-1) during the first inspection inter-val. Bolting may be examined (a) in place:under tension,-(b) when the. connection is disassembled, or (c) when the bolting is removed. A.1. 8 CATEGORY B-H, VESSEL SUPPORTS i l A.1.8.1 Integrally Welded Attachments in Reactor Vessel, Item-B8.10 The' attachment weld ' joining the; reactor; vessel support to the! pressure-retaining membrane of the reactor vessel where the' support base material 3 design thickness is 5/8 in. or' greater shall be surface or volumetrically t examined, as applicable.. in accordance with Figures -IWB-2500-13, -14,. and during the first inspection interval. Weld buildupsion nozzles 'that i serve as supports are ex1cuded. The examination includes essentially 100%. of the length of the weld to the reactor vessel and the integral attachment weld to a cast or forged integral attachment.to the reactor. vessel, as applicable. One hundred percent of the welding of each lug on the vessel. is included in the examination. A.1.8.2 Integrally Welded Attachments in ' Pressurizer, Item B8.20 - The attachment weld joining the pressurizer support to the pressure-retaining membrane of the vessel where the support base material design thickness is 5/8 in, or greater shall be surface or. volumetrically examined, as applicable, in accordance 'with Figures IWB-2500-13 -14', and -15 during the first inspection interval. Weld buildups on; nozzles that ' serve as supports are excluded. The examination-includes ' essentially 100% of the length of the weld to the pressurizer and the integral attachment weld to a cast or forged integral attachment to the pressurizer, as applic-able. One hundred percent of the welding of each lug on.the' pressurizer is; included in the examination. A.1.8.3 Integra11y' Welded Attachments in Steam Generators, Item B8.30 The attachmdnt weld joining the steam generator support to the-pressure-retaining membrane of the vessel where the support base material design ' thickness is 5/8 in. or greater shall be surface or volumetrically A-11 .1

1 examined,' as applicable. Jin accordance with Figures IWB-250N13 -14, 'and - c -15 during the first. inspection interval. Weld buildups oa ;ozzles that-- - H serve as supports are excluded... The examination includes essentially 100%- t of the length of. the weld to the steam generator and the integral attach-l ment weld to a cast.or forged integral-attachment to the steam generator, as applicable. One hundred percent of the welding of each lug on the steam generator is -included in -the examination. The examination is-limited to 1 the attachment weld on one steam generator.- i l A.1'.8.4 Integrally Welded Attachments in Heat Exchangers Item B8.40' The attachment weld joining the' heat-exchanger support to the pressure-- l retaining membrane of the vessel where the support base-material-design l thickness is 5/8 in. or. greater shall be surface 'or-volumetrically- ~ examined _ as applicable, in accordance.with, Figures. IWB-2500-13 -14,: and' -15 during the first inspection _ interval. Weld' buildups on-nozzles' that '- - serve as. supports are excluded.; The examination-includes essentially 100%- . of the length of the weld to the heat 1 exchanger and the integral _ attachment weld to a cast or forged integral attachment to the heat exchanger,:as applicable. - One hundred' percent of-the welding of each21ug on the heat. exchanger is included in the examination.; The examination isilimited to the attachment weld on one heat' exchanger. A.1. 9 CATEGORY B-J, PRESSURE-RETAINING WELDS IN. PIPING - A.1.9.1 Nominal Pipe Size 4 In. and Greater, Item B9.10: i. A.1.9.1.1 Circumferential Welds, Item B9.11' L For circumferential welds in pipe of nominal pipe size'4 in, and greater, surface plus volumetric examinations shall be performed in' ac-- -l cordance with Figure IWB-2500-8 over essentia11y 100% of:the-weld length during each inspection interval. The examinations shall include the j following: (a) All terminal ends in each pipe or branch run connected to = vessels. (b) All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed the following n ifmits under loads associated with specific seismic events and operational conditions. (1 ) primary plus secondary stress intensity of 2.4Sm for ferritic steel and austenitic steel, and (2) cumulative usage factor U of 0.4. 3-l A-12

h a i (c) Allt dissicilar metal welds between combinations of i (a) carbon or low alloy steels to high alloy steels; P (b) carbon or low alloy steels to high nickel alloys; and; j l-(c) high alloys steels-to high nickel-alloys. l (d), Additional piping welds so that the total equals 25% of the cir-cumferential joints. in the reactor coolant piping _ system. This t l total 'does not include welds excluded by IWB-1220.. These addi-l tional welds may be located in one loop (one loop is currently. l defined for both PWR and BWR plants in the 1977 Edition). j i 1.ongi tudinal Wel ds,' Item B9.12 - - A.1. 9.1. 2 9 For longitudinal: welds in pipe of nominal ' pipe size 4 in, and greater, I surface plus volumetric examinations shall be. performed in accordance with-- Figure > IWB-2500-8 for at least a pipe-diameter. length, but not more: than 12-in, of each longitudinal weld intersecting.the circumferential welds. required.to be examined. -A.1.9.2 Nominal Pipe Size Less Than 4 In., Item B9.20. A.1. 9. 2.1 Circumferential' Welds, Itet 89.21 a For circumferential welds in pipe of ' nominal pipe size less than 4 in., surface examinations shall; be performed in'accordance with Figure.IWB-3 2500-8 over essentially 100% of the weld length during each inspection - ~ interval. The examinations shall. include the following:- (a) All terminal ends in each pipe or branch run connected to. vessels. (b) All terminal ends and joints in-each pipe or branch' run connected to other components where the stress -levels exceed the-following-t i limits.under loads associated with specific seismic events and, operational conditions. l l< (1) primary plus secondary stress intensity of-2.4Sm for ferritic l steel and austenitic steel, and, i (2) cumulative usage' factor U of 0.4. l (c) All dissimilar metal welds between combinations of: ] -t (a) carbon or low alloy steels to high alloy, steels; (b) carbon or low alloy steels to high nickel alloys;.and (c) high alloys steels to high = nickel alloys. (d) Additionaf. piping welds so that the total' equals 25% of the cir-cumferential joints in the reactor coolant piping system. This total does not include welds excluded by IWB-1220. These addi-tional welds may be located in one-loop (one loop is current 1y' defined for both PWR and BWR plants in the 1977 Edition), f 1 A-13 1 ....... ~ ,._...,w

~ ~ - A.1.9.2.2 longitudinal Welds, Item B9.22. For longitudinal welds in pipe of nominal pipe size less than ~4 in., - surface examinations shall be performed in accordance with Figure IWB 2500-8 for at least a pipe-diameter length, but not more than 12 in of1 each longitudinal weld intersecting ~ the circumferential welds required to be examined.- + 2 e A.1.9.3 Branch Pipe Connection ' Welds. Item B9.30-- A.1. 9. 3.1 Nominal Pipe Size. Greater Than 2 In.,-Item B9.31 For welds in branch connections greater than 2 in., surface plus volumetric examinations shall be performed in accordance.with Figures - IWB-2500-9. -10 and -11 over essentially 100% of the weld ' length during each f a.spection interval. The examinations shall include the following - (a) All terminal ends in' each pipe or branch run_ connected to vessels'. (b) All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed the following-limits under loads associated with specific-seismic events and- 'l operational conditions. li (1) primary plus secondary stress; inter.sity of 2.4Sm for ferritic L q steel and austenitic steel,. and .j 1 ( 2) cumulative usage factor U of 0.4. 1 (c) A11~ dissimilar metal welds between combinations of: t (a) carbon or low alloy steels.to =high alloy steels; (b) carbon or low alloy-steels -to-high nickel alloys; and . i (c) high alloys ' steels to high nickel alloys. i (d) Additional piping welds so that the total equals 25% of the cir-1 cumferential joints in the reactor coolant piping system.- This 4 total does not include welds excluded by IWB-1220. These,addt-tional welds may be located in one loop (one loop is currently. defined for both PWR and BWR plants in the 1977 Edition). i l 4 a A-14

q i ' A.1.9.3.2 Nominal Pipe Size Less Than or. Equal; to In.. Item 39.32 For welds in branch pipe connections less than or-equal to 2 in.. ~ surface examinations _ shall-be performed in accordance with Figures IWB-2500-9,.-10, and -11 over essentially 100% of the weld length during each inspection interval.. The examinations shall; include the following (a) All terminal ends in. each ' pipe or branch run connected to vessels.- l -l (b) All terminal ends and joints in each pipe or branch run connected to other components where the stress levels-exceed the following limits under loads associated with specific seismic events and' ' operational conditions. (1 ). steel and austenitic steel : and i primary plus secondary stress intensity, of 2.4Sm for ferritic ( 2) cumulative-usage factor U of 0.4. .(c) All dissimilar metal welds between combinations;of:- (a) carbon or low alloy steels.' to high alloy steels; (b) carbon or_ low alloy steels to high nickel alloys; and (c) high alloys steels to high nickel alloys. I (d) Additional piping welds so that the total equals I25% off the cir-cumferential joints in the reactor coolant piping system. Thi.s - I total does not include welds excluded by -IWB-1220. These.addi-tional welds may be-located in one loop (onc' loop is currently. l defined for both PWR'and BWR plants in the.1977 Edition). 1 ~ l A.1.9.4 Socket Welds, Item B9.40 Socket welds shall be surface examined over essentially 100% of the weld length during each inspection interval. The' examinationsT shall include the following:

l (a) All terminal ends in each pipe or branch r'un connected to vessels.

(b) All terminal ends and' joints in each pipe or branch run connected t to other components where the stress levels exceed the following limits under loads associated with specific seismic events and-j operational conditions. 1 (1) primary plus secondary stress intensity of 2.4Sm' for ferritic l steel,and austenitic steel, and. l-t ( 2) cumulative usage factor U of 0.4. l A-15 1: . I'

q e i (c) - All ' dissimilar metal welds between combinations of: l t (a) carbon orclow alloy steels to high alloy steels; a (b) carbon or. low alloy? steels ' to high nickel alloys; and-1 ~ (c) high. alloys steels to high ' nickel alloys... (d). Additional' piping welds so that the total equals 25% of the cir-- 4 cumferential: joints in the reactor coolant piping system.' This total does not-include welds excluded by IWB-1220.. These addi = tional welds may be' located in one loop (one loop is currently' defined for'both PWR and _BWR plants in the 1977 Edition). i 2 A.1.10 CATEGORY B-K-1, SUPPORT MEMBERS FOR' PIPING,' PUMPS, AND VALVES., [ 1 a A.1.10.'1' Integrally Welded Attachments on Piping.- Item 810.10 Volumetric.or surface examinations, as applicable, per Figures, IWB-2500-13, -14, and'-15 of essentially 100% of. the, weld length are r required for all integrally welded support attachments of piping required. to be examined by Examination Category B-J. Only those: attachments whose base material design thickness is 5/8-in, or greater need.to be examined.. i A.1.10.2 Integrally Welded Attachments on Pumps', Item 810.20-- 1 Volumetric or surface examinations, as applicable, per Figures IWB-2500-13, -14, and'-15 of essentially.100% of. the weld length are required for. all welded support attachments.of; pumps' integral;to piping. required to be examined by Examination Category B-J. Only Lthose; attach-ments whose base material design thickness is-5/8 in. or greater'need..to be examined. 1 A.1.10.3 Integrally Welded Attachments on Valves, Item B10.30. Volumetric or surface examinations, as applicable,'per Figures IWB-2500-13 -14, and -15 :of essentia11y'100% of the weld length are required.for all welded support attachments of valves integral. to piping-required to be examined by Examination Category B-J. Only those attach-- ments whose base material design thickness is 5/8 in or greater need to-be examined. 4 L 4 A-16

i A.1.11 CATEGORY 8-K-2, COMP 0 NEWT SUPPORTS FDR PIPING, PUMPS, AND VALVES ) A.1.11.1 Component' Supports for Piping. Item ' B11.10 f All component supports for piping shall be visually examined (VT-3, 4) in accordance with IWA-2210.3 and IWA-2210.4 during each inspection in-terval.. The component supports extend from piping to, and include the-attachment to, the~ supporting structure. The settings of snubbers, shock 4. abscrbers, and spring-type hangers shall be verified. 1 1 A.1.11.2 Component Supports for Pumps. -Item B11.20 All component supports for pumps shall be visually examined (VT-3[4) in accordance with IWA-2210.3 and IWA-2210.4 during each-inspection in-. terval. The component supports extend from piping to, and' include the. attachment to, the supporting structure. ; The. settings of snubbers, shock. absorbers, and spring-type' hangers'shallibe verified. 1 A.1.11.3 Component Supports for Valves. Item B11.30-All component supports for valves.shall be visually examined (VT-3, 4) in accordance with IWA-2210.3 and IWA-2210.4 during each inspection in-terval. The component supports extend from piping to, and include the. attachment to, the supporting structure. The settings of snubbers, shock absorbers, and spring-type hangers shall be verified. A.1.12 CATEGORIES B-L-1 and B-M-1, PRESSURE-RETAINING WELDS IN PUMP CASINGS AND VALVE BODIES, AND B-L-2-and B-M-2,' PUMP CASINGS AND-i VALVE BODIES l A.1.12.1 Pump Casing Welds, Item B12.10 I Essentially 100% of the pressure-retaining welds in at least one pump in each group of pumps performing similar functions <in the system (e.g., recirculating coolant pumps) shall be surface and volumetrically examined in accordance with Figure IWB-2500-16 during each inspection interval. The examinations' may be performed at or near the end-of the inspection interval. 4 9 A.1.12.2 Pump Casings, Item B12.20 The internal surfaces of at least one pump in each group of pumps per-forming similar furictions in the s shall be visually examined (YT-1) ystem (e.g., recirculating coolant pumps) during each inspection interval. The examination may't* performed on the same pump selected for volumetric examination of welds. The examinations may be performed at or near the end i of the inspection interval. i A-17 4 ~ ___---__,.__,2.___ t

+ u l A.1.12.3 valve Body Welds. Itea B12.30 : Essentially:1 DOS of the pressure-retaining welds >in at least one valve in each group of valves with the same construction design (e.g., globe, - gate, or check valve) and manufacturing method that perform similar func-tions in the system (e.g., containment' isolation and system overpressure protection) shall be surface and, volumetrically examined in accordance with_. Figure IWB-2500-17 during each inspection interval. The examinations may: be performed at or near the end of:the inspection-interval. A.1.12.4 Valve Body Exceeding 4 In. Nominal Pipe Size, Item B12.40' ~ The. internal surfaces of at least one valve in-each group of valves with the same construction' design (e.g., glote, gate, or check valve) and manufacturing: method that perform similar functions in the system (e.g... containment isolation and system overpressure protection);shall belvisually examined (VT-1) during each inspection interval. The examination may be: performed on the same valve selected for volumetric' examination of welds. The examinations may be performed at or near the end of the inspection interval. A.1.13 CATEGORIES B-N-1, INTERIOR OF REACTOR VESSEL; B-N-2, -INTEGRALLY-WELDED CORE SUPPORT STRUCTURES-AND-INTERIOR ATTACH 4ENTS TO REA VESSELS; and B-N-3, REMOVABLE' CORE SUPPORT STRUCTURES .7 A.1.13.1 Reactor Vessel Interior. Item B13.10 The accessible areas of the reactor vessel interior, including the spaces above and below the reactor core that are made ' accessible by removing components during normal refueling outages, shall be visually examined (VT-3) during the first refueling outage:and subsequent refueling ' outages at approximately 3-year intervals, a A.1.13.2 Boiling Water Reactor Vessel Interior Attachments, Item B13.20' 1 The accessible welds in the reactor vessel interior. attachments shall be visually examined (VT-1) during each inspection interval. The examina-tions may be performed at or near the end of the inspection interval. I A.I.13.2.1' Boiling Water Reactor Core Support Structure, ' Item B13.21 The accessible surfaces of the core support structure shall be visually examined (VT-1) during each inspection interval. 3 Tne examinations may be 1 performed at or,near the end of the inspection interval. A-18 ll 11

f A.1.13.3 Core Support structure for Pressurized' Water Reactor Vessels. 3 Item 513.30 ? The accessible welds and surfaces 'of the core support structure shall be visually examined (VT-3)- each inspection interval. The structure shall be removed from the reactor vessel for examination. The examinations may be performed at or near the end of the inspection interval. 1 y l A.1.14 CATEGORY B-0, PRESSURE-RETAINING WELDS IN CONTROL R0D HOUSINGS -l A.1.14.1 Welds in Control Ro'd Drive Housings. - Item B14.10 The welds in 10% of the peripheral control rod drive housings 'shall be surface or volumetrically examined in accordance with Figure IWB-2500-18 i o during each inspection interval. The examinations may_ be performed at or near the end of the inspection; interval.. I A.1.15 CATEGORY B-P, ALL PRESSURE-RETAINING COMPONENTS A.1.15.1 Reactor Vessel Pressure-Retaining Boundary, Item B15.10 The reactor vessel pressure-retaining boundary shall.be visually examined (YT-2) during the system leakage test performed in accordance with IWB-5221 during each refueling outage. -The. entire pressure-retaining 4 boundary of the reactor coolant system is subject to system pressure tests conducted in accordance with IWA-5000.with the exceptions specified=in : IWA-5214 when system pressure tests are' conducted for repaired,> replaced,- or altered components, q l. A.1.15.1.1 Reactor Vessel Pressure-Retaining Boundary, Item B15.11 The reactor vessel pressure-retaining boundary shall be visually examined (VT-2) during the system hydrostatic. test performed in accor-dance with IWB-5222 once per inspection interval. The examinations may be performed at or near the end of the inspection interval. The entire pressure-retaining boundary of the reactor coolant system is subject to system pressure tests conducted in accordance with IWA-5000 with the-exceptions specified in IWA-5214 when system pressure tests are conducted i for repaired, replaced, or altered components. A.1.15.2 Pressurizer Pressure-Retaining Boundary. Item B15.20 The pressuriper pressure-retaining boundary shall be visually examined i (VT-2) during the system leakage test performed in accordance with IWB-5221 during each refueling outage. The entire pressure-retaining boundary of the reactor coolant system is subject to system pressure tests conducted 4 A-19 i ~ -

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l

in accordance with IWA-5000 with the exceptions specified in1WA-5214 when system pressure tests are conducted-for repaired, replaced. or altered components. A.1,15.2.1. Pressurizer Pressure-Retaiding Boundary, Item B15.21 The pressurizer pressure-retaining boundary shall be visually examined- -(VT-2) during the system hydrostatic test performed'in accordance with IWB-- 5222 once per inspection: interval. The examinations may be-performed at or. near the end of the inspection interval. The entire pressure-retaining bound.ry of the reacter coolant system is subject to. system pressure tests-conducted in acaordance' with IWA-5000 with the exceptions specified in IWA- . 5214 when_ system pressure tests are conducted for repaired, replaced,' or-altered components. A.1.15.3 Steam Generator Pressure-Retaining Boundary, ' Item B15.30- .The steam generator pressure-retaining boundary shall belvisually examined (VT-2) during the system leakage test performed inLaccordance:. with IWB-5221 during each refueling outage. The entire pressure-retaining boundary of the ' reactor coolant-system.is. subject to system pressunt tests -. conducted in accordance with IWA-5000 with the exceptions specified in IWA-5214 when system pressure -tests are conducted for repaired replaced,-or altered components. 1 i A.1.15.3.1 Steam Generator Pressure-Retaining Boundary, Item.B15.31 j The steam generator pressure-retaining boundary shall be _ visually: ) examined (VT-2) hring the system. hydrostatic -test performed in accordance-with IWB-5222 during each refueling outage.. The examinations may be per-- formed at or'near the end of the inspection interval.:.The entire pressure-retaining boundary of the reactor coolant system is subject to system i pressure tests conducted in accordance' with IWA-5000 with the exceptions 1 specified in IWA-5214 when system pressure tests are conducted for repaired, j replaced, or altered components. i A.1.15.4 Heat Exchanger Pressure-Retaining Boundary, Item B15.40 The heat exchanger pressure-retaining boundary shall' be. visually examined (VT-2) during the system leakage test performed in accordance = l l with IWB-5221 during each refueling outage..The eretire pressure-retaining. boundary of the reactor coolant system is subject to system pressure tests conducted in accordance with IWA-5000 with the exceptions-specified in IWA-5214 when system prtssure tests are conducted for repaired, replaced, or y altered components. 4 i A-20 I

A.1.15.4.1 Heat Exchanger Pressure-Retaining Boundary. Item B15.41 The heat exchanger pressure-retaining boundary shall be visually examined (VT-2) during the system hydrostatic test performed in accordance with IWB-5222 once per inspection interval. The examinations may.be per-formed at or near the end of the inspection interval. = The entire ~ pressure-retaining boundary of the reactor coolant system is-subject lto ' system pressure tests conducted in accordance with IWA-5000 with the : exceptions specified in IWA-5214 when system pressure tests are conducted for repaired, replaced. or altered components. A.1.15.5 Piping Pressure-Retainihg Boundary. Item B15.50 The piping: pressuro-retaining boundary shall be: visually examined ~ . (VT-2) during the: system leakage test performed in accordance with IWB-5221 during each refueling outage. The entire-pressure-retaining boundary of the reactor coolant system is subject to system pressure tests conducted in accordance with IWA-5000 with the exceptions specified in IWA-5214 when : system pressure tests are conducted for repaired, replaced or altered components. A.1.15.5.1 Piping' Pressure-Retaining Boundary, Item: B15.51 1 1 The piping pressure-retaining boundary shall be visually examined (VT-2) during the system hydrostatic test performed in accordance with 1 1 IWB-5222 once per inspection interval. The~ examinations may be performed at or near-the end of the inspection interval. The entire-pressure-J retaining boundary of the reactor coolant system is subject to system pressure tests conducted in accordance with IWA-5000 with the exceptions a specifled in IWA-5214 when system pressure tests are conducted for. repaired, replaced, or altered components.- A.1.15.6 Pump Pressure-Retaining Poundary, Item B15.60 - The pump pressure-retaining boundary shall be visually examined (VT-2) during the system leakage test performed in accordancecwith IWB-52211during-1 each refueling outage. The entire pressure-retaining boundary of the a reactor coolant s3 stem is subject to system pressure tests conducted in accordance with IWA-5000 with the exceptions specified in IWA-5214 when j system pressure tests are conducted for repaired, replaced, or altered j components. j A.1.15.6.1 Pump Pressure-Retaining Boundary, Item B15.61 The pump pressure-retaining boundary shall be visually examined (VT-2) durir.g the system hydrostatic test performed in accordance with IWB-5222 once per inspection interval. The examinations may be performed'at or' near the end of the inspection interval. The entire pressure-retaining boundary of the reactor coolant system is subject to system pressure tests conducted. A-21 b

~ fj m in' accordance.with IWA-5000 with the exceptions specified in IWA-5214 when - system pressure: tests are conducted.for repaired,- replaced, Mr altered. components. - -i A.1.15.'7L Valve Pressure-Retaining Boundary, Item B1530 The valve pressure-retaining boundary shall: be visually examined. (YT-2), during the system leakage test performed in accordance with IWB-5221Lduring: each refueling-outage. _ The entire pressure-retaining boundary of the reactor coolant system is. subject to. system pressure tests conducted in - accordance with IWA-5000 with the exceptions specified:in IWA-5214 when, system pressure tests. are conducted for, repaired, replaced.: or altered - componen ts..- A.1.15.7.1 Valve Pressure-Retaining Boundary, B15.71 The valve pressure retaining boundaryL shall: be ; visually examined (VT-2): -during the system hydrostatic test performed in accordance with'IWB-5222 once per inspection-interval.. The examinations may be performed at or near: the end of the inspection interval. The entire pressure-retaining boundary of the reactor = coolant system is subject to system pressure tests conducted-in accordance with IWA-5000 with the exceptions' specified in'IWA-5214.when-3 system pressure tests are conducted _ for: repaired, replaced,;or altered ' components. l l l A.1.16 CATEGORY B-Q, STEAM GENERATOR TUPING q l A.1.16.1 Steam Generator Tubing, Straight Tube Design, Item B16.'10 - L The entire length of the steam generator tubing shall be volumetrically-examined in 3% of the heating surface in each generator during the first ~ inspection interval. The heat transfer surface is,specified in terms of the number of tubes to be examined. i A.1.16.2 Steam Generator Tubing,. U-Tube Design, Item B16.20 Steam generator tubing-(hot leg side),- U-bend portion, and cold leg side (optional) shall be volumetrically examined 11n 3% of the; heating surface in each generator during the first inspection interval.. ~ 4 ~ A-22 E . ~..

+ ~ A.2' CLASS 2 REQUIREMENTS A. 2.1 CATEGORY C-A, PRESSURE-RETAINING WELDS IN PRESSURE, VESSELS. '_ A. 2.1.1 Shell Circumferential Welds,. Item C1.10 Essentially 100% of.the shell' circumferential welds at gross structural discontinuities shall:be volumetrically' examined in accordance-with Ftgure IWC-2520-1 during-each inspection-interval.. A gross structural discontinu ' ity is defined in NB-3213.2.. Examples are junctions-between shells of' different thicknesses, cylindrical shell-to-conical ^ shell. junctions,. and'. shell- (or head)-to-flange. welds, and head-to-shell--welds. For multiple: -f vessels dth similar design, size, and service (such= as steam generators and heat exchangers),4 the required examinations may be limited to one vessel 1or distributed among the vessels-3 A.2.1'.2 Head Circumferential Weld, Item C1.20: .{ Essentia11y'100% of the circumferential head-to-shell. weld shall be' i volumetrically examined in accordance with Figure!!WC-2520-1 during each. inspection interval.. For multiple vessels with'similar defign, size, and i service' (such as steam _ generators and heat exchangers), 'the required exami-nations may be limited to-one' vessel or-distributed among the vessels. - A. 2.1'. 3 Tubesheet-to-Shell Weld, Item C1.30 f Essentially 100% of. the tubesheet-to-shell weldishallf be volumetrically examined in.accordance with Figure IWC-2520-2 during:eachlinspection ~ interval. For multiple vessels with similar. design,. size ;and service (such as steam generators and heat exchangers), the required examinations-i - may be limited to one vessel or distributed.among the vessels. ' A.2.2 CATEGORY C-B, PRESSURE-RETAINING N0ZZLE WELDS IN_ VESSELS l A.2.2.1 Nozzles in Vessels 1/2 In. or less in Nominal Thickness. Item C2.10 All nozzles in vessels 1/2 in, or less in nominal-thickness, at. terminal ends of piping runs shall be surface examined.in accordance with Figure IWC-2520-3 :during each inspection interval. Terminal: ends are the extremi-i ties of piping runs that connect to vessels. Only.those piping runs selected for examination under Examination. Category C-F are included. 4 4 k f i A-23 r

A.2.2.2 Nozzles in Vessels Over 1/2 In. in Womir.el' Thickness. Item C2.20 = All nozzles in vessels over:1/2 in in nominal thickness at terminali ends of piping runs shall be surface'end volunetrically examined in accor-' dance with Figure IWC-2520-4 during each inspection interval. ' Terminal. ends, are the ektremities of piping runs that connect to vessels.. Only. those piping runs selected for examination under Examination Category C-F-. are included. A.2.3 CATEGORY'C-C AND C-E. SUPPORT MEMBERS - A.2.3.1 Integrally Welded Support Attachments in Pressure Vessels, Item C3.10 The surfaces of 100% of each. integrally welded support attachment in[ ' pressure vessels shall be surface examined in. accordance with Figure IWC - 520-5 during each inspection interval'. Examination 'is '11mited to plate-or: shell-type-and. linear-type supports whose base material: design thickness exceeds 1/2 in,. For multiple vessels of'similar design and service, the' required examinations may be conducted on only one vessel. -Where. multiple ej vessels are provided with a number of, similar supporting elements,Jthe examination of the support elements /may be distributed among the vessels. ] A.2.3.2 Component Supports, Item C3.20 Integrally welded pressure' vessel supports of those components required i to be examined under Examination Categories C-Frand C-G shall be visually l examined (VT-3) in accordance with IWF requirements. j l< A.2.3.3 Mechanical and Hydraulic Supports, Item C3.30 Mechanical and hydraulic snubbers and. shock absorbers in pressure vessels shall be visually examined (VT-4) in accordance with-IWF require- ] ments. ' The functional operability of these components shall be confirmed. A.2.3.4 Integrally Welded Support Attachments-in piping. Item C3.40 The surfaces of 100% of each integrally welded support attachment in piping shall be surface examined in accordance with Figure IWC-2520-5. Examination is limited to plate-or shell-type and linear-type supports - whose base material. design t.htckness exceeds 3/4 in. In addition, exami-nations are limited to supports of those components required to be examined under Examinatioti Categories C-F and C-G. I A-24

m i 0-A.2.3.5 Piping toewnent Sooports. Item C3.50 Supports of that piping required to be examined under Examination Categories C-F and C-G shall be visually examined (YT-3) in accordance with IWF requirements. A.2.3.6 Mechanical and' Hydraulic Supnorts in Piping. Item C3.60 Mechanical and hydraulic spring-type hangers, snubbers, and shock absorbers in piping shall be vicuelly examined (VT-4) in accordance with IWF requirements. The functional operability of these components shall be confirmed. Examinations are limited to supports of those components required to be examined under Examination Categories C-F and C-G. A.2.3.7 Integrally Welded Fump Support Attachments. Item C3.70 The surfaces of 100% of each integrally welded support attachemt in purips shall be examined in accordance with Figure IWC-2520-5. Examination is limited to plate-or shell-type and linear type supports whose base material design thickness exceeds 3/4 in. Examinations are limited to supports of those components required to be examined under Examination Categories C-F and C-G. A.2.3.8 Pump Component Supports, Item C3.80 Supports of those pumps required to be examined under Examination Cate-gories C-F and C-G shall be visually examined (VT-3) in accordance with IWF requirements. A.2.3.9 Pump Mechanical and Hydraulic Supports. Item C3.90 Pump mechanical and hydraulic spring-ty absorbers shall be visually examined (YT-4)pe hangers, snubbers, and shock in accordance with IWF-requirements. The functional operability of these components shall be confirmed. Examinations are limited to supports of those components required to be examined under Examination Categories C-F and C-G. A.2.3.10 Integrally Welded Valve Support Attachments. Item C3.100 The surfaces of 100% of each integrally welded valve support attachment shall be examined in accordance with Figure IWC-2520-5 during each inspection interval. Examination is limited to plate-or shell-type and linear-type supports whose base material. design thickness exceeds 3/4 in. Examinations are limited to supports of those components required to be examined under Examination Categories C-F and C-G. i A-25 -l N

A0 2.3.11 Valve Component Supports. Item C3,110 ] Supports of those vahss Nuired to be examined under Examination i Categories C-F and C-G shall be visually examined (VT-3) in accordance with i IWF requirements. ) i A.2,3.12 Valve Mechanical and Hydraulic Supports Item C3.120 i Valve mechanical and hydraulic spring-t absorbers shall be visually examined (VT-4)ype hangers, snubbers, and shock in accordance with IWF require-ments. The functional operability of these components shi.11 be confirmed. i Examinations are limited to supports of those components required to be l examined under Examination Categories C-F and C-E. i -l 1 l A.2.4 CATEGORY C-D. PRESSURE-RETAINING BOLTING EXCEEDING 2' INCHES. IN DIAMETER A.2.4.1 Bolts and Studs in Pressure Vessels. Item C4.10 For bolts and studs in pressure vessels,100% of the bolts and studs at each bolted connection of components required to be inspected shall be l volumetrically examined in accordance with Figure IWC-2520-6 during each 1 inspection interval. Bolting may be examined on one vessel in each system required to be examined that is similar in design, size, function, and i service. In addition, where the component contains a group of bolted con-nections of similar Mign and size (such as flange connections and manway 1 covers), only one bM ud connection among the group need be examined. t Bolting may be examined in place under load or upon disassembly of the l connection. A.2.4.2 Bolts and Studs in Piping, Item C4.20 One hundred percent of the bolts and studs at each bolted piping connection shall be volumetrically examined in accordance with Figure IWC-2520-6. The examination of flange bolting in piping systems required to be examined may-be limited to the flange connections in pipe runs selected for examination under Examination Category C-F. Bolting may be examined in place under load or upon disassembly of the connection. t l A.2.4.3 Bolts and Studs in Pumps, Item C4.30 l-For pumps,1005 of the bolts and studs at each bolted connection of pumps shall be volu, metrically examined in accordance with Figure IWC-I 2520-6. Bolting on only one pump among a group of pumps in each system ' required to be examined that have similar designs, sizes, functions, and. service is required to be examined. In addition, where one pump contains a group of bolted connections of similar design and size (such as flange A-26 I

t connections and manway covers), the exa ;ination may be conducted on one j bolted connection among the group. Bolting any be examined in place under load and upon disasseely of the connection. i A.2.4.4 Bolts and Studs in Valves, Item C4.40 i For valves,100% of the bolts and studs at each bolted connection of valves shall be volumetrically examined in accordance with Figure IWC-2520-6.. Bolting on only one valve among a group of valves in each system required to be examined that have similar designs, sizes, functions, and service is required to be examined. In addition, where the valve contains a group of bolted connections of similar design and size (such as fitnge connections and manway covers), the examination may be conducted on one bolted connection among the group. Bolting may be examined in place under load and upon disasseely of the connection. A. 2. 5 CATEGORY C-F, PRESSURE-RETAINING WELDS IN PIPING l A.2.5.1 Piping Welds 1/2 In. or Less Nominal Wall Thickenss, Item C5.10 l A.2.5.1.1 Circumferential Welds, Item C5.11 i The surfaces of 100% of each circumferential weld 1/2 in, or less i nominal wall thickness shall be examined in accordance with Figure-IWC-2520-7 during each inspection interval. The welds selected for j examination shall include a. all welds at locations where the stresses under the loadings resulting from Normal and Upset plant conditions as calculated by 4 the sum of Equations 9 and 10 in NC-3652 exceed the specified j value; b. all welds at terminal ends (see (e) below) of piping or branch runs; c. all ditsimilar metal welds; ~ dditional welds, at structural discontinuities-(see (f) below) d. a such that the total nunber of welds selected for examination in-l cludes the following percentages of circumferential piping welds; For boiling water reactors: l l 1. none [f the welds exempted by IWC-1220; i 2. none of the welds in residual heat removal and emergency core l cooling systems (see (g) below); 3. 50% of the main steam system welds; 4. 25% of the welds in all other systems. l L .A-27 i l.. ~ ..a A

f~ Tor pressurized water reactorst 1, 1none of the welds exempted by IWC-1220; 1 none of the welds in residual heat removal and emergency core cooling systems; 1 10% of the main steam system welds 8 in nominal pipe size and smaller; 4. 25% of the welds in all other systems, e. terminal ends are the extremities of piping runs that connect to structures, components (such as vessels, pumps, and valves) or pipe anchors, each of which act as rigid restraints or provide et least two degrees of restraint to piping thermal exptssion; f. structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as - elbows, tees, reducers, and flanges conforming to ANSI Standard B16.9), and nine branch connections and fittings; g. examination requirements are under development. A. 2. 5.1. 2 Longitudinal Welds, Item C5.12 Longitudinal welds 1/2 in, or less nominal wall thickness shall be surface examined in accordance with Figure IWC-2520-7 (2.5 t at the intersecting circumferential weld) during each inspection interval.- A.2.5.2 Piping Welds Over 1/2 In.- Nominal Wall Thickness -Item C5.20 A. 2. 5. 2.1 Circumferential Welds, Item C5.21 One hundred percent of each circumferential weld over 1/2 in, nominal wall thickness shall be surface and volumetrically examined in accordance with Figure IWC-2520-7 during each inspection interval. The welds selected for examination shall include 1 all welds at locations where the stresses under the loadings a. resulting from Normal and Upset plant conditions as calculated by the sum of Equations 9 and 10 in NC-3652 exceed the specified value; b. all welds at terminal ends (see (e) below) of piping or branch runs; c. all dissimilar metal welds; d. additio'nal welds, at structural discontinuities (see (f) below) such that the total number of welds selected for examination in-cludes the following percentages of circumferential piping welds; A-28

For boiling cater reactors: 1. none of the welds exempted by IWC-1220; 2. none of the welds.in residual heat removal and emergency core cooling systems (see (g) below) 3. 50% of the main steam system welds; 4. 25% of the welds in all other systems. For pressurised water reactors: 1. none of the welds exempted by IWC-1220; 2. none of the welds in residual heat removal and emergency core cooling systems;

3. -10% of the main steam system welds 8 in, nominal pipe size and smaller; 4.

25% of the welds in all other systems. terminal ends are the extremities of piping runs that connect to e. structures, components (such as vessels, pumps, and valves) or pipe anchors, each of which act as rigid restraints or provide at least two degrees of restraint to piping thermal expansion; f. structural discontinuities include pipe weld joints to' vessel nozzles, valve bodies, pump casings, pipe fittings (such as, elbows, tees, reducers, and flanges conforming to ANSI Standard B16.9), and nine branch connections and fittings; 9 examination requirements are under development. A.2.5.2.2 Longitudinal Welds, Item C5.22 Longitudinal welds over 1/2'in. nominal wall thickness-shall' be surface and volumetrically examined in accordance with Figure IWC-2520-7 (2.5 t at the intersecting circumferential weld) during each inspection interval. A.2.5.3 Pipe Branch Connections, Item C5.30 A. 2. 5. 3.1 Circumferential Welds, Item C5.31 The surfaces of 100% of each circumferential weld in pipe branch connections shall be examined in accordance with Figure IWC-2520-9 during each inspection interval. The welds selected for examination shall include all welds'at locations where the stresses under the loadings a. resulting from Normal and Upset plant conditions as calculated by the sum of Equations 9 and 10 in NC-3652 exceed the specified value; i 1 A-29 1'

o b. all welds at terminal ends-(see (e)~ below) of piping'or branch runs; c. all dissimilar natal welds; d. additional welds, at structural discontinuities (see (f) below) such that the total number of welds selected for examination in-cludes the following percentages of circumferential piping welds; For boiling water reactors: 1. none of the welds exempted by IWC-1220; 2. none of the welds in residual heat removal and emergency core cooling systems (see (g) below); 3. 50% of the main steam system welds; 4 25% of the welds in all other systems. For pressurized water reactors: 1. none of the welds exempted by IWC-1220; 2. none of the welds in residual heat removal and emergency core cooling systems; _ 3. 10% of the main steam system welds 8'in. nominal pipe size and smaller; 4. 25% of the welds in all other systems, e. terminal ends are the extremities of piping runs that connect to structures, components (such as, vessels, pumps, and valves) or pipe anchors, each of which act as rigid restraints or provide at least two degrees of restraint to piping thernal expansion; f. structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as, elbows, tees, reducers, and flanges conforming to ANSI Standard B16.9), and nine branch connections and fittings'; g. examination requirement's' are under development. A.2.5.3.2 Longitudinal Welds, Item C5.32 Longitudinal welds in pipe branch connections shall be surface examined in accordance with Figure IWC-2520-7 (2.5 t at the intersecting circumfer-ential weld) during each inspection interval. 4 A-30

1l-A.2.6 CATEGORY C-G,' PRESSURE-RETAINING WELDS IN PUMPS AND VALVES I A.2.6.1 Pumo Casing Welds, Item C6.10 ~ One hundred percent of all pump casing welds in each piping run examined under Examination Category C-F shall be surface examined in accordance with Figure IWC-2520-8 during each inspection interval. For multiple pumps of similar design, size, function, and service in a system, only one pump among each group of multiple pumps is required to be examined. The examination may be performed from either the inside or outside surface. A.2.6.2 Valve Body Welds. Item C6.20 One hundred percent of all valve body welds in each piping run examined under Examination Category C-F shall be surface examined in accordance with Figure IWC-2520-8 during each inspection interval. For multiple valves of similar design, size, function, and service in a system, only one valve among each group of multiple valves-is required to be examined. The examination may be performed from either the inside or outside surface. A.2.7 CATEGORY C-H, ALL PRESSURE-RETAINING COMPONENTS A.2.7.1 Pressure Yessels, Item C7.10 Pressure vessel pressure-retaining boundaries (other than open-ended portions of systems) shall be visually examined (VT-2) during the system leakage test performed in accordance with IWC-5221 during each inspec-tion. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A. 2. 7.1.1 Pressure Vessels, Item C7.11 Pressure vessel prc1sure-retaining boundaries (other than open-ended portions of systems) shall be visually examined (VT-2) during'the system hydrostatic test. performed in accordance with IWC-5222 during each ins'pec-tion period. There are no exemptions or exclusions from these requirements except as specified in IWA 5214. The system hydrostatic test shall be conducted at or near the end of each inspection interval. or during the same inspection period of each inspection interval of Inspection Program B. A.2.7.2 Piping Item C7.20 Pipin systems) g pressure-retaining boundaries (other than open-ended portions of shall be visually examined (VT-2) during the system leakage test performed in accordance with IWC-5221 during each inspection period. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A-31

A. 2. 7. 2.1 Piping, Item C7.21 Piping pressure-retaining boundaries (other than open-ended portions of systems) shall be visually examined (VT-2) during the system hydrostatic test performed in accordance with IWC-5222 during each inspection period. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. The system hydrostatic test shall be conducted at or near the end of each inspection interval or during the same inspection period of each inspection interval of Inspection Program B. A.2.7.3 Pumps, Item C7.30 Pump pressure-retaining boundaries (other than open-ended portions of systems) shall be visually examined (VT-2) during the system leakage test performed in accordance with IWC-5221 during each inspection period. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A. 2. 7. 3.1 Pumps, item C7.31 Pump pressure-retaining boundaries (other than open-ended portions of systems) shall be visually examined (VT-2) during the system hydrostatic test performed in accordance with IWC-5222 during each inspection period. There are no exemptions or exclusions from these requirements except as specified in IWA-5214 The system hydrostatic test shall be conducted at or near the end of each inspection interval or during the same inspection period of each inspection interval of Inspection Program B. A.2.7.4 Valves, Item C7.40 Valve pressure-retaining boundaries other than open-ended portions of systems) shall be visually examined (VT-2) in accordance with IWC-5221 during each inspection period. There are no exemptions or exclusions from these requirements except as specifled in IWA-5214. A system functional test (IWC-5221) serves as a required system pressure test. A.2.7.4.1 Valves, Item C7.41 Valve pressure-retaining boundaries (other than open-ended portions of systems) shall be visually examined (VT-2) during the system hydrostatic test performed in accordance with IWC-5222 during each inspection period. There are no exemptions or exclusions from these requirements except as specified in IWA-5214 The system hydrostatic test shall be conducted at or near the end of each inspection interval or during the same inspection period of each inspection interval of Inspection Program B. A-32

l A. 3 CLASS 3 REQUIREMENTS l A.3.1 CATEGORY D-A A. 3.1.1 Item D.1.1 Pressure-retaining components within the boundary of systems or. portions of systems required to operate in support of normal plant safety J functions of shutting down and maintaining the reactor in the cold shut-down condition shall be visually examined during system pressure tests. These components shall be visually examined (VT-2) during the system in-1 service test (IWD-5221) during each inspection period. Ln addition, these components shall be visually examined (VT-2) during the system hydrostatic : 1 test (IWD-5223) during the same period of each inspection interval of - Intpection Prograin B. The system boundary includes only those portions of the system required to operate or support the safety function up to and-including the first normally closed valve or valve capable of automatic I closure when the safety function is required. A. 3.1. 2 Item D.1.2 The component supports and restraints within the boundaries of the systems described for Item D.1.1 for components exceeding 4 in nominal pipe size shall be ' visually examined (VT-3) during each inspection period. A. 3.1. 3 Item D.1.3 Mechanical and hydraulic snubbers, spring loaded and constant weight support hangers for components exceeding 4 in, nominal pipe size shall be visually examined (VT-4) during each inspection period. A.3.2 CATEGORY D-B A. 3. 2.1 Item D.2.1 Pressure-retaining components within the boundary of systems or portions of systems required to operate in support of the post-accident j safety functions of emergency core cooling, containment heat removal, and atmospheric cle&nup and long-term residual heat removal from the reactor-shall be visually examined during system pressure tests. These components "3 shall be visually examined (VT-2) during the system functional test (IWD-5222) at least once at or near the end of each inspection period, coinciding with a system functional test. In addition, these components shall be visually examined (VT-2) during a system hydrostatic test (IWD-5223) during the same period of each inspection interval of Inspection Program B. A-33 s

A.3.2.2 Item D.2.2 Component supports and restraints within the boundaries of the systems described for Item D.2.1 for components exceeding 4 in, nominal pipe size shall be visually examined (VT-3) during each inspection period. A.3.2.3 Item D.2.3 Mechanical and hydraulic snubbers, spring loaded and constant weight support hangers for components exceeding 4 in, nominal pipe size shall be visually examined (VT-4) during each inspection interval. A.3.3 CATEGORY D-C A.3.3.1 Item D.3.1 Pressure-retaining piping, pumps, and valves within the boundary of systems or portions of systems required to operate in support of residual heat removal from the spent fuel storage pool shall be visually examined during system pressure tests. These components shall be visually examined (VT-2) at least once during the system inservice test IWD-5221 during each inspection period. In addition, these comoonents shall be visually examined (VT-2) during a system hydrostatic test (IWD-5223) during the same period of each inspection interval of Inspection Program B. A 3.3.2 Item D.3.2 Component supports and restraints within the boundary of the systems described for Item D.3.1 for components exceeding 4 in nominal pipe size shall be visually examined (YT-3) during each inspection period. A.3.3.3 Item D.3.3 Mechanical and hydraulic snubbers and spring loaded and constant weight support hangers for components exceeding 4 in. nominal pipe size shall be visually examined (VT-4) during each inspection period. 4 A-34 -}}