ML20202J455

First Interval Inservice Insp Program,Grand Gulf Nuclear Station, Technical Evaluation Rept

ML20202J455
Person / Time
Site:	Grand Gulf
Issue date:	03/28/1986
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20202J461	List: ML20202J455 ML20203G425 ML20203G436
References
CON-NRC-03-82-096, CON-NRC-3-82-96 SAIC-86-1626, TAC-55522, TAC-55524, TAC-57222, NUDOCS 8604160051
Download: ML20202J455 (98)

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ATTACHMENT 2 SAIC-86/1626 TECHNICAL EVALUATION REPORT FIRST INTERVAL INSERVICE INSPECTION PROGRAM GRAND GULF NUCLEAR STATION Submitted to

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U. S. Nuclear Regulatory Commission Contract No. NRC-03-82-096 Submitted by _

Science Applications International Corporation Idaho Falls, Idaho 83402 l

March 28, 1986 I

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CONTENTS

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . I 3

2. EVALUATION OF INSERVICE INSPECTION PLAN ...........

2.1 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 DOCUMENTS EVALUATED .........-............ .3 2.3

SUMMARY

OF REQUIREMENTS ................... 3 4

2.3.1 Code Requirements . . . . . . . . . . . . . . . . . .

2.3.1.1 Class 1 Requirements ................ .

4 2.3.1.2 Class 2 Requirements ................ 5 2.3.1.3 Class 3 Requirements ................ 5 2.3.1.4 Component Supports ................. 5 2.3.2 Preservice Inspection Commitments . . . . . . . . . . . 5 6

2.4 COMPLIANCE WITH REQUIREMENTS . . . . . . . . . . . . . . . . .

1 2.4.1 Appl icable Code Edition . . . . . . . . . . . . . . . . 6

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2.4.2 Code Requirements . . . . . . . . . . . . . . . . . . . 6 2.4.3 Preservice In;pection Commitments . . . . . . . . . . . 7 7

2.5 CONCLUSION

S AND RECOMMENDATIONS ...............

3. EVALUATION OF RELIEF REQUESTS ................ 9

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3.1 CL ASS 1 COMPON EN TS . . . . . . . . . . . . . . . . . . . . . . 9 3.1.1 Reactor Vessel ................... 9 ,

3.1.1.1 Relief Request No. I-00004, RPV Lower Head-to-Shell Weld A-A, Category B-A, Items Bl.ll and B1.21 . . . . . . . . . . . . . . . . . . . . . . . . 9 i 3.1.1.2 Relief Request No. I-00005, RPV Seam Weld A-8 Lower-Half, Category B-A, Item B1.11 ........ 14 3.1.1.3 Relief Request No. I-00006, RPV Seam Weld A-C, Category B-A, Item Bl.ll .............. 19 3.1.1.4 Relief Request No. I-00008, Control Rod Drive and In-Core Housing Welds and Flange Bolting, Category B-0, Item B14.10, and Category B-G-2, Item B7.10 .. ,23 3.1.2 Pressurizer (not applicable to BWRs) 3.1.3 Steam Generators and Heat Exchangers (no relief requests) i

3.1.4 Piping Pressure Boundary .............. 26 3.1.4.1 Relief Request No. I-00007, Piping Welds Within Guard Pipes, Category B-J, Items 89.11 and 89.21 .. 26 3.1.4.2 Relief Request No. I-00010 and I-00012, Pressure Retaining Piping Welds on RHR, RCIC, MS, RECIRC, and RWCU Systems, Category B-J, Items B9.11 and 89.12 . . 31 3.1.5 Pump Pressure Boundary (no relief requests) 3.1.6 Valve Pressure Boundary (no relief requests) i-3.2 CLASS 2 COMPONENTS . . . . . . . . . . . . . . . . . . . . . 37 3.2.1 Pressure Vessels and Heat Exchangers (no relief requests) 3.2.2 Piping Pressure Boundary .............. 37 i

3.2.2.1 Relief Request No. I-00002, Pressure Retaining -

Weld in Scram Discharge Volume Piping, Category C-F, Item C5.21 ................... 37 -

3.2.2.2 Relief Request No. I-00003, Circumferential Dissimilar Metal Weld, Category C-F, Item C5.21. . . 40 3.2.2.3 Relief Request No. I-00010, Pressure Retaining Piping Welds on RCIC System, Category C-F, Item C5.21 ..................... 43 3.2.2.4 Relief Request No. I-000ll, Thermal Tee Sleeve Welds, Category C-F, Item C 5.21 .......... 47 3.2.3 Pump Pressure Boundary ............... 49 3.2.3.1 Relief Request No. I-00009, Pump Casing Welds, Category C-G, Item C6.10 .............. 49 3.3 CLASS 3 COMPONENTS (no relief requests) i 3.4 COMPON EN T SUP PORTS . . . . . . . . . . . . . . . . . . . . . . 53 3.4.1 Plant and Shell-Type Supports (no relief requests) 3.4.2 Linear Type Supports (no relief requests) i

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3.4.3 Component Standard Supports . . . . . . . . . . . . . 54 i 3.4.3.1 Relief Request No. I-00001, Pipe Supports Within l

Guard Pipes, Category F-C, Item F-2 . . . . . . . . . 54 1

3.5 PRESSURE TESTS (no relief requests) 3.6 GENERAL (no relief requests)

4. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . 57 APPENDIX A: REQUIREMENTS OF SECTION XI, 1977 EDITION, l WITH ADDENDA THROUGH SUMMER 1979 .......... A-1 Q

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TECHNICAL EVALUATION REPORT FIRST INTERVAL INSERVICE INSPECTION PROGRAM Grand Gulf Nuclear Station

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 f acilities . Incorporated by reference in this regulation isSection XI of the Boiler and Pressure Vessel Code published by the Anerican Society of Mechanical Engineers (ASME), which provides the basis for implementing inservice inspection.*

Two types of inspections are required: (1) a preservice inspection conducted before comercial operation to establish a baseline and (2) peri-odic inservice inspections conducted'during 10-year intervals that normally start from the date of comercial operation. Separate plans for completing preservice inspection and each 10-year inservice inspection must be formu-lated and submitted to the Nuclear Regulatory Commission (NRC). The plan for each 10-year interval must be submitted at least 6 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 Grand Gulf Nuclear Station - Unit 1 (GGNS-1) has been written to the 1977 edition with addenda through Sumer 1979.

Section 2 of this report evaluates the first interval ISI Plan developed by the licensee, Mississippi Power and Light Company (MP&L) for GGNS-1 for (a) compliance with this edition of Section XI, (b) compliance with ISI-related commitments identified during the NRC's review before granting an Operating License, (c) acceptability of examination sample, and (d) exclusion criteria.

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

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t The Grand Gulf Nuclear Station, Unit No.1, received a construction permit on Septemoer 4,1974. The preservice inspection is based on confor-mance with the ASME Code,Section XI,1977 Edition, including addenda through September 1978. The ASME first published rules for inservice inspection in the 1970 Edition of Section XI. No preservice inspection requirements existed before that date. Since the Grand Gulf Unit No.1, plant system designs and ordering of long lead time components were well underway by the time the Section XI rules became effective, full compliance with the access and inspectability requirements was not always practical because of limitations of design, geometry, and materials of construction of components and systems that were designed before Section XI. The regulation therefore permits exceptions to impractical examination or testing requirements of the current Code to be requested. Relief from these requirements may be granted, provided the health and safety of the public are not endangered, giving due consideration to the burden placed on the licensee if the requirements were imposed. Section 3 of this report 3

evaluates requests for relief dealing with inservice examinations of

components and with system pressure tests.

The regulation also provides that ISI Programs may meet the requirements of subsequent Section XI editions a.1d addenda, incorporated by reference in the Regulation, subject to approval by the NRC. Portions of such editicns or addenda may be used, provided all related requirements of the respective editions or addenda are met. These instances are addressed on a case-by-case basis in Section 3 of this report. Likewise,Section XI provides that certain components and systems may be exempted from its requirements. In some instances, however, these exemptions are not accep-table to the NRC or are acceptable only with restrictions. As appropriate, exemptions are also discussed in Section 3 of this report.

Preliminary material on the Preservice Inspection (PSI) Program for GGNS-1, a General Electric boiling water reactor (BWR), was submitted in

References 1 through 3. This material was evaluated and additional required material identified in q NUREG-0831, dated SeptemberThe 1981.

license 4) submitted Safety Evaluation additional Report (SER material on the PSI program, including relief requests, exemption criteria, and isometric drawings in References 5 and 6. The PSI program and religf requests were approved in Supplement No. 2 to the SER, dated June 1982.\7)

An additional PSI relief request was submitted in Reference 8 ggd subsequently approved by the staff in Supplement No. 4 to the SER.(V1 Reference 9 also included the license condition that the first interval ISI program be submitted by June 30, 1983. References 10,11, and 12 deal further with establishing the schedule for submitting the ISI program, re-sulting in a required submittal date of September 1,1984. References 13, The first 14, and 15 contain additional relief requests interval ISI program was submitted July 25, 1984. IW for(P}" and ISI.

References 17, 18,

- 19, and 20 contain additional material related to the ISI program.

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2. EVALUATION OF INSERVlCE INSPECTION PLAN 2.1 Introduction The approach being taken in this evaluation is to-[e' view the applicable program documents to determine the adequacy o7 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

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recommendations are given. ,

2.2 Documents Evaluated , ,

A chronology of documents received for the evaluatien is given in Section 1 of this report. Those documents that impact this evaluation are (1) the ISI Program, (2) the licensee's response to the staff's request for additional information, (3) additional relief requests submitted by the licensee, and (4) to a lesser extent previous submittals on tne PSI program.

2.3 Summary of Reouirements The requirements on which this review is focused include the following:

(1) Comoliance with Acolicable Code Edf.tions. The Inservice Inspection Program snall De caseo on tne Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). The licensee for GGNS-1 has written the first interval prograns to the 1977 Edition with addenda through Summer 1979. These Code requirements are '

summarized in 2.3.1 below, and detailed Code requirements are given in Appendix A. The 1974 Edition, Sumer 1975 Addenda should be used for selecting Class 2 welds in systems providing the 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).

(2) Acceptability of the Examination Sample. Inservice volumetric, surface, anc visual examinations snall oe performed on ASME Code Class 1 and 2 components and their supports using sampling schedules described in Section XI of the ASME Code and 10 CFR 50.55a(b). Sample size designations are identified as part of the Code requirements in Appendix A.

! m (3) Exclusion Criteria. The criteria used to exclude components f rom examination snall be consistent with IWB-1220, IWC-1220, and j

10 CFR 50.55a(b). -

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(4) PSI Commi tments. The Inservice Inspection Program should address all license conditions, qualified acceptance conditions, or other ISI related commitments described in the SER and its supplements for the preservice examination.

2.3.1 Code Requirements The following requirements are summarized from the 1977 edition of Section XI with addenda through summer 1979. Many requirements call for 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. For detailed requirements, see Appendix A of this report or the Code itself.

2.3.1.1 Class 1 Requirements 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 inches in Diameter (7) Pressure Retaining Bolting, 2 inches and Smaller in Diameter (8) Integral Attachments fer Vessels (9) Pressure Retaining Welds in Piping (10) Pump Casings and Valve Bodies, including Pressure Retaining Welds (11) Interior of Reactor Vessel, including Integrally Welded Core Support Structures, Interior Attachments, and Removable Core Support Structures (12) Pressure Retaining Welds in Control Rod Housings 1

(13) All Pressure Retaining Components - Pressure Tests l (14) Steam Generator Tubing.

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2.3.1.2 Class 2 Requirements The following Class 2 components are to be examined in the first interval in accordance with Table IWC-2500-1:

(1) Pressure Retaining Welds in Pressure Vessels

-(2) Pressure Retaining Nozzle Welds in Vessels

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(3) Integral Attachments for Vessels, Piping, Pumps, and Valves (4) Pres'sure Retaining Bolting Greater than 2-inch Diameter (5) Pressure Retaining Welds in Piping (6) Pressure Retaining Welds in Pump and Valves (7) All Pressure Retaining Components - Pressure Tests.

2.3.1.3 Class 3 Requirements The following Class 3 reactor-connected and associate systems are to be examined in the first interval in accordance with IWD-2500-1:

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(1) Systems in Support of Reactor Shutdown Function (2) Systems in Support of Emergency Core Cooling, Containment Heat Removal, Atmosphere Cleanup, and Reactor Residual Heat Removal (3) Systems in Support of Residual Heat Rerroval from Spent Fuel ,-

Storage Pool. a a

2.3.1.4 Component Supports The following examination and inspection of component supports are to ,

be examined in the first interval in accordance with IWF-2500-1:

(1) Plate and Shell-Type Supports i

(2) Linear-Type Supports (3) Component Standard Supports.

lx 2.3.2 Preservice Inspection Commitments The only license condition related to ISI for GGNS-1 was 2.C.(18),

which required that the ISI program for the first interval be submitted to l the NRC by September 1, 1984. (See References 9,10,11, and 12). The program for the first interval was actually submitted on July 25,1984.( ,

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l 2.4 Compliance with Requirements 2.4.1 Applicable Code Edition _

The initial inservice inspection interval examination program must comply (10 CFR 50.55a(g)(4)(i)) with the requirements of the latest edition and addenda of Section XI incorporated into 10 CFR 50.55a on the date 12 months before the date of issuance of the operating license. Based on a June 16,1982, operating license for GGNS-1, the Code applicable to the first interval program is the 1977 Edition with addenda through Sumer 1979.

2.4.2 Code Requirements -

The first interval the GGNS-1 ISI program.(jST program O The ISI programofsubmitted record iswas contained reviewedin Revision 0 of (exclusive of pump and valve testing), and the following observations were noted:

(a) The GGNS-1 ISI program has been prepared in accordance with the applicable version of the Code.

(b) As specified in 10 CFR 50.55a(b)(2)(iv), Class 2 welds in the RHR and ECCS systems are to be selected for examination in accordance -

with the 1974 Edition of ASME,Section XI, with Addenda through Sumer 1975. GGNS-1 has used an alternate criterion based on application of IWC-1220(a) and (d) for exemptions and selection of 10% of the nonexempted welds at terminal ends of pipe at vessel nozzles, welds at locations of high stress, dissimilar metal welds, welds that cannot be pressure tested and welds at structural discontinuities. For the GGNS-1 plant, the number of welds examined on the RHR and ECCS over the life of the plant is increased by nearly a factor of two using the GGNS-1 selection criteria. Although the weld distribution differs slightly, the GGNS-1 approach includes consideration of high stress locations and dissimilar metal welds in accordance with the more recent codes. The weld selection criteria used for RHR and ECCS are considered to be at least equivalent and p.robably superior to that required by the regulations. ,

(c) Sample size and weld selection have been implemented in accordance with the Code and appear to be correct.

(d) In accordance with 10 CFR 50.55a(g)(4)(iv), portions of the 1980 i Edition of ASME,Section XI, with addenda through Winter 1980 have been adopted as discussed in Volume 1, Section 5 of the ISI Plan.

(e) Exclusion criteria have been applied in accordance with the Code and appear to be correct as discussed in Volume 1, Section 6 of the ISI Plan.

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(f) GGNS-1 has developed an unique plan for inservice inspection of pipe supports in lieu of using the methodology provided in IWF.

The GGNS-1 plan for ISI of pipe supports is occumented in Volume 1, Section 7, of the ISI Plan. The proposed approach is based on statistical sampling to detect failure of more than 10% of the supports with a 95% confidence level.

For Class 1 supports,101 out of a total of 194 supports are scheduled for examination in the GGNS-1 plan. All Class 1 supports would be examined if the failure criteria were exceeded in the original sample. Application '

of IWF methodology would require that 154, or about 50% more, supports be examined.

Class 2 and Class 3 supports are combined into a total of 1495 If defective supports are located i supports, of which 280 will be examined.

in the initial sample, an additional 265 supports will be examined.

' Application of IWF would result in examination of 660 supports, or about twice as many as included in the GGNS-1 plan. Snubber inspection is to be included in the ISI plan at a later time.

2.4.3 Preservice Inspection Commitments 4

k cense condition 2.C.(18) was satisfied with the submittal of the ISI I plan.t 41

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2.5 Conclusions and Recommendations Based on the preceding evaluation, it is. concluded that the first .

interval ISI program reviewed generally meets the Code to which it was written and NRC regulations in regard to examination sample and exclusion '

criteria. There are, however, open items to be addressed by the licensee or the staff as applicable. These items relate to the acceptability of the GGNS-1 component supports examination program and reporting requirements. .

Statistical sampling methods should be acceptable for selecting i supports for examination and are likely to be implemented in future i versions of the Code. Generally the statistical sampling approach results '

in fewer examinations than required by the Code. As illustrated above 101 l

supports out of 194 Class 1 supports would be examined on a statistical sampling basis whereas IWF would require examination of 154 supports out of 194.

The combining of, Class 2 and Class 3 supports into a common statistical base for selection is not consistent with the conceptual or historical development of the Code. This approach results in examination of fewer Class 2 supports then is consistent with the safety importance of

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Class 2 systems. The data provided by the licensee is not sufficient to determine the exact impact of combining Class 2 and Class 3 supports on the '

number of Class 2 supports examined. The combined population of Class 2 i

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, , a and Class 3 supports is stated to be 1495, which includes 575 supports between 2-1/2" and 4" pipe size that would be excluded by the Code. Of the 4 1495, 280 or slightly less than 20% will be examined under the G11S-1 sampling plan. Assuming that a random selection process is emp; ed, 20%

of the Class 2 supports would be examined. Application of IWF nc mally results in examination of about 40-50% of the Class 2 supports. Corres-ponding, IWF normally results in examination of about 5% - 10% of Class 3 supports. The combining of Class 2 and Class 3 supports results in reduced examination of Class 2 supports and increased examination of Class 3 supports. We recommend that Class 2 and Class 3 supports not be combined and that the extent of examination of Class 2 supports be maintained consistant with the higher safety importance of Class 2 systems.

I It should also be noted that snubber examinations are to be included in the plan and the licensee has committed to provide these at a later time. The staff may wish to request a firm schedule for providing the snubber examination plan.

Several relief requests include the commitment by'the licensee to evaluate the growth of existing flaws or other unusual examination results

.' and to consider enhanced examinations following the detection of flaw '

growth or other unusual results. No commitment to infcrm the staff of unusual results or obtain staff approval of resolution is included in the plan. The staff may wish to require reporting requirements of the licensee to assure timely reporting and resolution of any abnormalities. _

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3. EVALUATION OF RELIEF REQUESTS The following sections evaluate the 12 pending relief requests. The material included in the paragraphs titled Code Relief Request, Proposed Alternative Examination, and Licensee's Basis for Requesting Relief is quoteo directly from tne reller request except for minor eoitorial changes such as removing references to figures and tables not included in this report.

3.1 CLASS 1 COMPONENTS Subsections IWA and IWB of the Code govern the examination of Class 1 piping and components. Specific requirements are given in Table IWB-2500-1.

3.1.1 Reactor Vessel 3.1.1.1 Relief Reauest No. I-00004, RPV Lower Head-to-Shell Weld A-A, Lategory 6-A, Items 81.11 ano 81.21 Code Reauirement The upper portion of this weld is a circumferential shell weld (Bl.ll), and the lower portion of this weld is a circumferential head weld (B1.21). ,

Circumferential Shell Welds, Item Bl .11. All pressure-retaining circumferential snell welos in tne reactor vessel shall be volumetric-ally examined in accordance with Figure IWB-2500-1 over essentially 100% of their lengths during the first inspection interval. Examina-tions may be performed at or near the end of the interval.

Circumferential Head Welds, Item 81.21. All pressure-retaining circumferential nead welos in tne reactor vessel head shall be volu-metrically 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.

Code Relief Request Relief is requested to delete all ultrasonic inservice inspections of the lower one-half (Category B-A, Item Bl.21 requirement) of the entire circumference of the A-A weld, except as noted under alternate examinations.

Proposed Alternative Examination Instead of examining the entire lower one-half of the A-A weld, it is proposed that manual ultrasonic examinations be performed only of the section of the weld in which .three recordable indications were found (approximately a 12-inch by 12-inch surface area) once per inservice inspection interval. The sizes of the indications will be 9

l monitored and, if any of the indications appears to be increasing in size, the indications will be evaluated and appropriate actions taken, which may include ultrasonic examinations of other sections of the weld . The anticipated exposure for performing the alternative examination would be less than 500 millirem.

Licensee's Basis for Requesting Relief The A-A weld joins the lowest ring of circumferential shell plates on the reactor pressure vessel (RPV) to the RPV bottom head and is located 80.66 inches above vessel zero. The bottom of the core is located at 216.31 inches above vessel zero. The weld is approximately 135.65 inches below the bottom of the core. In addition, the weld is approximately 91.2 inches below the centerlines of the recirculation pump suction nozzles (N1 nozzles) and approximately 98.4 inches below the centerlines of the recirculation pump discharge nozzle (jet pump suction nozzles-N2).

The upper portion of the A-A weld is a typical circumferential shell weld. Automated ultrasonic examination procedures and equipment have been developed which will permit the required inservice volumet-ric inspections to be performed remotely. The lower portion of the A-A weld is a circumferential head weld. Due to the curved geometry of this portion of the weld, automated means for ultrasonic examina-tion of this portion of the weld have not been developed; thus, this portion of the weld must be ultrasonically examined by manual procedure.

The containment design of Grand Gulf Nuclear Station Unit 1 is designated Mark III. A feature of this design is that an annulus space of approximately 30 inches width exists between the reactor vessel outer circumference and the biological shield wall inner -

circumference. The examiners must enter this annulus space to perform manual ultrasonic examination of the A-A weld. Contact radiation levels on, and area radiation levels near, the recirculation inlet and outlet nozzles recorded at other BWR plants during the first three to six years of reactor operation have been in the range of 200 mR/hr to 2000 mR/hr. We anticipate the area radiation levels at the A-A weld to be approxe.iately the same as those near the nozzles due to their proximity to each other, the constricted space in annulus, the proxi-mity to the core, which is treated as an area source, and possible reflection from the metallic insulation on the inner surface of the biological shield wall. For purposes of estimating exposure, we have assumed an area radiation level at the A-A weld of 800 mR/hr at the end of the first 40-month inservice inspection period. The level is expected to increase as the plant ages.

Due to the large amount of weld area required to be examined and the nature of the radiation sources in the area, shielding is not practical. Such shielding would need to shield the entire body, would be heavy and difficult to move and would require significant exposure to erect and move.

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• a Based on the results of the examinations performed during the preservice inspections, it is estimated that 16 man-hours will be required to perform the required manual ultrasonic examinations. This

' time does not incicde tne time required for personnel to enter and exit the annulus space; however, it does include an allowance for the 4 extra time required by personnel due to wearing protective clothing and for mapping three recordable, but not reportable indications, which were found during preservice inspections.

We estimate that the manual ultrasonic examination of the A-A weld will require approximately 12,800 millirem of personnel exposure. Entry and exit of the annulus region plus support personnel exposure during the examinations are estimated to require an additional 1,700 millirem for a total estimated exposure of 14,500 millirem.

Relief from the ultrasonic inservice inspections of the A-A weld is requested for the following reasons:

1. The upper one-half (Category B-A, item B1.11) of the A-A weld was examined by remote ultrasonics as a preservice l

inspection in accordance with ASME Section XI, and no recordable indications were found.

2. The lower one-half (Category B-A, item Bl.21) of the A-A _

weld was examined by manual ultrasonics as a preservice inspection in accordance with ASME Section XI, and a total of three recordable, but no reportable, indications were found. The examination report shows that the indications are outside the heat affected zone of the weld. .

3. The entire reactor pressure vessel was subjected to a hydrostatic pressure test in accordance with ASME Section

- III.

The upper one-half of the A-A weld will be examined by 4.

remote ultrasonics during the first inservice inspection interval in accordance with the requirements of ASME Section XI.

) The entire reactor pressure vessel will be subjected to a 5.

system leakage test at each refueling outage and to a system hydrostatic test each inservice inspection interval in accordance with the requirements of ASME Section XI.

6. The lower one-half.of the A-A weld is a circumferential head weld. Due to the curved geometry of this portion of the weld, automated means for ultrasonic examination of this portion of the weld have not been developed; thus, this portion of the weld must be ultrasonically examined by manual procedure. Based on the results of the examinations L

performed during the preservice inspections, it is estimated that the manual ultrasonic examination of the A-A weld will 11

a require approximately 12,800 millirem of personnel exposure.

Entry and exit of the annulus region plus support personnel exposure during the examinations are estimated to require an additional 1,700 millirem, for a total estimated exposure of 14,500 millirem.

Evaluation The lower head-to-shell weld A-A on the GGNS-1 RPV is the S

transition from the cylindrical shell to the spherical lower head.

The weld is accessible through an annular gap approximately 30 inches wide between the RPV and the biological shield wall. The upper portion of the weld will be examined over 100% of the Code required volume using remote, automated ultrasonic inspection equipment.

However, the automated equipment, as currently configured, cannot be

) used to examine the lower portion of the weld due to the spherical curvature of the lower head. Manual ultrasonic examination of the lower portion of the weld is the only alternate currently available.

The lower portion of the A-A weld was examined over 100% of the length during the preservice examination using manual ultrasonic

, methods. Based cn this examination, the licensee estimates that approximately 16 man-hours are required exclusive of exit and entry time to conduct a complete manual ultrasonic scan of the lower portion _

of the A-A weld. The radiation field in the region of the A-A weld has been estimated to be 800 mR/hr based on radiation fields in similar BWRs following 40-months operation. Combining the estimates 9

of inspection time and radiation level, the licensee estimates that a total exposure of 14,500 millirem would be required to conduct the inservice examination of the lower portion of the A-A weld in the first period. Higher exposures would occur if the examinations are deferred. .

). As an alternate to 100% examination of the A-A weld, the licensee has proposed manual ultrasonic examination of a 12" by 12" section of _

the A-A weld in which three recordable, but not reportable, indica-tions were found during the preservice examination. The alternate examination is estimated to result in less than 500 millirem exposure.

If a change in the indications were observed, more extensive examina-g tion woulo be evaluated. Since a 12" by 12" patch of the A-A weld can be examined with manual volumetric methods without excessive radiation exposure and since monitoring an additional 12" by 12" patch where no flaws had occurred prior to PSI would give a measure of the condition of the remainder of the weld, we recommend that an additional 12" by 12" patch of the A-A weld be volumetrically examined. The second or

$ reference patch should be at least 90 0 from the patch containing the fl aws.

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Based on the high radiation exposure that would result from conducting complete volumetric examinations of the lower portion of the A-A weld, the examination is considered impractical. The 0 alternative examinations proposed, which include monitoring the reportable indications found in the lower portion of the A-A weld 12

)

% 6 during preservice, eill assure that the necessary level of structural reliability is achieved.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the lower head-to-shell weld discussed above, adherence to the Code require-ments is impractical. It is further concluded that the proposed examinations will provide the necessary assurance of structural

) reliability during this interval. Therefore, relief is recommended as requested provided:

(a) the upper portion of the A-A weld is volumetrically examined over 100% of the weld length, 4 (b) the manual volumetric examinations, over a 12" by 12" area, of the reportable defects and a reference patch in the lower portion of the A-A weld are performed and evaluated, and (c) the Code-required system pressure tests are performed.

4 The requested reiief has been recommended based on the impracticality of conducting manual ultrasonic examinations in radiation fields estimated to exist in the examination area. If actual radiation fields are lower, for example if the core were removed, more extensive examinations should be conducted, i It is further recommended that improvements in automated ultrasonic inspection technology be monitored and that the entire length of the lower portion of the A-A weld be examined if techniques become available during the 10-year interval.

References References 16, 20, 21, and 22. _

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3.1.1.2 Relief Request No. I-00005, RPV Seam Weld A-B Lower Half, Lategory o-a, item oi.ii.

Code Requirement All pressure-retaining circumferential'shell welds in the reactor vessel shall be volumetrically examined in accordance with Figure IWB-2500-1 over essentially 100% of their lengths during the first inspection interval. Examinations may be performed at or near the end of the interval.

Code Relief Reauest Relief is requested to delete all ultrasonic inservice inspection on the lower one-half (Category B-A, Item B1.11 requirements) of the entire circumference of the A-B weld, except as noted under alternate examinations.

Proposed Alternative Examination Instead of examining the entire icwer one-half of the A-B weld, -

it is proposed that manual ultrasonic examinations be perforned only on the section of the weld in which the four recordable indications

. were found (approximately a 13-inch by 22-inch surface area), once per inservice inspection interval. The sizes of the indications will be monitored and, if any of the indications appear to be increasing in size, the indications will be evaluated and appropriate action taken, which may include ultrasonic examinations of other sections of the weld. The anticipated exposure for performing these alternative examinations would be less than 4600 millirem.

~

i Licensee's Basis for Requesting Relief The circumferential weld A-B joins together ring 1 and ring 2 of the reactor pressure vessel and is located 210.66 inches above vessel zero. The bottom of. the core is located at 215.31 inches above vessel zero. The weld is approximately 4.65 inches below the bottom of the core. The centerlines of the recirculation pump discharge nozzles (jet pump suction nozzles-N2) and the recirculation pump suction nozzles (N1 nozzles) are located approximately 30.96 inches and 37.92 l inches, respectively, below the weld.

The upper portion of seam A-B is a typical circumferential shell I weld. An automated ultrasonic examination procedure and equipment  ;

have been developed which will permit the required inservice volumetric inspections to be performed remotely.

e 14

However, due to the nozzle interference with the automated ultrasonic equipment and the irregular geometry on the lower portion of weld A-B, remote examination is not possible. Therefore, this portion of the weld must be examined by the manual procedure. The containment design of Grand Gulf Nuclear Station Unit 1 is designated Mark III. A feature of this design is an annulus space of approxi-mately 30 inches width which exists between the reactor vessel outer circumference and the biological shield wall inner circumference. The examiners must enter this annulus space to perform the manual ultra-sonic examination of weld A-B. since we have no actual radiation data from GGNS, we are basing our estimates on data from other plants. The information we received from the other BWR plants indicated that seam A-B is located in a radiation field of approximately 8-10 R/hr. This is a conservative estimate for the end of the first fuel cycle (first refueling outage). The actual exposures will probably be less; how-ever, the area and contact readings can be expected to rise as the plant ages. .

Based on the results of the preservice inspections,. it is estimated that 16 man-hours will be required to perform the manual ultrasonic examinations of weld A-B. This time also does not include the time required for personnel to enter and exit the annulus space; i

however, it does include an allowance for the extra time required by personnel due to wearing protective clothing and for mapping four recordable, but not reportable, indications which w're found during the preservice examination. _

As stated earlier, personnel performing manual ultrasonic examinations on this particular weld will be subject to very high area radiation fields, in the neighborhood of 8-10 R/hr. Even assuming GGNS daily and weekly administrative whole-body limits are waived and the inspectors are allowed to accumulate their entire GGNS quarterly administrative limit of 2500 mrem in one dose, personnel stay times in this area would be less than 20 minutes.

We estimate that the manual ultrasonic examination of the A-B weld will require approximately 128,000 millirem of exposure per _

10-year interval or 42,666 millirem each inspection period. The calculations do not include exposure to support personnel.

Due to the large amount of weld area required to be examined and the nature of the radiation sources in the area, temporary shielding  !

l is not practical. Such shielding would need to protect the entire '

body, would be heavy and difficult to move, and would require signifi-cant exposure to erect in and remove from the annulus region. If the shielding were worn on the body (lead vests, for example), the extra weight of the shielding would slow personnel entry and exit, thus negating any advantages.

Relief from ultrasonic inservice inspection of weld A-B is requested for the following reasons:

l 15 1

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1. The entire reactor pressure vessel was subjected to a hydrostatic pressure test in accordance with ASME Section III.

2. The entire reactor pressure vessel will be subjected to a system leakage test at each refueling cutage and to a system hydrostatic test each inservice inspection interval in accordance with the requirements of ASME Section XI.

3. The upper one-half (Category B-A, Item 81.11) of the A-B weld was examined by remote ultrasonics as a preservice inspection in accordance with ASME Section XI and 13 recordable, but not reportable, indications were found.

4. The lower one-half (Category B-A, Item B1.11) of the A-B weld was examined by manual ultrasonics as a preservice inspection in accordance with ASME Section XI, and a total of four recordable, but not reportable, indications were found. These indications were laminar reflectors in the base metal and are located outside the heat affected zone of the weld.

1

5. Based on the information we received from other BWR plants, l

we have estimated that the manual ultrasonic examination on _

weld A-B could require as many as 48 examiners.

6. Weld A-B will be examined by remote ultrasonics from the #2 ring side of the weld during the first inservice inspection interval (first 10 years). It is not practical to perform a remote ultrascnic examination from the #1 ring side of the weld due to nozzle interference. However, a minimum of 70%

coverage of the weld will be attained with the 00, 450, 600 transducers scanning from the #2 ring. Full coverage is not possible with the remote equipment because of the seam configuration. Based on the results of preservice exami- .

nations and radiation field data from other plants, it is estimated that manual ultrasonic examination of the A-B weld could result in approximately 128,000 millirem of exposure per 10-year interval o. 42,666 millirem each inspection period. The calculations do not include exposure to support personnel .

Evaluation Shell weld A-B is a circumferential weld which joins. ring 1 and ring 2 in the cylindrical portion of the vessel at an elevation slightly less than 5 inches below the bottom of the core. The weld is accessible through an annular gap approximately 30 inches wide between the RPV and the biological shield wall. The upper portion of the weld will be examined over 100% of the Code required volume using remote, automated ultrasonic inspection equipment. However, the automated equipment, as currently configured, cannot be used to 16

examine the lower portion of the weld due to interference with the recirculation pump suction and discharge nozzles. Manual ultrasonic examination of the lower portion of the weld is the only alternate currently available.

The lower portion of the A-B weld was examined over 100% of the length during the preservic examination using manual ultrasonic me thods . Based on this examination, the licensee estimates that approximately 16 man-hours are required exclusive of exit and entry time to conduct a complete manual ultrasonic scan of the lower portion of the A-B weld. The radiation field in the region of the A-B weld has been estimated to be 8-10 R/hr based on radiation fields in similar BWRs following 40-months operation. Combining the estimates of inspbetion time and radiation level, the licensee estimates that a total exposure of 128,000 millirem would be required to conduct the ,

inservice examination of the lower portion of the A-B weld.

As an alternate to 100% examination of the A-B weld, the licensee has proposed manual ultrasonic examination of a 13" by 22" section of the A-B weld in which four reportable, but not recordable, indicators were found during the preservice examination. The alternate examina-tion is estimated to result in less than 4600 millirem exposure. If a change in the indications were observed, more extensive examination would be evaluated. _

Based on the high radiation exposure that would result from conducting volumetric examinations of the lower portion of the A-B weld, the examination is considered impractical. Considering that 100% of the code required volume from both sides of the weld was volumetrically axamined during the preservice examination, that 100%

of the Code required volume will be examined from the upper side of the weld during the first interval, that the reportable indications found in the lower half of the weld during preservice will be examined volumetrically during the first interval, that a VT-2 visual examina-tion of Class I components will be conducted during system hydrostatic ~

and leakage tests, the necessary level of structural reliability can be achieved.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the circumferential shell weld discussed above, adherence to the Code requirements is impractical. It is further concluded that the proposed examinations will provide the necessary assurance of structural reliability during this interval. Therefore, relief is recommended as requested provided:

17

(a) the upper portion of the A-B weld is volumetrically examined over 100% of the weld length, (b) the manual volunetric examinations of the reportable defects in the lower portion of the A-B weld are performed and evaluated, and (c) the Code-required system pressure tests are performed.

The requested relief has been recommended based on the impracticality of conducting manual ultrasonic examinations in radiation fields estimated to exist in the examination area. If actual radiation fields are lower, for example if the core were removed, more extensive examinations should be conducted.

It is further recommended that improvements in automated 1

ultrasonic inspection technology be monitored and that the entire length of the lower portion of the A-B weld be examined if techniques become available during the 10-year interval.

References References 16, 20, 21 and 22.

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3.1.1.3 Relief Request No. I-00006, RPV Seam Weld A-C, Category B-A, Item 81.11.

I Code Requirement All pressure-retaining circumferential shell welds in the reactor '

vessel shall be volumetrically examined in accordance with Figure IW8-2500- 1 over essentially 100% of their lengths during the first inspection interval. Examinations may be performed at or near the end of the interval.

Code Relief Request

Relief is requested to delete al! manual ultrasonic inservice i inspection on the areas of weld A-C affected by the presence of intruding N6 and N12 nozzles. This length is approximately 23.5 ft of the approximately 69 ft circumference of the A-C weld (138 feet for both sides); therefore, approximately 83 percent of the required examinations is performed by the remote method.

4 Proposed Alternative Examination

] ,

l None other than Code required pressure tests.

Licensee's Basis for Requesting Relief The circumferential weld A-C joins together ring 2 and 3 of the ,

i

' reactor pressure vessel (RPV) and is located 388.84 inches above vessel zero. The top of the core is located 366.61 inches above vessel zero, and the seam is 22.23 inches above the top of the core.

! Both the upper and lower portions of seam A-C are typical -

circumferential shell welds. Automated ultrasonic examination procedure and equipment have been developed which will permit the required inservice volumetric inspections to be performed remotely.

However, due to the nozzle interference with the automated ultrasonic l equipment, a remote examination is not possible. Therefore, this i

portion of the weld must be examined by the manual procedure.

Except for areas around the N6 nozzles and the N12 nozzles, seam A-C was examined with the automated ultrasonic equipment. Approxi-mately 23 inches on each side of each N6 nozzle's centerline and 18 inches on each side of each N12 nozzle's centerline required manual ultrasonic inspection. The nozzles are located so close to the seam that they interfere with the automated equipment.

l The containment design of Grand Gulf Nuclear Station Unit 1 is designated Mark III. A feature of this design is that an annulus '

, space of approximately 30 inches width exists between the reactor 19

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pressure vessel outer circumference and the biological shield wall inner circumference.

The examiners must enter this annulus space to perform the manual ultrasonic examination on weld A-C. Since we have no actual radiation data from GGNS, we based our estimates on data from other plants. The information we received from other BWR plants indicated the A-C weld is located in a radiation field of approximately 8-10 R/hr. These are conservative estimates for the end of_ the first fuel cycle (first refueling outage). The actual exposures will probably be less; how-ever, the area and contact readings can be expected to rise as the plant ages.

1 The total length of weld subject to manual inspection is approxi-mately 23.5 feet; therefore, it will take approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to perform the manual inspection on weld A-C. This estimation. includes 4

an allowance for the extra time required by personnel due to wearing protective clothing, but it does not include the time required by personnel to enter and exit the annulus area.,

As stated earlier, personnel performing manual ultrasonic examinations on this particular weld will be subject to very high radiation fields, in the neighborhood of 8-10 R/hr. Even assuming

i. GGNS daily and weekly administrative whole-body limits are waived and the inspectors are allowed to accumulate their GGNS quarterly admini-strative limit of 2500 mrem for one dose, personnel stay times in this -

area would be less than 20 minutes. .

Our calculations indicate. that seam A-C will require approxi-mately 32,000 millirem of personnel exposure per 10-year interval or 10,666 millirem each inspection period. These figures do not include radiation exposure to support personnel.

Due to the amount of weld area required to be examined and the nature of the radiation sources in the area, tenporary shielding is not practical. Such shielding would need to protect the entire body, -

would be heavy and difficult to move, and would require significant exposure to erect in and remove from the annulus region.

Relief from the manual ultrasonic inservice inspection of weld A-C is requested for the following reasons:

1. The entire reactor pressure vessel was subject to a hydrostatic pressure test in accordance with ASME Section III.

2. The entire reacter pressure vessel will be subject to a system leakage test at each refueling outage and to a system hydrostatic test each inservice inspection interval in accordance with the requirements of the ASME Section XI.

i

! I 20 O

3. The entire A-C weld, except for the areas around N6 and N12 nozzles, was examined by remote ultrasonics from both sides of the weld as a preservice inspection in accordance witn ASME Section XI, and 19 recordable, but not reportable, indications were found. Manual ultrasonic inspection of the areas not scanned by the remote equipment revealed no

~

recordable indications.

4. The accessible A-C weld will be examined by remote ultra-sonics during the first inservice inspection interval (first 10 years) in accordance with the requirements of ASME Section XI. Complete coverage of the weld around the N6 and N12 nozzles will be attained with the 00, 450, and 600 transducers scanning from the side of the seam opposite the nozzles.

5. Based on radiation data from other plants, calculations indicate that seam A-C will require approximately 32,000 millirem of personnel exposure per 10-year interval or 10,666 millirem each inspection period. These figures do not include radiation exposure to support personnel.

Evaluation _

Shell weld A-C is a circumferential weld that joins ring 2 and ring 3 in the cylindrical portion of the vessel at an elevation sligntly greater than 22 inches above the top of the core. The weld is accessible through an annular gap approximately 30 inches wide between the RPV and the biological shield wall. The weld will be examined over approximately 83% of the weld length using remote,

~

automated ultrasonic inspection equipment. However, the automated 't equipment, as currently configured, cannot be used to examine the areas close to the N6 and N12 nozzles. Manual ultrasonic examination of the weld is the only alternate currently available in the areas ~

adjacent to the nozzles.

The A-C weld was examined over 100% of the length during the preservice examination using both manual and automated ultrasonic me thods . Based on this examination, the licensee estimates that approximately 4 man-hours are required exclusive of exit and entry time to conduct manual ultrasonic scans of the A-C weld near the N-6 and N-12 nozzles. The radiation field in the region of the A-C weld has been estimated to be 8-10 R/hr based on radiation fields in similar BWRs following 40 months of operation. Combining the estimates of inspection time and radiation level, the licensee estimates that a total exposure of 32,000 millirem would be required

, to conduct the inservice examination of the A-C weld in the areas of the N-6 and N-12 nozzles.

21

' j Based on the high radiation exposure that would result' from l conducting manual volumetric examinations of the A-C weld in the area of the N-6 and N-12 nozzles, the examination is considered impracti-cal. Considaring that 100% of the Code required volume from both sides of the weld was volumetrically examined during the preservice examination, that 83% of the weld length will be examined during the first interval, that a VT-2 visual examination of Class 1 components will be conducted during system hydrostatic and leakage tests, the necessary level of structural reliability can be achieved.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the circumferential shell weld discussed above, adherence to the Code requirements is impractical. It is further concluded that the proposed examinations will provide the necessary assurance of structural reliability during this interval. Therefore, relief is recommended as requested provided:

(a) the volumetric examinations are performed to the maximum extent practical and (b) the Code-required system pressure tests are performed. .

It is further recommended that improvementi in automated ultrasonic inspection technology be monitored and that the entire length of the lower portion of the A-C weld be examined if techniques become available during the 10-year interval.

References i

References 16, 20, 21 and 22. _

22 4

3.1.1.4 Relief Request No. I-00008, Control Rod Drive and In-Core Housing Welos and Flange Bolting, Category 8-0, Item 614.10, ano Category B-G-2, Item B7.10.

Code Requirement Welds in Control Rod Drive Housings, Item B14.10. The welds in 10% of tne peripneral control roo orive nousings snall be surface or volumetrically examined in accordance with Figure IWB-2500-18 during

) each inspection interval. The examinations may be performed at or near the end of the inspection interval.

Bolts, Studs, and Nuts in Reactor Vessel, Item B7.10. The surfaces of all ooits, stuos, and nuts 2 in, or less in ojameter in the reactor vessel 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.

Code Relief Recuest Relief is requested to exempt from inservice inspection, the peripheral CRD housing welds (tube-to-tube, tube-to-flange), the eight _

(8) bolts associated with each flange of 193 CRD housings, and the four (4) bolts associated with each flange of 58 incore housings.

Proposed Alternative Examination None other than code-required system pressure tests.

' Licensee's Basis for Requesting Relief The weld areas and bolting are not accessible for inspection unless the control rod drive support structure is removed. A 3600 surface examination required by the Code cannot be accurately accomplished from the outside due to interference from adjacent CRD housings. Inspection of the weld from the inside of the CR0 housing would require that the control rod drive mechanism be removed, which could result in damage to the drive. With removal of the drive, a ,

small amount of reactor water would escape to the CRD cavity area, '

possibly causing contamination of personnel and equipment. The time frame associated with the CRD support structure removal and CRD mechanism would be approximately six (6) man-hours per drive. Dosage received by personnel in this interval cannot justify the inspection process to possibly find a fault which could be discovered by excessive leakage in the drywell sump monitored per Operating License Manual (Technical Specification) limits in effect.

l l

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Request for exemption from inservice inspection should be granted for the following reasons:

1. The peripheral CR0 housing welds have been examined by radiography and liquid penetrant methods and have been hydrostatic tested in accordance with ASME Section III Code requirements.

2. All incore and CRD housing bolting has been examinec in accordance with the requirements of ASME Section III, which exceed the Section XI (VT-1) visual examination requirements.

3. The welds and bolting will be subject to hydrostatic testing per the requirements of ASME Section XI upon completion of each outage.

4. If the welds and/or bolts fail while in operation, the maximum leakage rate, by calculation, will occur at the peripheral CR0 housing tube-to-flange weld. The maximum calculated leak rate is 681 gpm. By criteria established in Subarticle IWB-1200, " exemptions by makeup capacity," the normal makeup capability for GGNS is 878 gpm, which exceeds the calculated maximum leakage.

Leak detection is provided with the leakage detection 5.

system, with continuous monitoring in the control room.

6. The CRD housing supports would prevent ejection of the housings in case of total failure of the welds or bolts.

7. Removal of the control rod drive support structure would ,

result in hardships with no compensatory increase in the "_

level of quality and safety.

t Evaluation Relief was requested and granted to eliminate examination of the control rod drive and in-core housing weld and flange bolting during

) the preservice examination. The primary basis for establishing the structural integrity of these components is, therefore, the construc-tion code examinations and system leakage and hydrostatic tests. ASME Code Section III was the construction code and, as a result, the peripheral CRD housing welds were examined by radiography and liquid penetrant methods. Examination of the bolting in accordance with

) Section III was also completed.

The licensee has shown that failure of a CR0 housing and/or bolt would result in a leak rate from the system which would be less than the capacity of the makeup system. In addition, CR0 housing supports 4

24 4

would prevent ejection of the housings in case of total failure of the welds or bolts. During operation, leak detection equipment .is moni-tored continuously in the control room so leaks can be detected at an earlier stage.

Further examination of the CRD housings or bolts in accordance with the Code would require removal of the CRD support structure and driver. Removal of the CR0 support structure and driver would result in some personnel radiation exposure.

i Relief appears to be warranted on the basis that conducting the proposed examinations would be impractical. The preservice structural reliability was established through the construction examinations.

Hydrostatic and leakage tests can be used to continually update the structural reliability during ISI. In any event, failure of a CRD housing or the bolting would not result in a leakage rate above the a makeup capability of the system.

If it becomes necessary to remove the CRD support structure in conjunction with other maintenance activities, it appears prudent to conduct surface examination of the accessible portion of housing welds and visual examination of the bolting. Correspondingly, if a CR0 mechanism were removed for other purposes, volumetric examination of the housing welds should be conducted.

~

Conclusions and Recommendations i Based on the above evaluation, it is concluded that for the welds and bolting discussed above, adherence to the Code requirements is impractical. It is further concluded that the proposed examinations will provide the necessary assurance of structural reliability during this interval. Therefore, relief is recommended as requested provided:

)

(a) the Code-required system pressure tests are performed and (b) the Code-required examinations are performed on the acces-sible areas if the CR0 support structure or CR0s are removed for other maintenance activities.

)

References References 5, 7, 15, 16, and 20.

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3.1.2 Pressurizer Not applicable to BWRs.

3.1.3 Steam Generators and Heat Exchangers No relief requests.

3.1.4 Piping Pressure Boundary 3.1.4.1 Relief Request No. I-00007, Piping Welds Within Guard Pipes, Gategory 6-d, items 89.11 ano 69.21.

Code Requirement Circumferential Welds, Item 89.11. For circumferential welds in pipe or nominal pipe size 4 in, ano greater, surface plus volumetric examinations shall be performed in accordance with Figure IWB-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 otner components where the stress levels exceed the following limits 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. j (c) All dissimilar metal welds between combinations of l

(1) carbon or low alloy steels to high alloy steels, (2) carbon or low alloy steels to high nickel alloys, and (3) high alloy steels to high nickel alloys.

(d) Additional piping welds so that the total equals 25% of the l circumferential joints in the reactor coolant piping system.

This total does not include welds excluded by IWB-1220. These additional welds may be located in one loop (one loop is currently defined for both PWR and BWR plants in the 1977 Edi tion) .

t 26

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Circumferential Welds, Item B9.21.

For circumferential welds in pipe of nominal pipe size less than 4 in., surface examinations shall be performed in accordance with Figure IWB-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.

4 (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) primary plus secondary stress intensity of 2.4Sm for i

ferritic steel and austenitic steel, and (2) cumulative usage factor U of 0.4. -

(c) All dissimilar metal welds between combinations of (1) carbon or low alloy steels to high alloy steels, (2) carbon or low alloy steels to high nickel alloys, and (3) high alloys steels to high nickel alloys.

t (d) Additional piping welds so that the total equals 25% of the circumferential joints in the reactor coolant piping system.

IWB 1220. These This total does additional weldsnot include may weldsinexcluded be located one loopby(one loop is t currently defined for both PWR and BWR plants in the 1977 .

Edition).

t .

Code Relief Request Relief is requested to permit ultrasonic examination of only the accessible areas of the welds, except as noted under alternate

) examination, for the piping welds listed in Table 1.

Proposed Alternative Examination The accessible length of each weld will be ultrasonically

) examined in accordance with ASME Section XI, Table IWB-2500-1 Examination Category B-J. Should signs of weld deterioration or discrepancies be noted during regular inspections, evaluation of the conditions will be made.

4 27

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Table 1 REQUEST FOR RELIEF NO. I-00007 Item System Wel d Ph. ISO th. Pi pe Componen t Limi ted Type Class Reason for No . Ph . Size Area Scan Limi ta tions 1 B21 G12-Cl-A MS-ll-9 28" Pipe to Pipe 53% T 1 Guard Pipe 2 B21 G12-Al-A liS-11-3 28" Pipe to Pipe 53% T 1 Guard Pipe 3 8 21 G12-01 -A MS-11 -12 28" Pipe to Pipe 53% T 1 Gaard Pipe 4 B21 G12-81-A 11S - 1 1 - 6 28" Pipe to Pipe 53% T 1 Guard Pipe 5 B21 W2 FW 1 24" Pipe to Pipe 50% T 1 Giard Pipe g 6 B21 W18 Rl-11-7 24" Pipe to Pipe 50% T 1 Guard Pipe 7 B21 W9 SD-ll-2 3" Pipe to Pipe 50% T 1 Giard Pipe 8 E12 W47 Ril-l l-1 20" Pipe to Pipe 50% T 1 Girard Pipe 9 E51 W12 R I 9 6" Pipe to Pipe 50% T 1 Giard Pipe 10 E51 W7 RI-11-3 10" Pipe to Pipe 50% T 1 Guard Pipe 11 E51 W18 CU-ll - 3 6" Pipe to Pipe 50% T 1 Guard Pipe

. i s .,.lk,&, -

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Licensee's Basis for Requesting Relief The circumferential welds joining the flued head and the process pipe are encapsuled by the portion of the guard pipe which protrudes beyond the containment. To comply with the inservice inspection requirements of ASME Section XI, two 4 x 6 inch elliptical access ports spaced 180 degrees apart are provided for access to the welds.

After fabrication and installation of the process pipe, guard pipe and the flued head, it was determined that the entire length of the weld is not accessible through the two access ports. Approximately 50% of t

the weld is accessible for inservice inspection.

Exemption is requested for the inservice inspection of inaccessible portions of welds located inside guard pipes for the.

following reasons:

1 1. All but two of these lines were designed to high energy pipe break criteria. The exceptions are Q1E12G012W47 and QlE51G00lW12 which are classified as moderate energy. pipes.

2. These welds were designed and fabricated in accordance with ASME Section III, Class 1 requirements and were examined by radiographic and liquid penetrant techniques.

3. These welds have satisfactorily passed both liquid penetrant and _

ultrasonic examination in accordance with ASME Section XI, Class 1 requirements.

4. Class 1 isolation valves in the process pipe on both sides of the guard pipes are capable of completely isolating each pipe in the event of a pipe failure. ,

5. The guard pipes have been designed and constructed in accordance ;d with ASME Section III, Class 2 requirements and were 4 hydrostatically tested in accordance with ASME Section III, Class 2 requirements. _

6. The guard pipes are open to the drywell environment; thus, any leakage due to weld failure will be contained within the drywell. The guard pipes will prevent any leakage from escaping i

to the Containment Building.

7. The process pipes inside the guard pipes were hydrostatically tested in accordance with ASME Section III, Class 1 requirements.

8. The process piping inside each guard pipe assembly will be subject to periodic pressure tests in accordance with ASME Section XI, Table IWB-2500-1, Category B-P, requirements.

I

)

s .

d)

Evaluation

.The welds for that relief is requested are circumferential welds encapsulated in guard pipes which protrude beyond the containment wall.

Two 4 x 6-inch elliptical access ports spaced 1800 apart were provided in the guard pipes to permit examination of the circumferential welds.

However, with currently available ultrasonic equipment, only about 50%

of the length of each weld can be volumetrically examined through the access ports. The entire length of each weld can be surface examined.

The process piping for which relief is requested was designed to ASME Section III, and the circumferential welds were radiographed and liquid penetrant checked during construction. The sections of process piping within the guard pipes can be completely isolated by Class 1 valves on both sides of the guard pipes. The guard pipes are hydro-statically tested Class 2 piping designed to divert any leakage from 4

the process pipes to the drywell.

l The circumferential welds in the process pipe will be volumetri-cally examined over about 50% of the weld length in accordance with Section XI. In addition, the process pipe weld inside the guard pipe l

l j will be surface examined and visually examined during system pressure i tests.

1 Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds 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.

i Therefore, relief is recommended as requested provided:

'^

i (a) the volumetric examinations are perforned to the maximum extent practical and

- l (b) the Code-required surface examinations and system pressure

. tests are performed.

References

References 1 and 16.

1

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i 30

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p. -

, - - - c,. .T '-w---- -

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i . ,

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3.1.4.2 Relief Request No. I-00010 and I-00012, Pressure Retaining-Fiping ~

Welas on RnR, RCIC, MS, RECIRC, and RWGU Systems, Catenory B-d, items 69. i i ana b9. id. ' '

' e

. f.

Code Requirement .

Circumferential Welds, Item B9.11. For circumferential. welds in pipe of nominal pipe size 4 in, ano greater, surface plus volumetric examinations shall be performed in accordance with Figure IW3-2500-8 .

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 branc'n run connected to vessels. -

,e ,,

(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. -

r (1) primary plus secondary stre:s i.ntensity'$f 2.4Sm for ferritic steel and austenitic steel, and (2) cumulative usage factor U of 0.4$ _

(c) All dissimilar metal welds between combinations ~ of (1) carbon or low alloy steels tolhigh allby steels, (2) carbon or low alloy steels to high nickel alloys, and (3) high alloy steels to high nickel alloys. ,

4 (d) Additional piping welds so that t'h'e' total equals 25%'of the circumferential joints in the reactor coolant piping system. _

This total does not include welds excluded by IWB-1220. These additional welds may be located in one loop (one loop is currently defined for both PWR and BWR plants in the 1977 Edition) . ,

Longitudinal Welds, Item Bl.12. for longitudinal welds in' pipe of nominal pipe size 4 in, and greater, surface plus volumetric examinations shall be performed i'a accordance'with Figure 'IWB-2500-8 .

for at least a pipe-diameter length, but not more than 12 in.,fof each l longitudinal weld intersecting the circumferential welds required to ,

J be examined.

I i

m i y

31 j; i

~

~

t .

Code Relief Request Relief is requested to exempt from volumetric examination the inaccessible portions of the Class 1 welds listed on-Table 2.

Proposed Alternative Examination All the welds identified in Table 2 will be inspected twice by volumetric examination during the 10-year interval as discussed by NRC in GGNS Safety Evaluation Report, Supplement No. 2. (In this supple-ment, the staff noted that "An augmented program may include increasing the sample size and/or frequency of accessible welds to ensure that generic degradation is not occurring in these piping welds.")

Licensee's Basis for Requesting Relief Portions of welds that were preservice ' examined have physical obstructions due to design. Due to this limited accessibility, it is impractical to volumetrically examine 100% of the welds listed on Table 2.

Request for an exemption should be granted for the following -

reasons:

1. The inaccessible portions of listed welds were examined by radiography, passed in accordance with ASME Section III, Class 1 i

requirements.

2. The inaccessible portions of listed welds were surface examined (magnetic particle or liquid penetrant), passed in accordance ,

with ASME III and/or XI, Class 1 requirements. y

3. The inaccessible portions of listed piping welds will be subject to a system leakage test after each refueling outage for Class 1 in accordance witn ASME Section XI requirements.

4 The inaccessible portions of listed piping welds will be subject.

to a system hydrostatic test each inspection interval in i accordance with ASME Section XI, Class 1 requirements.

5. The portions of listed welds inaccessible for volumetric examination will be surface examined each inspection interval, in accordance with ASME Section XI.

6. Accessible portions of listed welds will be volumetrically and surface examined each inspection interval in accordance with ASME Section XI. Should indications be found, an engineering i evaluation will be made to determine if the inaccessible portions of listed welds have been affected.

32 m ~

Table 2 RE(NEST FOR RELIEF N0. I-00010 Item System Weld th. ISO tb. Pipe Canponent Limi ted Type C1 ass Reason for No. tb . Size Area Scan Limi ta tions 1 E12 G014-FW-44 R il-8-8 6" Valve to Elbow 42% T 1 Radius of Elbow B21 G9-C1-B-L/B MS-11-8 28" Elbcw to long 32% T 1 Rupture Restraint 2

Seam 3 B21 Gil-DI-B-L/B MS-11-11 28" Elbow to long 32% T 1 Rupture Restraint Seam B21 G8-Al-B-L/B MS-11-2 28" Elbow to long 32% T 1 Rupture Restraint w 4

" Seam B21 G030 FW-23 FW 2 24" Valve to Pipe 5% T 1 So ck-0 -Le t 5

6 B21 G030 Rf-36 Rf- 8-4 24" Valve to Pipe 5% ,

T 1 Sock-0-let 7 B21 G026 FW-17 FW-11-7 6" Pipe to Tee 5% T 1 So ck-0 -Le t 8 B21 G001-W4 MS'-11-3 28" Valve to Pipe 13% T 1 Rupture Restraint 9 B21 G001-W4 MS 9 28" Valve to Elbow 13% T 1 Rupture Restraint B21 G9-Cl-3-L/A MS-11-8 28" Elbcw Long Seam 19% T 1 Rupture Restraint 10 1

6 . ij. "

. . _ .- m. _ _ _ _

P

~

Table 2 ( Continued)

REQUEST FOR RELIEF NO. I-00010 Pi pe Component Limi ted Type Class Reason for Item Sys tem Weld m. ISO th. Limi ta tions Size Area Scan 11 0 . m.

Pipe to 5% T Sweep-0-Le t 11 B33 G5-B1-E RR-ll-9 24" 1 Elbow Sweep Let 4" Elbow to Tee 24% T 1 Tee 12 B33 G024-W2 RR-11-19 CU-8-7 4" Elbow to 8% T 1 Elbow Radius 13 G33 G002-W179 Ven turi 28" Elbow long Seam 19% T 1 Rupture Restraint y 14 B21 Gil-Dl-B -L/A MS-11-ll 28" Valve to Pipe T 1 Rupture Restraint 15 B21 G001 W9 MS-ll-12 13%

28" Elbow long Seam 19% T 1 Rupture Restraint 16 021 G8-Al-B-L/A MS-11-2 R I-i l-4 6" Pipe to Valve 21 % T 1 Elbow Radius 17 E51 G001 W40 ,

RI-ll-7 6" Valve to Elbow 25% T 2 Elbow Radius 18 E51 502 e

l l ,

. _ = =

.- = _ _

. . I

7. Leak detection is provided, by way of leakage detection system with continuous monitoring, for the RHR, RCIC, MS, RWCU and RECIRC systems.

8. The failure of any of these welds would have no adverse effect on plant safety as there is isolation capability and/or shutdown capability as part of the plant design.

Evaluation Eighteen Class 1 piping welds in piping 4-inch diameter and greater for which volumetric examination over 100% of the weld is precluded have been identified in Relief Requests I-00010 and I-00012. Relief from full volumetric examination of the welds is reques ted.

All the welds were fabricated to ASME 'Section III and were accepted following Section III radiographic, surface, and hydrostatic testing. The accessible portions of the welds were volumetrically examined, and complete surface examinations were conducted during the PSI in accordance with Section XI. The licensee will conduct volu-metric examinations of the accessible portions of all the welds twice during the inspection interval . The licensee has also committed to conduct an engineering evaluation of the entire weld condition if indications are found during the volumetric examinations. The Code- -

required surface examination and pressure tests will be conducted on all the welds.

The licensee also states that failure of any of the subject welds would not have an adverse affect on plant safety since all of the areas can be isolated or the system shut down in a manner consistent a, with plant design. In addition, continuous leak detection monitoring . Ach is provided on all of the systems containing inaccessible welds. .y Conclusions and Recommendations _

Based on the above evaluation, it is concluded that for the welds 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 volumetric examinations are performed to the maximum extent (a) practical twice during the interval and (b) the Code-res aired surface examinations and system pressure tests are performed.

References References 1 and 16.

35

3.1.5 Pump Pressure Boundary No relief requests.

3.1.6 Valve Pressure Boundary No relief requests.

m w

6 0

t 1

i 36

3.2 CLASS 2 COMP 0NENTS Subsections IWA and IWC of the Code govern the examination of Class 2 piping and components. Specific requirements are given in Table IWC-2500-1.

3.2.1 Pressure Vessels and Heat Exchangers No relief requests.

I 3.2.2 Piping Pressure Boundary

) 3.2.2.1 Relief Request No. I-00002, Pressure Retaining Weld in Scram Discnarge Volume Piping, Category C-F, Item C5.21.

Code Requirement .

One hundred percent of each circumferential weld over 1/2 in.

j 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 i

a. all welds at locations where'the stresses under the loadings resulting from Normal and Upset plant conditions as calculated by the sums 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. additional welds at structural discontinuities (see (f) below) .

such that the total number of welds selected for examination includes 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;

3. 50% of the main steam system welds;

4. 25% of the welds in all other systens.

l 37 i

O a

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 4

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; Code Relief Request -

~ Relief is requested to delete the requirement to volumetrically examine the portion of weld no. 91 inaccessible due to the installed bracket.

Proposed Alternative Examination The accessible portion of the weld will be volumetrically and surface examined each inspection interval in accordance with ASME Section XI, Table IWC-2500-1. Should indications be found, an engineering evaluation will be made to determine if the inaccessible portion of this weld has been affected. The inaccessible portion of the weld represents approximately 14 percent of the total weld length. _

Licensee's Basis for Requesting Relief The installation of a bracket directly above the circumferential weld no. 91 has limited the accessibility to the weld for ultrasonic  %,.

examination. The inaccessible portion of the weld is approximately 7 5.5 inches of the circumference of the 12 inch NPS pipe, or 14 percent of the weld length. '

Exemption from inservice volumetric (ultrasonics) examination of -

the inaccessible portion of this weld is requested for the following reasons:

1. This weld was examined by radiography and passed in accordance with ASME Section III, Class 2 requirements.

2. This weld was surface examined (liquid penetrant) and passed in accordance with ASME Section XI, Table IWC-2500-1, Category C-F requirements.

3. The accessible portion of this weld was volumetrically examined using ultrasonic techniques and passed in accordance with ASME Section XI, Table IWC-2500-1 requirements.

4. This weld has been hydrostatically tested in accordance with ASME Section III, Class 2 requirements.

e 38

5. This weld will be subject to a system functional test and a system hydrostatic test in accordance with ASME Section XI, Table IWC-2500-1 requirements.

6. The portion of this weld inaccessible for volumetric examination will be surface examined each inspection interval in accordance with ASME Section XI, Table IWC-25000-1, Category C-F requirements.

Evaluation Access to circumferential weld no. 91 in the Scram Discharge Volume Piping has been restricted by installation of a bracket directly above the weld. Approximately 5.5 inches of weld length on the surface of the 12-inch NPS, or about 14% of the weld length, is not accessible.

Weld no. 91 was fabricated and successfully radiographed and hydrostatically tested in accordance with ASME Section III Class 2 requirements. During the PSI, the accessible portion of the weld was volumetrically examined with ultrasonic equipment, and the complete length of the weld was surface examined with liquid penetrant. _

During the first interval ISI examination, the weld will be volumetrically examined over all but 14% of the total weld length, surface examined over the entire weld length, and visually examined during Code-required pressure tests.

Complete examination of weld no. 91 using volumetric techniques I?

appears to be impractical due to access restrictions. The alternate "'

examinations should be adequate to verify structural reliability.

Conclusions and Reconsendations Based on the above evaluation, it is concluded that for the weld discussed above, acherence to the Code requirements is impractical.

It is further concluded that the proposed examinations will provide the necessary assurance of structural reliability during this in terval . Therefore, relief is recommended as requested provided:

(a) the volumetric examinations are performed to the maximum ,

extent practical and l (b) the Code-required surface examinations and system pressure tests are performed.

References References 1 ano 16.

39

o 3.2.2.2 Relief Reouest No. I-00003, Circumferential Dissimilar Metal Welo, Category C-F, Item C5.21.

Code Requirement 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

a. all welds at locations where the stresses under the loadings resulting from Normal and Upset plant conditions as calculated by the sums 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. additional welds at structural discontinuities (see (f) below) such that the total number of welds selected for examination includes the follewing 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 ,

~j core cooling systems; . .:e n

3. 50% of the main steam system welds; -

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 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; Code Relief Request j Relief is requested to exempt RCIC turbine exhaust sparger weld 1

(FW-13) from inservice inspection, except as noted in alternative t 'ing.

1 40 9

Proposed Alternative Examination No alternative testing is proposed at this time. However, if the suppression pool is drained for other reasons, inspection of the weld will be performed.

Licensee's Basis for Requesting Relief

1. The subject weld has been volumetrically examined by radiography and found acceptable in accordance with ASME Section III, Class 2 requirements.

2. The subject weld will only be exposed to design pressure and temperature a small percentage of the time the plant operates.

The balance of that time it will see static conditions at near atmospheric pressure, along with relatively low temperatures.

3. The open area of the sparger hole is over 300% of the cross-sectional area of the pipe itself. The steam will flow through the sparger relatively unimpeded, permitting only low pressure buildup in the pipe.

4. The subject weld is in constant contact with the water in the suppression pool. When the suppression pool is at its lowest -

level (Post-LOCA 111'-4 3/4"), the subject weld is submerged under 6'-4 3/4" of water.

Evaluation Relief from inservice examination of the 20" circumferential ~:

weld, W-13, between Reactor Core Isolation Cooling (RCIC) line -

• 20"-HBB-53 and 20"-HCB-27 is requested. The basis for requesting ,

relief is that the external surface of the weld is submerged about 6' 4-3/4" below the water level in the suppression pool. Since the -

suppression pool is not normally drained, examination of weld W-13 is impractical. If the suppression pool were drained for other reasons the weld would be examined.

The subject weld was radiographed in accordance with ASME,Section III, Class 2 following construction. Since loading on this pipe segment is expected to be infrequent and of low level, there appears to be little chance of inducing failure. Accordingly, the construction inspection should provide a reliable basis for verifying the structural integrity of the piping. The licensee has committed to volumetrically examine the weld if the suppressed pool is drained.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds discussed above, adherence to the Code requirements is impractical.

It is further concluded that the examinations conducted during 41

-~ ..

4 fabrication provide the necessary assurance of structural reliability during this interval since no significant mechanism for inducing failure is postulated.

1 References .

References 1 and 16.

l.

3 i

p 3

i

}

l

/

a c.- r ,

j n . ..,

1 i

4 1

4 J

42 I

- - - . . - - , . , . . - , , , - ,-n- ..n, ,,, , , . ~, n -., _ , - , , . , , . - , , . .-e,---. , . , , - , - - , , . , , - - . ~ , - - . - ,

3.2.2.3 Relief Request No. I-00010, Pressure Retaining Piping Welds on KLic dystem, Lategory L-r, item Lo.21.

Code Reauirement 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

a. all welds at locations where the stresses under the loadings resulting from Normal and Upset plant conditions as calculated by -

the sums 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. additional welds at structural discontinuities (see (f) below) such that the total number of welds selected for examination includes 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 76 core cooling systems; si$1

.;$f

3. 50% of the main steam system welds;

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 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; 1

43

. o Proposed Alternative Examination All the welds identified in Table 3 will be inspected twice by volumetric examination during the 10-year interval as discussed by NRC in GGNS Safety Evaluation Report, Supplement No. 2.

Licensee's Basis for Requesting Relief Portions of welds that were preservice examined have physical obstructions due to design. Due to this limited accessibility, it is .

impractical to volumetrically examine 100% of the welds listed on Table 3.

I Request for an exemption should be granted for the following reasons:

1. The inaccessible portions of listed welds were examined by radiography, passed in accordance with ASME Section III, Class 2 requirements.

2. The inaccessible portions of listed welds were surface examined (magnetic particle or liquid penetrant), passed in accordance with ASME III and/or XI, Class 2 requirements.

3. The inaccessible portions of listed piping welds will be subject to a system leakage test after each refue,ing outage for Class 2 in accordance with ASME Section .XI requirements.

4. The inaccessible portions of listed piping welds will be subject to a system hydrostatic test each inspection interval in ,

accordance with ASME Section XI, Class 2, requirements.

~

5. The portions of listed welds inaccessible for volumetric examination will be surface examined each inspection interval, in accordance with ASME Section XI. -

6. Accessible portions of listed welds will be volumetrically and surface examined each inspection interval in accordance with ASME Section XI. Should indications be found, an engineering evaluation will be made to determine if the inaccessible portions of listed welds have been affected.

7. Leak detection is provided by way of leakage detection system with continuous monitoring for the RCIC system.

8. The failure of any of these welds would have no adverse effect on plant safety as there is-isolation capability and/or shutdown capability as part of the plant design.

44

n _. . . - - _ .. =

Table 3 REQUEST FOR RELIEF NO. I-00010 Item System Wel d M). ISO N>. Pipe Component Limi ted Type Class Reason for No. N). Si ze Area Scan Limi ta tions 1 E51 G004 8-8-1 RT-8-2 6" Elbw to Elbow 32% T 2 Elbw Radius 2 E51 G004 7-8-4 R I-8-1 10" Red cer to Tee 19% T 2 Tee 3 E51 G004 7-8-9 R I-8-1 10" Reducer to Tee 19% T 2 Tee N 4 E51 G004 7-8-8 RI-8-1 10" Redscer to' Tee 19% T 2 Tee 5 E51 G001 WI R I-8-12 6" Valve to Elbw 10% T 2 Elbw Radius

Evaluation Five Class 2 circumferential piping welds having a nominal wall thickness over 1/2 inch for which volumetric examination over 100% of the weld is precluded have been identified in Relief Request I-00010.

Relief from full volumetric examination of the welds is requested.

All the welds were fabricated to ASME Section III and were accepted following Section III radiographic, surface, and hydrostatic testing. The accessible portions of the welds were volumetrically examined, and complete surface examinations were conducted during the PSI in accordance with Section XI. The licensee will conduct volumetric examinations of the accessible portions of all the welds twice,during the inspection interval. The licensee has also committed to conduct an engineering evaluation of the entire weld condition if indications are found during the volumetric examinations. The .

Code-required surface examination and pressure tests will be conducted on all the welds. ,

~

The licensee also states that failure cf any of the subject welds would not have an adverse affect on plant safety since all of the areas can be isolated or the system shutdown in a manner consistent with plant design. In addition, continuous leak detection monitoring is provided on all of the systems containing inaccessible welds.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds 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: ,' t I

(a) the volumetric examinations are performed to the maximum extent practical twice during the interval and -

(b) the Code-required surface examinations and system pressure tests are performed.

References References 1 and 16. ,

t 6

46

• O 3.2.2.4 Relief Request No. I-00011, Thermal Tee Sleeve Welds' Category L-P, Item C5.21.

J Code Requirement 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 were selected for examination in accordance with the special GGNS-1 selection criteria described in Section 2.4.2 and References 5, 16, 21, and 22.

Code Relief Request Relief is requested for exemption from inservice volumetric (ultrasonics and radiography) inspections of thermal tee welds 00ll-A-1 and 0011-B-1 on the RHR return to RWCU line.

Proposed Alternative Examination The inservice inspection plan requires that the thermal tee sleeve weld be examined twice during the 10-year interval by the -

magnetic particle technique. -

Licensee's Basis for Requesting Relief Due to the design of the diermal tee sleeve, there is not "

sufficient area to perform a meaningful ultrasonic examination. Also, ."

the position of the thermal sleeve, as well as lack of internal access, precludes the use of radiography.

Request for exemption from inservice volumetric (ultrasonics and -

4 radiography) inspection of thermal tee welds 00ll-A-1 and 0011-B-1 on the RHR return to RWCU line should be granted for the following reasons:

l. The thermal tee welds have been volumetrically examined by radiography and found acceptable in accordance with ASME Section III, Class 2 requirements.

2. The thermal tee welds have been surface examined by magnetic particle and found acceptable in accordance with ASME Section XI, Class 2 requirements.

3. The thermal tee welds will be subject to magnetic particle inspection every 10-year interval in accordance with ASME Section XI, Class 2 requirements.

4 47

i

} 4. Thermal tee welds will 'be subject to a system pressure test l in accordance with ASME Section XI, Class 2 requirements.

5. Leak detection is provided for, the main steam tunnel via the 4

leak detection system in the control room which would detect a leak in the thermal tee sleeve weld, i

6. The system design would allow for isolation of the thermal sleeve without adversely affecting plant safety.

t Evaluation

) Relief from volumetric examination of the two thermal tee sleeve l

welds, 00ll-A-1 and 001108-1, located on the residual heat removal (RHR) return from the reactor water cleanup (RWCU) line is requested.

The configuration of the thermal sleeve is such that there is not sufficient area to conduct an ultrasonic examination, and the position of the thermal sleeve and lack of internal access preclude radiography.

l i

The subject welds were radiographed in accordance with ASME, j Section III, Class 2, following construction. During the PSI, the welds were surface examined with magnetic particle inspection i techniques.

• =

/

During the first interval ISI, the thermal tee sleeve welds will be surface examined using magnetic particle inspection twice. The l welds will also be subjected to visual examinations during the .

j Code-required pressure tests.

f 'In addition to ISI examinations, leak detection is provided in @

the main steam tunnel that would detect a leak in the thermal sleeve. ,d If a leak were detected, the thermal sleeves could be isolated without '$

adversely affecting plant safety. v-Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds discussed above, adherence to the Code requirements is impractical.

! It is further concluded that the proposed examinations will provide d

necessary assurance of structural reliability during this interval.

Therefore, relief is recommended as requested provided:

(a) the magnetic particle surface examinations are performed and i

1 (b) the Code-required system pressure tests are performed.

I i References 1

) References 1, 3, and 16.

}

l 48

(

3.2.3 Pump Pressure Boundary 3.2.3.1 Relief Request No. I-00009, Pump Casing Welds, Category C-G, item c6.10.

Code Requirement One hundred percent of all pump casing welds in each piping run i examined under Examination Category C-F snall 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.

Code Relief Recuest. .

Relief is requested to exempt from inservice inspection the inaccessible portiens of the pump casing welds listed on Table 4.

Proposed Alternative Examination _

Code-required system pressure tests will be performed and pumps will be tested at least once per 31 days per Technical Specification

) to ensure operability.

-W Licensee's Basis for Requesting Relief g

+

':za Inaccessible pump casing welds are located where the concrete Yi

) pump support encasement only allows a 3" clearance between the pump . 1 casing and the concrete encasement wall. Due to this limited .

accessibility, it is impractical to surface examine those portions of the welds located within the surrounding concrete pump support encasement.

) Request for an exemption should be granted for the following reasons:

1. These welds have been volumetrically examined by radiography and passed in accordance with the ASME Section III, Class 2 requirements.

2. The accessible length of each applicable weld will be surface examined (magnetic particle method) in accordance with ASME Section XI. Inaccessible cump casing welds are

. located where the concrete pump support encasement only allows a 3" clearance between the pump casing and the 4

49

) .

• O Table 4 REQUEST FOR RELIEF NO. I-00009 LIST OF PUMP CASING WELDS i

l E12 - RHR Pumo "B" Casing (Pump 2.1 E12 C0028)

WELDS SURFACE THAT SHALL BE EXAMINED D H-1 0H-4 DH-7 DH-25 SB-5 DH-2 DH-5 DH-11 SB-3 SB-6 DH-3 DH-6 DH-12 SB-4 SB-7 I WELDS THAT CAN BE PARTIALLY EXAMINED WELDS THAT CANT 0T BE SURFACE EXAMINED SB-2 (18" Accessible, 54" Inaccessible) ,

53-1 (Inaccessible) .

E21 - LPCS PUMP CASING (Pump 2.1 E21 C001)

WELDS SURFACE THAT SHALL BE EXAMINED

~

D H-1 DH-4 ' DH-7 DH-27 SB-5 DH-2 DH-5 DH-11 58 -3 SB-6 DH-3 DH-6 DH-12 53-4 SB-7 WELDS THAT CAN BE PARTIALLY EXAMINED WELDS' THAT CANNOT BE EXAMINED .

58-2 (3" Accessible, 69" Inaccessible) S3-1 (Inaccessible) , k.i E22 - HPCS PUMP CASING (Pump tb.1E22 C001) _

WELDS SURFACE THAT SHALL BE EXAMINED D H-4 DH-7 DH-19 58 - 6 2

DH-1 DH-2 DH-5 DH-11 SH-28 SB-7 i

I DH-3 DH-6 DH-12 S8-5 i WELDS THAT CAN BE PARTIALLY EXAMINED WELDS THAT CANNOT BE EXAMINED SB-4 ( 68" Accessible, 4" Inaccessible) 58-1 (Inaccessible)

SB-2 ( Inaccessible)

SB-3 (Inaccessible) 50

concrete encasement wall. Due to this limited accessibility it is impractical to surface examine those portions of the welds located within the surrounding concrete pump support encasement.

3. The failure of these welds, thus leading to failure of the pump, would have no adverse effect on plant safety, as redundant emergency core cooling systems are provided.

4. Annunciators (i.e. low suction pressure, discharge pressure abnormal, etc.) are provided in the control room, along with other system indicators, to alert the operators to abnormal operating conditions.

5. The systems, including the pumps, are tested at least once per 31 days per Operating License Manual (Technical Specification) requirements to ensure operability.

6. Pumps will be subject to a system pressure test in accordance with ASME Section XI, Class 2. requirements.

7. Approximately 87 percent of the welds on the subject pump, which require surface examination, are accessible. Per-formance of the required examinations on these accessible welds should ensure that generic degradation is not occurring in these pump casing welds. _

Evaluation .

Portions of the external pump casing are inaccessible where the ,

pump passes through the concrete support floor on three safety system Jyj pumps. Relief from surface examination of inaccessible casing welds .15 and inaccessible portions of casing welds is requested on three Class ' d4,<

2 pumps. The following summarizes the examinations to be conducted on 'f;;

the three pwnps:

100% Surface Partial Surface No Surface Examination Examination Examination RHR "B" Casing 15 Welds 25% One Weld One Weld (lE12 C0028) ,

LPCS Pump Casing 15 Welds 4% One Weld One Weld l (lE21 C001)

HPCS Pump Casing 14 Welds 94% One Weld Three Welds (lE22C002) 4 51 t

--r- _,m . - - , , - -. - _ , ,

1

. . 1 No less than 83% of the welds on any one pump will be examined, and 87% of the welds on all three pumps will be examined. j The pump casings were fabricated to ASME Section III, Class 2, requirements and radiographed prior to installation.

Pump performance is continuously monitored in the reactor control .

room, and the pumps are performance tested every 31 days.

The requested relief is justified considering that at least 83%

of the casing welds on any one pump will be examined and should give a reliable overall assessment of pump casing integrity. In addition,

- die pump casings will be visually examined during Code-required pressure tests. Pump manufacturing history and on-line operational monitoring provides further assurance that casing integrity will be maintained.

Since the surface examinations can be conducted from either the external or internal surface of the pump casing, an attempt should be made to examine the portions of the casing welds, inaccessible on the external surface, on an internal surface if the pumps are disassembled for maintenance.

Conclusions and Recommendations

! Based on the above evaluation, it is concluded that for the welds i discussed above, adherence to the Code requirements is impractical.

It is further concluded that the proposed examinations will provide . .e necessary assurance of structural reliability during this interval. _ 27dO Therefore, relief is recommended as requested provided:

. . 413

                                                                                                                                            ..v . :

(a) the surface examinations are performed to the maximum extent practical, (b) the Code-required system pressure tests are performed, and (c) the surface examinations are completed from the internal surface if a pump is disassembled for maintenance. References References 1, 3,' and 16. i d 1 52

  - , - - -   ,n..-   , , - - ,      -.-.,--m.         ~ , - - - - -        ,    , - - - -   n -,.-._ , ,. - . .- -   ,m ..     , -,.   , ,

3.3 CLASS 3 COMPONENTS No relief requests. 3.4 COMPONENT SUPPORTS Subsection IWF of the Code governs the examination of component supports. Specific requirements are given in Table IWF-2500-1. 3.4.1 Plant and Shell-Type Supports No relief requests. 3.4.2 Linear Type Supports No relief requests.

                                                            /

r we'

                                                                                  ,  1L fr 53 s :- i mm.                    s --.  - - . .m . - - - _          < any1           =h

1 I 3.4.3 Component Standard Supports 3.4.3.1 Relief Request No. I-00001, Pipe Supports within Guard Pipes, Category F-C, Item F-2. Code Requirement

            .        All supports within the examination boundaries of IWF-1300 having components and piping required to be examined during the first inspection interval by IWB-2500, IWC-2500, and IWD-2500 shall be visually examined (VT-4) each inspection interval. The areas subject to examination are mechanical connections to the pressure-retaining component; weld and mechanical connections to the building structure; weld and mechanical connections at intermediate joints in multiconnected integral and nonintegral support; and spring-type supports, constant load-type supports, snubbers, and shock absorbers.

Code Relief Request Relief is requested to delete all visual examinations for the supports identified as follows: - Guide No. Support System Identification

    -         2                 QlE51G001C09            Reactor Core Isolation Cooling     '

2 QlESlG004C03 Main Steam to RCIC 3 QlE51G001C08 Reactor Core Isolation Cooling Reactor Water Cleanup 3 Q1G33G002C04 3 QlB21G021C02 Main Steam Drain - 2 QlB21G021H02 Main Steam Drain 3 Q1821G531C01 Main Steam 3 Q1821GS31C02 Main Steam , 3 QlB21G531C03 Main Steam 3 Q1821G531C04 Main Steam 4 Proposed Alternative Examination No alternative testing is proposed at this time. i 54 3

Licensee's Basis for Requesting Relief The component supports were provided for the guard pipes as required by stress analysis to act as restraints which limit lateral movement and as dead weight supports for the process pipe within the guard pipe. These inaccessible supports are located inside guard pipes that extend from the drywell wall through the containment building to the auxiliary building. During installation of the process pipe and guard pipe assembly between the drywell wall and the first outboard isolation valve, the supports were accessible for examination because of the installation sequence used. After guard. pipe components were assembled and welded in place, the access to the supports was no longer available. Exemption is requested for inservice visual examination of these component supports for the following reasons:

1. The supports are inaccessible because they are located in guard pipes which are filled with reflective insulation that cannot be removed.

2. These supports were visually inspected in accordance with Section III, Subsection NF.

3. These supports were preservice examined in accordance with _

Section XI.

4. These supports serve as pipe guides to restrict lateral movement of the pipe to which they are attached. In addition, these supports also serve as dead weight support for the process pipe.

Additional restraints are provided outside the no-break zone as required by pipe break analysis. The failure of any of these ' inaccessible supports would not damage the pressure boundary of the process pipe or affect the integrity of the guard pipe. Evaluation The 10 supports for which relief is requested are nonintegral lateral supports for the short, straight sections of process piping within the guard pipe between the containment wall and drywell wall. The primary load on the supports is the deadweight load of the piping. The supports are encased in the guard pipe which is filled with reflective insulation that was not designed for removal. Prior to final assembly of the penetrations, the supports were examined in accordance with Section III and in accordance with Section XI during , PSI. 55

     , - -     -     -.,,sy      .- -    m       , , , _ . _ , . _ . - --
                                                                                                                      --7 y-- . - -...,-

Failure of these nonintegral lateral supports within the guard pipe appears very unlikely. Additional restraints on the process piping are provided outside of the guard pipes in accordance with pipe break analysis. In the very unlikely event of failure of the supports within the guard pipe, the licensee states that the integrity of the process pipe pressure boundary or the guard pipe would not be compromised. Examination of the subject supports is impractical. Failure of the supports is very unlikely but would have no adverse consequence if it did occur. Accordingly, relief from the visual support examination is warranted. Conclusions and Recommendations Based on the above evaluation, it is concluded that for the supports discussed above, the Code requirements are impractical and, if implemented, would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, the requested relief should be granted. References References 1 and 16. b' 3.5 PRESSURE TESTS 7 I No relief requests. 3.6 GENERAL No relief requests. i l 56 ) t

4. REFERENCES

1. L. F. Dale (MP&L) to H. R. Denton (NRC), October 2,1980; PSI Program attached, including Attachments 1, 2, and 3.

2. L. F. Dale (MP&L) to H. R. Denton (NRC), October 16, 1980; submits PSI /ISI relief request from inspection of main steam line welds located in guard pipes.

3. L. F. Dale (MP&L) to H. R. Denton (NRC), April 6,1981; submits abstract of PSI progress.

4 Safety Evaluation Report, NUREG-0831, September 1981; partial review of PSI request for additional information from licensee.

5. L. F. Dale (MP&L) to H. R. Denton (NRC), September 1,1981; attached were PSI /ISI relief requests from inspection of control rod drive housing welds (00002/I-00008), from inspection of pump casing welds (00003/I-00009), from inspection of RHR heat exchanger welds (00004/NA), from injection of thermal tee sleeve welds (00005/I-000ll), from inspection of RC pump casing internal.;

(00006/NA); ASME Section XI Code Exemption Criteria; responses to FSAR questions 121.7,121.10,.121.11,121.12,121.13 and 121.14; and PSI isometric drawings. _

6. L. F. Dale (MP&L) to H. R. Denton (NRC), October 15, 1981; submits PSI /ISI relief request for limited ultrasonic scanned piping welds and information on stainless steel piping systems.

7. Safety Evaluation Report, NUREG-0831, Supplement No. 2, June 1982; h grants PSI relief requests 00001 through 00007. ,%

                                                                                   .dM'

8. L. F. Dale (MP&L) to H. R. Denton (NRC), April 30, 1982; submits PSI /ISI relief request from inspection of RCIC dissimilar metal weld located in the suppression pool. -

9. Safety Evaluation Repart, NUREG-0831, Supplement No. 4, May 1983; PSI program approved, PSI relief request 00008 granted, and operating

, license condition 2.C.(18) added.

10. J. P. McGaughy (MP&L) to H. R. Denton (NRC), May 31, 1983; proposed amendment to operating license to require submittal of the ISI document by April 1, 1984.

11. L. F. Dale (MP&L) to H. R. Denton (NRC) March 16, 1984; proposed

.I amendment to operating license to require submittal of ISI document by August 1,1984.

12. Safety Evaluation Report, NUREG-0831, Supplement No. 5, August 1984; amends operating License to require submittal of ISI document by September 1, 1984.

57 ) 1..

 . .                                                                                         l l

13. L. F. Dale (MP&L) to H. R. Denton (NRC), August 25, 1983; PSI relief request number 24 attached.

14. L. F. Dale (MP&L) to H. R. Denton (NRC), May 11, 1984; PSI relief request number 12.

15. L. F. Dale (MP&L) to H. R. Denton (NRC), May 22, 1984; reserialization of PSI relief request numbers.

16. L. F. Dale (MP&L) to H. R. Denton (NRC), July 25, 1984; ISI Program, Volumes I, II, and III.

17. L. F. Dale (MP&L) to H. R. Denton (NRC), May 11, 1984; ISI sample plan for inspection of pipe supports.

18. J. G. Cesare (MP&L) to H. R. Denton (NRC), June 3,1984; MP&L response

          .to Generic Letter 84-11, inspection of stainless steel piping, IGSCC          ~

augmented inspection. -

19. L. F. Dale (MP&L) to H. R. Denton (NRC), September 20, 1984; PSI /ISI relief request from inspection of RCIC piping weld (00015/I-00012).

20. L. F. Dale (MP&L) to H. R. Denton (NRC), June 29, 1984; submits PSI relief requests I-00001 through I-000ll. _

21. NRC to MP8L, August 22, 1985; RAI - request for additional information.

22. L. F. Dale (MP&L) to H. R. Denton (NRC), October 31, 1985; response to RAI, Reference 21.

iff 16' . O 58

l APPENDfX A p Requirements of Section XI of the Anerican Society = of Mechanical Engineers Boiler and Pressure Code, , 1977 Edition with Addenda through Summer 1979 A.1 CLASS 1 REQUIREMENTS

                                                                           ,                                    1 A.1.1        CATEGORY B-A, PRESSURE-RETAINING WELDS IN REACTOR VESSEL                       '

A.1.1.1 Shell Welds, Item Bl .10 A.l .l .l .1 Circumferential and Longitudinal Welds, Items B1.11 and Bl .12, l All pressure-retaining circumferential and longitudinal shell 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 4 first inspection interval. Examinations may be performed at or near the end of the interval.

                                                                               ~

A.1.1. 2 Head Welds, Item Bl .20 f ) A.1.1. 2.1 Circumferential and Meridional Head Welds, Items B1.21 and 81.22 All pressure-retaining circumferential and meridional head wolds in the reactor vessel head shall be volumetrically examined in accordance with cc, 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 9, ) examined at or near the end of the interval. A.1.1. 3 Shell-to-Flange Weld, Item Bl .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 first inspection interval. If the examinations are conducted from the flange face, the remaining examination required to be conducted from the vessel wall may be performed at or near the end of each inspection interval. A.1.1. 4 Head-to-Flange Weld, Item B1.40 Essentially 100% of the length of the head-to-flange weld shall be volumetrically examined in accordance with Figure IV3-2500-5 during the first inspection interval. If the examinations are conducted from the l flange face, the remaining examination required to be conducted from the } vessel wall may be performed at or near the end of each inspection interval. l A-1

                                                                                  -   - - -          ---      w

A.1.l .5 Repair Welds, Item Bl .50 A.1.1. 5.1 Repair Welds in the Beltline Region, Item B1.51 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. The beltline region extends for the length of the vessel thernal shield, or in the absence of a thermal shield, the effective length of reactor fuel elements. If the location of the repair is not positively and accurately known, then the individual shell plate, forging, or shell-course containing the repair shall be included. A.1. 2 CATEGORY B-8, PRESSURE-RETAINING WELDS IN VESSELS OTHER TRAN REACTOR VESSELS 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 circunferential shell-to-head welds in the pressurizer shall be volumetrically examined in accordance with Figure IWB-2500-1 over - essentially 100% of their length during the first inspection interval. i A.1. 2.1. 2 Longitudinal Shell Weld, Item B2.12 One foot of 'se selected longitudinal shell weld in the pressurizer _ intersecting the examined circumferential shell-to-head weld shall be volumetrically examined in accordance with Figure IWB-2500-2 during the - first inspection interval . 4 A.1. 2. 2 Head Welds in Pressurizer Vessels, Item B2.20 l A.1. 2. 2.1 Circumferential and Meridional Head Welds, Items 82.21 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 All circumferential and meridional head welds in the primary side of the steam generators shall be volumetrically examined in accordance with Figure IWB-2500-3 over essentially 100% of their length during the first inspection interval . I A-2

A.1. 2. 4 Tubesheet-to-Head Weld, Item B2.40 The tubesheet-to-head weld 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, j Item B2.50 i A.1. 2. 5.1 Circumferential Welds, Item B2.51 i All circumferential shell (or head) welds in the primary side of the heat exchangers shall be volumetrically examined in accordance with Figures IWB-2500-1 and -3 over essentially 100% of their length during the first inspection interval . A.1.2.5.2 Longitudinal (or Meridional) Welds, Item B2.52 All longitudinal (or meridional) welds in the primary side of the heat 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) weld shall be volumetrically examined in accordance with Figure IWB-2500-6 over essentially 100% of its length during the first interval . ' u A.1. 3 CATEGORY B-D, FULL PENETRATION WELDS OF N0ZZLES IN VESSELS _ (INSPECTION PROGR#1 B) A.1.3.1 Reactor Vessel Nozzle-to-Vessel Welds, Items B3.90 and B3.100 All nozzle-to-vessel welds and inside radius sections in the reactor vessel 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 the inspection in terval . If examinations 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 inside diameter may be performed at or near the end of each inspection in terval . 4 l & A-3 i

i l A.1. 3. 2 Pressurizer Nozzle-to-Vessel Welds, Items B3.110 and B3.120 All nozzle-to-vessel welds and inside radius sections in the pressur- 1 izer shall be volumetrically examined in accordance with Figure IWS-2500-7 j during the first interval of operation. The nozzle-to-vessel weld and ad,jacent 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 . i A.1. 3. 3 Steam Generator Nozzle-to-Vessel Welds, Items 83.130 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-2500-7 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 the inspection interval. A. l . 3. 4 Heat Exchanger Nozzle-to-Vessel Welds, Items B3.150 and B3.160 All nozzle-to-vessel welds and 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 incl uded. 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. 4 A.1. 4 CATEGORY B-E, PRESSURE-RETAINING PARTIAL PENETRATION WELDS IN , VESSELS h-A.1.4.1 Reactor Yessel Partial Penetration Welds, Item B4.10 A.1. 4.1.1 Vessel Nozzles, Item B4.12 > The external surfaces of partial penetration welds on 25% of reactor 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. 1 ) A-4 l

i . A.1. 4.1. 2 Control Rod Drive Nozzles, Item B4.13 The external surfaces of partial penetration welds en 25% of the control rod drive nozzles shall be visually examined (VT-2) during the first inspection interval . The examinations shall cumulatively cover the specified percentage among each group of penetrations of comparable size and function. A. l . 4.1. 3 Instrumentation Nozzles, Item B4.14 $ The external surfaces of partial penetration welds on 25% of the instrumentation nozzles shall be visually examined (VT-2) during the first inspection interval. The examinations shall cumulatively cover the specified percentage among each group of penetrations of comparable size and function. 9 A.1. 4. 2 Heater Penetration Welds on the Pressurizer, Item B4.20 The external surfaces of 25% of the heater penetration welds on the pressurizer shall be visually examined (VT-2) during the first inspection . in terval . The examinations shall cumulatively cover the spect fled per-centage among each group of penetrations of comparable size and function.

                               ~

A.1. 5 CATEGORY B-F, PRESSilRE-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 j shall be surface and volumetrically examined in accordance with Figure '9 4 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 conbinations of - (a) carbon or low alloy steels to high alloy steels, (b) carbon or low alloy steels to high nickel alloys, and (c) high alloy steel to high nickel alloys are included. Examinations are required of each safe-end weld in 4 each loop and connecting branches of the reactor coolant-system. A.1.5.2 Pressurizer Nozzle-to-Safe End Welds, Item B5.20 p 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. 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. Examinations are required of each , ) safe-end weld in each loop and connecting branches of the reactor coolant system. , l I A-5

A.1.5.3 Steam Generator Nozzle-to-Safe End Welds, Item B5.30 All dissimilar metal nozzle-to-safe end welds in the steam generator l shall be surface and volumetrically examined in accordance with Figure l IWB-2500-8 during the first inspection interval. Dissimilar metal welds ! between continations 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. Examinations are required of each safe-end weld in each loop and connecting branches of the reactor coolant system. A.1.5.4 Heat Exchanger Nozzle-to-Safe Ends Welds, Item B5.40 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. 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. Examinations are required of each safe-end weld in each loop and connecting branches of the reactor coolant system.

                                                                                        ~

A.1.5.5 Piping Safe End Welds, Item B5.50 All dissimilar metal safe end welds in piping shall be surface and i volumetrically examined in accordance with Figure IWB-2500-8 during the first inspection interval . 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. Examinations are required of each safe-end weld in each loop - and connecting branches of the reactor coolant system. A.l .6 CATEGORY B-G-1, PRESSURE-RETAINING BOLTING LARGER THAN 2 INCHES IN DIAMETER A.1.6.1 Reactor Closure Head MJts, Item B6.10 The surfaces of all reactor closure head nuts larger than 2 in. in diameter shall be examined during the first inspection interval. Bol ting i may be examined (a) in place under 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 inspection interval. l

 )

4 A-6

l A.l.6.2 Reactor Closure Studs, Items B6.20 and B6.30 l All closure studs in the reactor vessel larger than 2 in. in diameter shall be volumetrically examined in accordance with Figure IWB-2500-12 during the first inspection interval. A surface examination is also required when T.he studs are removed. Examinations may be performed at or near the end of the inspection interval. A.1.6.3 Threads in the Flange in the Reactor Vessel, Item B6.40 All threads in the flange in the reactor vessel shall be volumetrically examined in accordance with Figure IWB-2500-12 during the first inspection 11terval . Examination includes threads in base metal and is required only w1en the annection is disassembled. Examinations may be performed at or n'ar the end of the inspection interval. A.I.6.4 Reactor Closure Washers and Bushings, Item B6.50 The surfaces of all closure washers and bushings on bolting larger than 2 in. in diameter in the reactor vessel shall be visually examined (VT-1) during the first inspection interval . The examinations are required only when the connection is disassembled and may be performed at or near the end of the inspection interval . Bushings may be examined in place. _ A.1.6.5 Pressurizer Bolts and Studs, Items B6.60 and B6.70 All bolts and studs larger than 2 in. in diameter in the pressurizer shall be volumetrically 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 a performed at or near the end of the inspection interval . _ A.1.6.6 Pressurizer Bolting, Item B6.80 The surfaces of all nuts, bushings, and washers on bolting larger than 2 in. in diameter in base material and threads in flange stud holes in the 4 pressurizer 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. Bushings and threads in base material of flanges are required to be examined only when the connections are disassembled. Bushings may be inspected in place. Examinations may be performed at or near the end of the inspection interval. A.1.6.7 Bolts and Studs, in Steam Generators,- Items 86.90 and B6.100 All bolts and studs larger than 2 in. in diameter in the steam generators shall be volumetrically examined in accordance with Figure IWB-1 2500-12 during the first inspection interval . A surface examination is A-7

i also required when the bolts and studs are removed. Examinations may be performed at or near the end of the inspection interval. A.1.6.8 Bolting in Steam Generators, Item B6.110 The surfaces of all nuts, bushings, and washers in bolting larger than 2 in, in diameter in base material and threads in flange stud holes in steam generators 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. Bushings and threads in base materials of flanges are required to be examined only when the connections are disassembled. Bushings may be inspected in place. Examinations may be performed at or near the end of the inspection interval . A.1.6.9 Bolts and Studs, in Heat Exchangers, Items 86.120 and B6.130 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 surface examination 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 ~ 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. A.1.6.10 Bol ting in Heat Exchangers, Item 86.140 The surfaces of all nuts, bushings, and washers on bolting larger than 2 in. in diameter in base material and threads in flange stud holes in heat exchangers shall be visually examined (VT-1) during the first inspection i in terval . Bolting may be examined (a) in place under tension, (b) when the connection is disassembled, or (c) when the bol ting is removed. Examina- _ tions may be performed at or near the end of the inspection interval. Bushings and threads 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 bol ts and studs on com-4 ponents selected for examination under Examination Categories B-B, B-J, B-L-1, and B-M-1, as applicable. A.1.6.11 Bolts and Studs in Piping, Items B6.150 and 86.160 All bolts and studs larger than 2 in. in diameter in piping shall be volumetrically examined in accordance with Figure IWB-2500-12 during the first inspection interval . A surface examination is also required when the bol ts and studs are removed. Examinations may be performed at or near the end of the inspection interval. A-8 e

A.1.6.12 Bolting in Piping, Item B6.170 The surfaces of all nuts, bushings, and washers on bolting larger than 2 in. in diameter in base material and threads in flange stud holes in piping shall be visually examined (VT-1) during the first inspection interval . Bolting nay be examined (a) in place under tension, (b) when the connection is disassembled, or (c) when the bolting is removed. Bushings and threads in base materials of flanges are required to be examined only when the connections are disassembled. Bushings may be inspected in place. Examinations may be performed at or near the end.of the inspection in terval . 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 volumetrically examined in accordance with Figure IWB-2500-12 during the first inspection interval . A :urface examination is also required when the bol ts and studs are removed. Examinations nay be performed at or near the end of the inspection interval. 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-1, as applicable. A.1.6.14 Bol ting in Pumos, Item B6.200 The surfaces of all nuts, bushings, and washers in bolting larger than 2 in, in diameter in base material and threads in flange stud holes in pumps shall be visually examined (VT-1) (a) in place under tenston, (b) when the connection is disassembled, or (c) when the bolting is removed. Examinations may be performed at or near the end of the inspection inter-val . Bushings and threads in base materials of flanges are required to be _ examined only when the connections are disassembled. Bushings may be Z inspected in place. Examinations are limited to bolts and studs on com- j pcnents selected for examination under Examination Categories B-8, B-J, B-L-1, and B-M-1, as applicable. A.1.6.15 Bolts and Studs in Valves, Items 86.210 and B6.220 All bolts and studs larger than 2 in. in diameter in valves shall be volumetrically 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 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 B44-1, as applicable. A.1.6.16 Bol ting in Valves, Item B6.230 The surfaces of all nuts, bushings, and washers in bolting larger than 2 in. in diameter in base materials and threads in flanged stud holes in valves shall be visually examined (VT-1) during the first inspection j A-9

interval . Bolting may be examined (a) in place under tension, (b) when the connection is disassembled, or (c) when the bolting is removed. Examina-tions nay be performed at or near the end of the inspection interval. Bushings and threads 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 com-ponents selected for examination under Examination Categories B-8, B-J, B-L-1, and B-M-1, as applicable. A.1.7 CATEGORY B-G-2, PRESSURE-RETAINING BOLTING 2 INCHES AND SMALLER IN DIAMETER A.1.7.1 Bolts , Studs , and Wts in Reactor Vessel , Item B7.10 The surfaces of all bolts, studs, and nuts 2 in. or less in diameter in the reactor vessel shall be visually examined (YT-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.1.7.2 Bolts, Studs, and Wts in Pressurizer, Item B7.20 The surfaces of all bolts, studs, and nuts 2 in. or less in diameter in the pressurizer shall be visually examined (YT-1) during the first in-spection interval . Bolting may be examined (a) in place under tension, (b) when the connection is disassembled, or (c) when the bolting is removed. A.1.7.3 Bolts, Studs , and Wts 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 (YT-1) during the first t 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.1.7.4 Bolts, Studs, and Wts in Heat Exchangers,' Item B7.40 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 bol ting is removed. $ A.1.7 . 5 Bolts, Studs, and Wts in Piping, Item B7.50 The surfaces of all bolts, studs, and nuts 2 in. or less in diameter in piping 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. l > A-10

A.1.7.6 Bolts, Studs, and MJts in Pumps, Item B7.60 The surfaces of all bol ts, studs, and nuts 2 in or less in diameter in pumps 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.1.7.7 Bolts, Studs, and MJts in Valves, Item B7.70 The surfaces of all bolts, studs, and nuts 2 in or less in 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, INTEGRAL ATTACmENTS FOR VESSELS A.1.8.1 Integrally Welded Attachments in Reactor Vessel, Item B8.10 The attachment weld joining the attachment to the pressure-retaining membrane of the reactor 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 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.

        /L1.8.2 Integrally Welded Attachments in Pressurizer, Item B8.20                      _

The attachment weld joining the attachment 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 3 applicable, in accordance wi th 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 applicable. One hundred percent of the welding of each lug on the pressurizer is included in the

 )

examination. A.1.8.3 Integrally Welded Attachments in Steam Generators, Item B8.30 The attachment weld joining the attachment 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 A-11 gw..

O s 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 steam generator and the integral attachment 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 the attachment weld on one steam generator. A.1.8.4 Integrally Welded Attachments in Heat Exchangers, Item 88.40 The attachment weld joining the attachment 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 heat 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 each lug on the heat exchanger is included in the examination. The examination is limited 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 A.1.9.1.1     Circumferential Welds, Item B9.11 For circumferential welds in pipe of nominal pipe size 4 in. and greater, surface plus volumetric examinations shall be performed in ac-          _.

cordance with Figure IWB-2500-8 over essentially 100% of the weld length during each inspection interval. The examinations shall include the following: i (a) All terminal ends in each pipe or branch run connected to vessels. ' l 1 i (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 ) primary plus secondary stress intensity of 2.4Sm for ferritic steel and austenitic steel, and i ( 2) cumulative usage factor U of 0.4. A-12

1 L i l t (c) All dissimilar metal welds between combinations of j (a) carbon or low alloy steels to high alloy steels; J (b) carbon or low alloy steels to high nickel alloys; and 1 (c) high alloys steels to high nickel alloys. l (d) Additional piping welds so that the total equals 25% of the cir- l cumferential joints in the reactor coolant piping system. This total does not include welds excluded by T4B-1220. These addi- 1 tional welds may be located in one loop (one loop is currently I defined for both PWR and BWR plants in the 1977 Edition). l 3 . A.l.9.1.2 Longitudinal Welds, Item B9.12 For longitudinal welds in pipe of nominal pipe size 4 in, and greater, surface plus volumetric examinations shall be performed in accordance with Figure FAB-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. l . 9. 2.1 Circumferential Welds, Item B9.21 For circumferential welds in pipe of nominal pipe size less than 4 in.,

surface examinations shall be performed in accordance with Figure T4B-2500-8 over essentially 100% of the weld length during each inspection in terval . 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 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. (c) All dissimilar metal welds between cortinations 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. (d) Additional 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 TAB-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). 9 A-13

o . 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 IWS-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. t A.1.9.3 Branch Pipe Connection Welds, Item B9.30 A.1.9. 3.1 !bminal 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 1007, 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 limits under loads associated with specific seismic events and operational conditions. 4 (1 ) primary plus secondary stress intensity of 2.4Sm' for ferritic steel and austenitic steel, and . ( 2) cumulative usage factor U of 0.4.

)

(c) All dissimilar metal welds between conbinations 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. j

)          (d) Additional piping welds so that the total equals 257, of the cir-           I 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).

) 1 ') A-14 l

e . A.1.9.3.2 Nominal Pipe Size Less Than or Equal to 2 In., Itcm B9.32 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. (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 ) primary'plus secondary stress intensity of 2.4Sm for ferritic steel and austenitic steel, and ( 2) cumulative usage factor U of 0.4. (c) All dissimilar metal welds between coabinations 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. (d) Additional 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 currently defined for both PWR and BWR plants in the 1977 Edition). A.l.9.4 Socket We lds , Item B9.40 Socket welds shall be surface examined in accordance with Figure . IWB-2500-8 over essentially 100% of the weld length during each insp?ction in terval . The examinations shall includa 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 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. > A-15

o- . (c) All dissimilar netal 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. (d) Additional piping welds so that the total equals 25% of the cir-cmnferential joints in the reactor coolant piping system. This total does not include welds excluded by IWB-1220. These addf-i tional welds may be located in one loop (one loop is currently defined for both PWR and BWR plants in the 1977 Edition). A.1.10 CATEGORY B-K-1, INTEGIAL ATTACR4ENTS FOR PIPING, Pil1PS, AfD VALVES A.l .10.1 Integrally Welded Attachments on Piping, Item B10.10 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 integrally welded 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. A.1.10.2 Integrally Welded Attachments on Pumps, Iten B10.20 Volumetric or surface examinations, as applicable, per Figures 4 IWB-2500-13, -14, and -15 of essentially 100% of the weld length are required for all welded attachments of pumps integral to 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. 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 essentially 100% of the weld length are i required for all welded attachments of valves integral to 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-16

f , A.1.11 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, PlNP CASINGS AND VALVE BODIES A.1.11.1 Pumo Casing Wel ds, Item B12.10 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 wi th Figure IWB-2500-16 during each inspection interval. The examinations may be performed at or near the end of the inspection interval. A.1.11.2 Pumo Casings, Item B12.20 The internal surfaces of at least one pump in each group of pumps per-forming similar functions in the system (e.g., recirculating coolant pumps) shall be visually examined (VT-1) during each 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 end of the inspection interval.

        , A.1.11.3 Valve Body Welds , Item B12. 30 Essentially 100% 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 j

Figure IWB-2500-17 during each inspection interval. The examinations may be performed at or near the end of the inspection interval. . A.1.11.4 Valve Body Exceeding 4 In. Nominal Pice Size, Item 812.40 The internal surfaces of 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 functions in the system (e.g., containment isolation and system overpressure protection) shall be visually examined ( VT-1) during each inspection interval. The examination my be t performed on the same valve selected for volumetric examination of welds. The examinations may be performed at or near the end of the inspection in terval . I l l - A-17 s

4 , A.1.12 CATE(DRIES B-N-1, INTERIOR OF REACTOR VESSEL; 8-N-2, INTEGRALLY WEl.DED CORE SUPPORT STRUCTURES AND INTERIOR ATTACffiENTS TO REACT YESSELS; and B-N-3, RDiOVABLE CORE SUPPORT STRUCTURES A.1.12.1 Reactor Vessel Interior, ' Item 813.10 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, sh.all be visually examined (VT-3) during the first refueling outage and subsequent refueling outages at approximately 3-year intervals. A.1.12.2 Boilina Water Reactor Vessel Interior Attachments , Item B13.20 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. A.1.12.2.1 Boiling Water Reactor Core Support Structure, Item 813.21 The accessible surfaces of the core support structure shall be visually examined (VT-1) during each inspection interval. The examinations may be performed at or near the end of the inspection interval. 3 A.l .12.3 Core Support Structure for Pressurized Water Reactor Vessels, - Item B13.30 The accessible welds and surfaces of the core support structure shall be visually examined (VT-3) each inspection interval. The structure shall The examinations may be removed from the reactor vessel for examination. be performed at or near the end of the inspecticn interval. A.1.13 CATEGORY B-0, PRESSURE-RETAINING WELDS IN CONTROL R0D HOUSINGS A.1.13.1 Welds in Control Rod 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 IW8-2500-18 during each inspection interval. The examinations may be performed at or near the end of the inspection interval. A-18

 - - . - - - ,           _  - - - , , . -           -                    , - - - - - -   -,     -,-,.n-,- -   m

b o A.1.14 CATEGORY B-P, ALL PRESSURE-RETAINING COMPONENTS A.1.14.1 Reactor Vessel Pressure-Retainino Boundary, Item B15.10 The reactor vessel pressure-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 wi th the exceptions specified in IWA-5214 when system pressure tests are conducted for repaired, replaced, or altered canponents. A.1.14.1.1 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 nay 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 for repaired, replaced, or altered components. A.1.14.2 Pressurizer Pressure-Retaining Boundary, Iten B15.20 The pressurizer pressure-retaining boundary shall be visually examined (VT-2) during the system leakage test perforned 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

componen ts.

A.1.14.2.1 P ressurizer P ressure-Retaining 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 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 4 al tered components. A-19

O O A.1.14.3 Steam Generator Pressure-Retaining Boundary, Item B15.30 The steam generator pressure-retaining boundary shall be visually examined ( VT-2) during t1e 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 al tered components. A.1.14.3.1 Steam Generator Pressure-Retaining Boundary, Item B15.31 The steam generator pressure-retaining boundary shall be visually examined (VT-2) during the system hydrostatic test performed in accordance with IWB-5222 during each re fueling outage. The examinations any 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 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 al tered componen ts. A.1.14.4 Heat Exchancer 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 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 al tered components. ) A.1.14.4.1 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 nay be per-formed at or near the end of the inspection interval. The entire i 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 reraired, replaced, or altered components. t A-20 i

A.1.14.5 Piping Pressure-Retainino Boundary, Item B15.50 The piping pressure-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.1.14.5.1 Piping Pressure-Retaining Boundary, Item B15.51 The piping 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 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.1.14.6 Pumo Pressure-Retaining Boundary, Item B15.60 The pump pressure-re taining 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 wi th the exceptions specified in IWA-5214 when system pressure tests are conducted for repaired, replaced, or altered componen ts. A.1.14.6.1 Pump Pressure-Retaining Boundary, Item B15.61 The pump pressure-re'taining boundary shall be visually examined ( VT-2) ' during the system hydrostatic test perforced in accordance with IWB-5222 -l

once per inspection interval. The examinations nay be performed at or near '

1 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 for repaired, replaced, or altered componen ts . A.1.14.7 Valve Pressure-Retaining Boundary, Item 815.70 The valve pressure-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 i reactor coolant system is subject to system pressure tests conducted in accordance with IWA-5000 with the exceptions specified in IWA-5214 when ! A-21 i l

     . system pressure tests are conducted for repaired, replaced, or altered componen ts.

A.1.14.7.1 Valve P ressure-Retaining Boundary, B15.71 The valve 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 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. 1 A.1.15 CATEGORY B-Q, STEAM GENERATOR TUBING A.1.15.1 Steam Generator Tubing, Straioht Tube Design, Item B16.10 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. A.1.15. 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 in 3% of the heating surface in each generator during the first inspection interval. t ) l 9 I I A-22 f .

A. 2 CLASS 2 REQUIREt1ENTS A. 2.1 CATEGORY C-A, PRESSURE-RETAINING WELDS IN PRESSURE VESSELS t 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 Figure IWC-2520-1 during each inspection interval. A gross structural discontinu-ity is defined in M3-3213.2. 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 mul tipl e vessels with similar design, size, and service (such as steam generators , and heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. A. 2.1. 2 Head Circumferential Weld, Item C1.20 Essentially 100% of the circumferential head-to-shell weld shall be volumetrically examined in accordance with Figure FAC-2520-1 during each inspection interval . For multiple vessels with similar design, size, and 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 Essentially 100% of the tubesheet-to-shell weld shall be volumetrically examined in accordance with Figure IWC-2520-2 during each inspection intervcl. For multiple vessels with similar design, size, and service 1 (such as steam generators and heat exchangers), the required examinations may be limited to one v9ssel or distributed among the vessels. A.2.2 CATEGORY C-B, PRESSURE-RETAINING N0ZZLE WELDS IN VESSELS A. 2. 2.1 Nozzles in Yessels 1/2 In. or Less in hbminal 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-ties of piping runs that connect to vessels. Only those piping runs selected for examination under Examination Category C-F are included. i A-23

A.2.2.2 Nozzles in Vessels Over 1/2 In. in Nominal Thickness, Item C2.20 All nozzles in vessels over 1/2 in. in nominal thickness at terminal ends of piping runs shall be surface and volumetrically examined in accor-dance with Figure IWC-2520-4 during each inspection interval. Terminal ends are the extremities of piping runs that connect to vessels. Only those piping runs selected for examination under Examination Category C-F are included. 1 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 One hundred percent of each integrally welded attachment in pressure vessels shall be surface examined in accordance with Figure IWC-2520-5 during each inspection interval. Examination is limited to integrally welded attachments whose base material design thickness is 3/4 in. or greater. 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 i of the attachments may be distributed among the vessels. A.2.3.2 Integrally Welded Attachments in Piping, Item C3.40 One hundred percent of each integrally welded attachment in piping shall be surface examined in accordance with Figure IWC-2520-5. Exami-nation is limited to integrally welded attachments whose base material design thickness is 3/4 in. or greater. In addition, examinations are limited to attachments of those components required to be examined under Examination Categories C-F and C-G. A.2.3.3 Integrally Welded Pumo Attachments, Item C3.70 One hundred percent of each integrally welded attachment in pumps shall be examined in accordance with Figure IWC-2520-5. Examination is limited to integrally welded attachments whose base material design thickness is 3/4 in, or greater. Examinations are limited to supports of those com- ' ponents required to be examined under Examination Categories C-F and C-G. l A.2.3.4 Integrally Welded Valve Attachments, Item C3.100 One hundred percent of each integrally welded valve support attachment shall be examined in accordance with Figure IWC-2520-5 during each i inspection interval. Examination is limited to integrally welded attachments whose base material design thickness is 3/4 in. or greater. Examinations are limited to supports of those components required to be examined under Examination Categories C-F and C-G. 1 A-24

i 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 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 service. In addition, where the component contains a group of bolted con-nections of similar design and size (such as flange connections and manway covers), only one bolted connection among the group need be examined. i Bolting may be examined in place under load or upon disassembly of the j 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 1 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. A.2.4.3 Bolts and Studs in Pumps, Item C4.30 For pumps,100% of the bolts and studs at each bolted connection of - pumps shall be volumetrically examined in accordance with Figure IWC- '_ 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 connections and manway covers), the examination may be conducted on one bolted connection among the group. Bolting may be examined in place under < load or upon disassembly of the connection. A.2.4.4 Bolts and Studs in Valves, Item C4.40

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 flange connections and manway covers), the examination may be conducted on one ,

bolted connection among the group. Bolting may be examined in place under l load or upon disassembly of the connection. l A-25

      - , - - - - - , -                                     , - . . , - - ,                -. -   - ,e e--..--,---r---       , - -

a o A. 2. 5 CATEGORY C-F, PRESSURE-RETAINING WELDS IN PIPING A. 2.5.1 Piping Welds 1/2 In. or Less Nominal Wall Thickness, Item C5.10 A. 2.5.1.1 Circumferential Welds, Item C5.11 One hundred percent of each circumferential weld 1/2 in, or less nominal wall thickness shall be surface examined in accordance with Figure , IWC-2520-7 during each inspection interval . The welds selected f]r examination shall include

a. all welds at locations where the stresses under the loadings 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. 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; i 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 thermal expansion; I

A-26

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.

s 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. fbminal 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 _

a. all welds at locations where the stresses under the loadings 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) belcw) of piping or branch runs;

c. all dissimilar metal welds;

d. additional welds, at structural discontinuities (see (f) below) such that the total nuser of welds selected for examination in-cludes the following percentages of circumferential piping welds.

For boiling water reactors:

1. none of the welds exenpted by P4C-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 P4C-1220; y 2. none of the welds in residual heat removal and emergency core cooling systems;

) , A-27

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 thermal expansion; structural discontinuities include pipe weld joints to vessel

                                                      ~

f. 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.2.5.2.2 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. - i 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 4 connections shall be examined in accordance with Figure IWC-2520-9 during each inspection interval. The welds selected for examination shall include _

a. all welds at locations where the stresses under the loadings 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)laf piping or branch runs;

c. all dissimilar metal welds; i

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;

)

4

         .                                     A-28

  . c                                        ,

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; 25% of the welds in all other systems.

4. For pressurized water reactors: .

1. none of the welds exemptad by IWC-1220;

2. none of the welds in res'idual heat removal and emergency core l 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. i

e. terminal ends are the extremitied of piping runs that connect to structures, components (such as, vessels, pumps, and valves) or i pipe anchors, each of which act as rigid restraints or provide at i least two degrees of restraint to piping thermal expansion-

f. structural discontinuities incluck pipe weld joints to vessel -

nozzles, valve bodies, pump casings, pipe fittings (such as, l elbows, tees, reducers, and flanges conforming to ANSI Standard B16.9), and nine branch connections and fittings; I

g. examination requirements are under' development.

                                                                                                                          .I 4        A. 2. 5. 3. 2 Longitudinal Welds, Item C5.32 Longitudinal welds in pipe branch connections shall be surface examined                                      -l in accordance with Figure IWC-2520-7,( 2.5 t:at the inteesecting circumfer-ential weld) during each inspection interval,.                                                                      i i

f 1 i i I i  ;

         -                                       A-29 J

l 1 A.2.6 CATEGORY C-G, PRESSURE-RETAINING WELDS IN PUMPS AND VALVES A.2.6.1 Pump 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. r A. 2. 6. 2 Valve Body Welds, Item C6.20 One hundred percent of all valve body welds in each pi;,ing 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 1 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 Vessels, 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'exe.nptions or exclusions from these requirements except ' as specified in IWA-5214. A. 2.7.1.1 Pressure Vessels, Item C7.11 Pressure vessel pressure-retaining boundaries (other than open-ended portions of systems) shall be visually examined (YT-2) during the system hydrostatic test performed in accordance with IWC-5222 during each inspec-tion period. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. The system byirostatic test shall be conducted at or near the end of each inspeGia interval or during the same inspection period of each inspection intmal -f Inspection Program B. A. 2.7. 2 Piping, Item C7.20 Piping 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-30

,             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 l period of each inspection interval of Inspection Program B. l i l A.2.7.3 Pumps, Item C7.30 1 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 4

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 l 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

                                                                               /

r 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 specified in IWA-5214. A system functional

 ;            test (IWC-5221) serves as a required system pressure test.

s i , 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. f . A-31 _ - - - _ , . _ _ . _ - ,_e _ . - -_ ,_ _ _ . - , , . _ . _ , -, . ; - , , . ,

A. 3 CLASS 3 REQUIRBIENTS A. 3.1 CATEGORY D-A, SYSTEMS IN SUPPORT OF REACTOR SHUTDOWN FUNCTION A. 3.1.1 Pressure-Retaining Components, Item D1.10 Pressure-retaining components in the pressure-retaining boundary shall be visually examined (VT-2) during the system pressure test (IWA-5000/ IWD-5221) or system hydrostatic test (IWD-5223) each inspection period. The system hydrostatic test shall be conducted at or near the end of each inspection interval or during the samc inspection period of each inspection interval for Inspection Program B. The system boundary extends up to and includes the first normally closed valve or valve capable of automatic closure as required to perform the safety-related system function. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. . A. 3.1. 2 Integral Attachment--Component Supports and Restraints, Item D1.20 Component supports and restraints shall be visually examined (VT-3) in ' accordance with Figure IWD-2500-1 during each inspection interval . Includes attachments for components exceeding 4 in. nominal pipe size whose structural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A 3.1.3 Integral Attachment--Mechanical and Hydraulic Snubbers, ) Item D1.30 , Mechanical and hydraulic snubbers shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in. nominal pipe size whose struc-tural integrity is relied on to withstand design loads when the system ) function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A. 3.1. 4 Integral Attachment--Spring Type Supports , Item D1.40 l 0 Spring type supports shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attach- l ments for components exceeding 4 in. nominal pipe size whose structural l integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements ) except as specified in IWA-5214. ,) A-32

G . A. 3.1.5 Integral Attachment--Constant Load Type Supports, Item 01.50 Constant load type supports shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in. nominal pipe size whose structural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A. 3.1.6 Integral Attachment--Shock Absorbers, Item 01.60 Shock absorbers shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in. nominal pipe size whose structural integrity is relied on to withstand design loads when the system function is re-quired. There are no exemptions or exclusions from these requirements except as speci fied in IWA-5214. A.3.2 CATEGORY D-B, SYSTEMS IN SUPPORT OF EMERENCY CORE COOLING, CONTAIMiENT HEAT REMOYAL, ATMOSPHERIC CLEANUP, AND REACTOR RESIDUAL HEAT REMOYAL A.3.2.1 Pressure-Retaining Components, Item D2.10 The pressure-retaining components in the pressure-retaining boundary shall be visually examined (VT-2) during the system pressure test (IWA- . 5000/IWD-5221) or system hydrostatic test (IWD-5223) each inspection s period. The system hydrostatic test shall be conducted at or near the end L of each inspection interval or during the same inspection period of each inspection interval for Inspection Program B. The system boundary extends - up to and includes the first normally closed valve or valve capable of automatic closure as required to perform the safety-related system function. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A.3.2.2 Integral Attachment--Component Supports and Restraints, Item D2. 20 Component supports and restrain'ts shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in nominal pipe size whose struc-tural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. i l A-33

A.3.2.3 Integral Attachment--Mechanical and Hydraulic Snubbers, ftem D2.30 Mechanical and hydraulic snubbers shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in. nominal pipe size whose struc-tural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A.3.2.4 Integral Attachment--Spring Type Supports, Item D2.40 Spring type supports shall be visually examined (W-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attach-ments for components exceeding 4 in. nominal pipe size whose structural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A.3.2.5 Integral Attachment--Constant Load Type Supports, Item D2.50 Constant load type supports shall be visually examined (VT-3) in ac-cordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in. nominal pipe size whose struc- _ tural int.egrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. - A.3.2.6 Integral Attachment--Shock Absorbers, Item D2.60

                                                                                           ~

Shock absorbers shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments f for components exceeding 4 in. nominal pipe size whose structural integrity l is relied on to withstand design loads when the system function is _ required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A.3.3 CATEGORY D-C, SYSTEMS IN SUPPORT OF RESIDUAL HEAT REMOYAL FROM SPENT FUEL STORAGE POOL A. 3. 3.1 Pressure-Retaining Components, Item D3.10 l The pressure-retaining components in the pressure-retaining boundary  ; shall be visually examined (W-2) during the system pressure test (IWA-5000/ IWD-5221) or system hydrostatic test (IWD-5223) each inspection period. 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 for Inspection Program B. The system boundary extends up to and i A-34 [ ,

includes the first normally closed valve or valve capable of automatic closure as required to perform the safety-related system function. There are no exemptions or exclusions from these requirements except as spect fied ) in IWA-5214.  ; 1 A.3.3.2 Integral Attachment--Component Supports and Restraints, Item D3.20 Component supports and restraints shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in. nominal pipe size whose struc-tural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A.3.3.3 Integral Attachment--Mechanical and Hydraulic Snubbers, Item D3.30 ) Mechanical and hydraulic snubbers shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in, nominal pipe size whose struc-tural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. A.3.3.4 Integral Attachment--Spring Type Supports, Item D3.40 Spring type supports shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attach-ments for components exceeding 4 in. nominal pipe size whose structural - integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements '. except as specified in IWA-5214. ~4 ) . A.3.3.5 Integral Attachment--Constant Load Type Supports, Item D3.50 -- Constant load type supports shall be vis'ually examined (VT-3) in accor-dance with Figure IWD-2500-1 during each inspection interval. Includes y attachments for components exceeding 4 in, nominal pipe size whose struc-tural integrity is relied on to withstand design loads when the system function is required. There are no exemptions or exclusions from these requirements except as specified in IWA-5214. ) A.3.3.6 Integral Attachment--Shock Absorbers, Item D3.60 Shock absorbers shall be visually examined (VT-3) in accordance with Figure IWD-2500-1 during each inspection interval. Includes attachments for components exceeding 4 in, nominal pipe size whose structural integrity is relied on to withstand design loads when the system function is ) required. There are no exemp*. ions or exclusions from these requirements except as speci fied in IWA-5214. A-35 p

A. 4 COMP 0NENT SUPPORTS A. 4.1 CATEGORY F-A, PLATE AND SHELL TYPE SUPPORTS, ITEMS F-1, F-2, AND F-3 All supports within the examination boundaries of IWF-1300 having canponents and piping required to be examined during the first inspection interval by IWB-2500, IWC-2500, and IWD-2500 shall be visually examined (VT-3) each inspection interval. The areas subject to examination are mechanical connections to the pressure-retaining component; weld and mechanical connections to the building structure; and weld and mechanical connections at intermediate joints in a multiconnected integral and non-integral support. i A.4.2 CATEGORY F-B, LINEAR TYPE SUPPORTS, ITEMS F-1, F-2, AND F-3 All supports within the examination boundaries of IWF-1300 having components and piping required to be examined during the first inspection interval by IWS-2500, E4C-2500, and IWD-2500 shall be visually examined (VT-3) each inspection interval . The areas subject to examination are mechanical connections to the pressure-retaining component; weld and mechanical connections to the building structure; and weld and mechanical connections at intermediate joints in a mul ticonnected ' integral and non-integral support. 4: A.4.3 CATEGORY F-C, COMPONENT STANDARD SUPPORTS, ITEMS F-1, F-2, F-3, AND F-4 _ All supports within the examination boundaries of D4F-1300 having components and piping required to be examined during the first inspection interval by D4B-2500, IWC-2500, and IWD-2500 shall be visually examined (VT-4) each inspection interval. The areas subject to examination are mechanical connections to the pressure-retaining component; weld and mechanical connections to the building structure; weld and mechanical connections at intermediate joints in multiconnected integral and non-integral support; and spring-type supports, constant load-type supports, snubbers, and shock absorbers. A-36

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