ML20246P967
| ML20246P967 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/31/1988 |
| From: | Beth Brown, Mudlin J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20246P927 | List: |
| References | |
| CON-FIN-D-6022 EGG-ESM-7914, NUDOCS 8905220320 | |
| Download: ML20246P967 (31) | |
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e sp wa r-EGG-ESM-7914 January 1988 I
INFORMAL REPORT I
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TECHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN: CAROLINA POWER AND LIGHT COMPANY, BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1.
DOCKET NUMBER 50-325 B. W. Brown J. D. Mudlin a
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TECHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
CAROLINA POWER AND LIGHT COMPANY, F
BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1, DOCKET NUMBER 50-325 B. W. Brown J. D. Mudlin Published January 1988 Idaho National Engineering Laboratory EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for:
U.S. Nuclear Regulatory Comission Washington, D.C. 20555 under DOE Contract No. DE-AC07-761001570 FIN No. 06022 (Project 5)
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t ABSTRACT This report presents the results of the evaluation of the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval Inservice Inspection (ISI)
Program Plan, through Revision 1, submitted August 19, 1987, including the requests for relief from the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code Section XI requirements which the Licensee has determined to be impractical. The Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous PSI and ISI reviews. The requests for relief from the ASME Code requirements which the Licensee has determined to be impractical for the second 10-year inspection interval are evaluated in Section 3 of this report.
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-l This work was funded under:
U.S. Nuclear Regulatory Commission FIN No. D6022, Project 5 Operating Reactor Licensing Issues Program, j
Review of ISI for ASME Code Class 1, 2, and 3 Components 11
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SU N RY The Licensee, Carolina Power and Light Company, has prepared the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval Inservice Inspection (ISI) Program Plan, through Revision 1, to meet the requirements of the 1980 Edition, Winter 1981 Addenda (80W81) of the ASME Code Section XI except that the extent of examination for Code Class 2. piping welds has been determined by ASME Code Case N-408. The second 10-year interval began July 10, 1986 and ends July 10, 1996.
The information in the BrunswicP. Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan, Revision 0, submitted November 10, 1986, was reviewed, including the requests for relief from the ASME Code Section XI requirements which the Licensee has determined to be impractical. As a result of this review, a Request for Additional Information (RAI) was prepared describing the information and/or clarification required from the Licensee in order to complete the review.
Sevision 1 of the ISI Program Plan was included in the Licensee's responses to the NRC's RAI.
Based on the review of the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan, through Revision 1, the Licensee's responses to the NRC's RAI, and the recommendations for granting relief from the ISI examination requirements that have been determined to be impractical, it has been concluded that the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan, through Revision 1, with the exception of Request for Relief RR-8 which is not evaluated.in this report, is acceptable and in compliance with 10 CFR 50.55a(g)(4).
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6 CONTENTS ABSTRACT.................................................................
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SUMMARY
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INTRODUCTION..........................................................
1 2.
EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN.......................
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2.1 Docume n ts Eval u at ed................................................ 3 2.2 Compliance with Code Requirements.................................. 3 2.2.1 Compliance with Applicable Code Editions.......................
3 2.2.2 Acceptability of the Examination Sampl e........................ 4 2.2.3 Excl u s i on C ri t e ri a...................,......................... 4 2.2.4 Augmented Examination Commitments..............................
4 2.3 Conclusions........................................................5 3.
EVALUATION OF RELIEF REQUESTS.........................................
6 3.1 Class 1 Components.................................................
6 3.1.1 Reactor Pressure Vessel........................................
6 3.1.1.1 Request for Relief No. RR-2, Revision 1, Examination Category B-A, Item B1.11, Reactor Pressure Vessel C i rcumfe ren t i al Shell Wel d................................. 6 3.1.2 Pressurizer (Does not apply to BWRs) 3.1.3 Heat Exchangers and Steam Generators (No relief requests) 3.1.4 Piping Pressure Boundary.......................................
8 3.1.4.1 Request for Relief No. RR-1, Revision 0, Examination I
Category B-J, Item B9.11, Pressure Retaining Class 1 Circumferential Piping Welds in Containment Penetrations... 8 3.1.5 Pump Pressure Boundary...................................<.....
9 3.1.5.1 Request for Relief No. RR-4, Revision 0, Examination Category B-L-2, Item B12.20, Internal Surfaces of Class 1 Pump Casings....................................... 9 l
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' 3.1. 6. Val ve Pres s ure Bounda ry....................................... 12 3.1.6.1 Request for Relief No. RR-3, Revision 0,. Examination Category B-M-2, Item B12.50, Internal Surfaces of Cl a s s 1 Val ve Bod i e s...................................... 12 l
3.1.7 General (No relief requests) l 3.2 Class 2 Comporents (No relief requests) 3.3 Class 3 Components (No relief requests) 3.4 Pressure Tests.....................................................
14 3.4.1 Class 1 System Pressure Tests (No relief requests) 3.4.2 Cl as s 2 Sys tem Pres sure Test s................................. 14 3.4.2.1, Request for Relief No. RR-7, Revision 0, Hydrostatic Test of the Class 2 Portion of the Main Steam System......14 3.4.3 Cl ass 3 System Pressure Tests................................. 16 3.4.3.1 Request for Relief No. RR-6, Revision 0, Hydrostatic Test of the Class 3 High Pressure Portion of the High Pressure Coolant Injection System.........................
16 3.4.4 General (No relief requests) 3.5 General...........................................................17 3.5.1 Ultrasonic Examination Techniques (No relief requests) 3.5.2 Exempted Components (No relief requests) 3.5.3 0ther.........................................................
18 3.5.3.1 Request for Relief No. RR-5, Revision 0, Component Supports on the Service Water Piping for the Four Diesel Fuel Storage Chambers.............................. 18 3.5.3.2 Request for Relief No. RR-8, Revision 0, Functional Te sting of Snubbers....................................... 20 4.
CONCLUSION...........................................................
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REFERENCES...........................................................
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TECHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
CAROLINA POWER AND LIGHT COMPANY,'
BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1, DOCKET NUMBER 50-325 1.
INTRODUCTION l
Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including supports) which are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vassel Code Class 1, Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," (Reference 2) to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during the second 120-month inspection l
interval shall comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the second 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code which are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein. The Licensee, Carolina Power and Light Company, has prepared the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval Inservice Inspection (ISI) Program Plan, through Revision 1, to meet the requirements of the 1980 Edition, Winter 1981 Addenda (80W81) of the ASME Code Section XI except that the extent of examination for Class 2 piping welds has been determined by ASME Code Case N-408. The second 10-year interval began July 10, 1986 and ends July 10, 1996.
As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain Code examination requirements are impractical and requests relief from them, the licensee shall submit information and justifications to the Nuclear Regulatory Commission (NRC) to support that determination.
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'. 4 Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's -
determinations under 10 CFR 50.55a(g)(5) that Code requirements are impractical. The NRC may grant relief and may impose' alternative l
l life or property or the common defense and security, and are otherwise in
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requirements that it determines are authorized by law, will not endanger the public interest, giving due consideration to the burden upon the
- licensee that could result if the requirements were imposed on the facility.
The information in the Brunswick Steam Electric Plant, Unit 1, Second i
10-Year Interval ISI Program Plan, Revision 0, submitted November 10, 1986 (Reference 3), was reviewed, including the requests for relief from the ASME Code Section XI requirements which the Licensee has determined to be impractical. The review of the ISI Program Plan was performed using the Standard Review Plans of NUREG-0800 (Reference 4), Section 5.2.4, " Reactor Coolant Boundary Inservice Inspections and Testing," and Section 6.6,
" Inservice Inspection of Class 2 and 3 Components."
In a letter dated April 3, 1987 (Reference 5), the NRC requested additional information that was required in order to complete the review of the ISI Program Plan. The requested information was provided by the Licensee in-letters dated June 4, 1987 (R % rence 6) and August 19, 1987 (Reference 7).
In the August 19, 1987 response, the Licensee added three relief requests and submitted Revision 1 of the ISI Program Pla1.
The Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan, through Revision 1, is evaluated in Section 2 of this report.
The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous PSI and ISI reviews.
The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI,
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I 1980 Edition including Addenda through Winter 1981. Specific inservice test (IST) programs for pumps and valves are being evaluated in other reports.
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2.
EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN t
This evaluation consisted of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements-and any license conditions pertinent to ISI activities. This section describec the submittals reviewed and the results of the review.
2.1 Documents Evaluated Review has been completed on the following information provided by the Licensee:
(a) Letter, dated November 10, 1986, submittal of Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan,.
Revision 0; (b) Letter, dated June 4, 1987, Licensee's response to the NRC's RAls;<
(c) Letter, dated August 19, 1987, additional information'with regard to the NRC's RAI and submittal of Revision 1 of the ISI Program Plan.
2.2 Comoliance with Code Requirements 2.2.1 Comoliance with Anolicable Code Editions The Inservice Inspection Program Plan shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of July 10, 1986 for the second 10-year interval, the Code applicable to the second 10-year inspection interval ISI program plan is the 1980 Edition with Addenda through Winter 1981. As stated in Section 1 of this ;eport, the Licenses has written the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan, Revision 1, to meet the requirements of the 1980 Edition, Winter 1981 Addenda of the Code except that the extent of examination for Code Class 2 piping welds has been determined by ASME Code Case N-408, " Alternative Rules for 3
C Examination of Class 2 Piping,Section XI, Division 1."
Code Case N-408 is referenced in Regulatory Guide 1.147, Revision 5 (Reference 8), as an NRC-approved code case and, therefore, may be used.
2.2.2 Acceptability of the Examination Samole Inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1, 2, and 3 components and their supports using sampling schedules described in Section XI of the ASME Code and 10 CFR 50.55a(b). Sample size and weld selection have been implemented in accordance with the Code and appear to be correct.
2.2.3 Exclusion Criteria The criteria used to exclude components from examination shzil be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, IWD-1220, and 10 CFR 50.55a(b). The exclusion criteria have been applied by the Licensee in acc.ordance with the Code as discussed in the ISI Program Plan and appear to be correct.
2.2.4 Auamented Examination Commitments In addition to the requirements as specified in Section XI of the ASME Code, the Licensee has comitted to meet the inspection requirements contained in the following documents:
(a) NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary P.iping," Revision 1
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(Reference 9);
(b) NUREG 0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line j
Nozzle Cracking" (Reference 10),
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IEB 80-13, Cracking in Core Spray Spargers (Reference 11);
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(d) NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping" (Reference 12); and (e) Regulatory Guide 1.150, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," Revision 1 (Reference 13).
2.3 Conclusions i
Based on the review of the documents listed above, it is concluded th.* the-Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan, through Revision 1, is acceptable and in compliance with 10 CFR 50.55a(g)(4).
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EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements which the Lic ensee has determined to be impractical for the second 10-year inspection interval are evaluated in the following sections.
3.1 Class 1 Comoonents 3.1.1 Reactor Pressure Vessel 3.1.1.1 Reouest for Relief No. PR-2. Revision 1. Examination Cateoory B-A. Item B1.11. Reactor Pressure Vessel Circumferential Shell Mald Code Requirement'.Section XI, Table IWB-2500-1, Examination Category B-A, Item B1.11 requires a 100% volumetric examination of one reactor pressure vessel circumferential shell weld in the beltline region as defined by Figure IWB-2500-1.
Licensee's Code Relief Reouest:
Relief is requested from examining 100% of the Code-required volume of RPV circumferential shell weld DC.
t Licensee's Procosed Alternative Examination: The Licensee states that Section XI permits this particular examination to be deferred to the end of the inspection interval. The Licensee will keep abreast of any technological advancements that could increase the examination coverage. This examination will be deferred until the last inspection period.
Based upon the technology available at that time, welds will be examined to the extent feasible, In addition, the remaining beltline longitudinal seam weld (E38) and the accessible portions of the other beltline circumferential weld (DB) will be examined to the extent practical. This will result in additional examination of approximately 373 inches of beltline region vessel weld.
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i J,1censee's Basis for Reouestino Relief: The
. ee states that the Brunswick plant examination accessibility was designed in accordance with the requirements of the 1974 Edition through the Summer 1975 Addenda of Section XI by the installation of permanent tracks to be used by automated equipment. This particular edition and addenda of Section XI required examination of 5% of the weld length of circumferential welds and 10% of the weld length of longitudinal seam welds. The new Code requirements represent a significant change in examination coverage.
The Licensee also states that space limitations between the RPV and the reflective mirror insulation do not provide the necessary clearance for util' ation of tracklest automated equipment such as magnetic cre.wlers. Removal of the reflective mirror insulation and permanent tracks is impractical due to the high radiation fields (5-15 R/hr) between the RPV and biological shield.
Insulation removal (including tracks) would require in excess of 300 man-hours with an estimated 1500-4500 R exposure.
Approximately 299 inches of the 751 inches of circumferential beltline region weld DC is examinable. Based on estimated accessibilities, an additional 373 inches of other beltline region circumferential and longitudinal weld lengths can be examined.
Evaluation: Based on the RPV design, the volumetric examination of beltline region weld DC, to the extent required by the Code, is impractical. An acceptable percentage of the Code-required examination will be performed on circumferential beltline region weld DC when the remaining beltline longitudinal seam weld (E38) and the accessible portions of the other beltline circumferential weld (DB) are substituted for the remainder. The RPV would have to be redesigned and prefabricated in order to complete the volumetric craminations as required by the Code.
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Conclusions:==
Based on thel above evaluation, it is concluded that the proposed alternative examination ensures an acceptable level of inservice structural integrity and that compliance with the' specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in.the level of quality.and safety. Therefore,.it is recommended that relief be granted as requested.
3.1.2 Pressurizer (Does 'not apply to BWRs) c 3.1.3 Heat Exchanaers and Steam Generators (No relief requests) 3.1.4 Pioina Pressure Boun'darv 3.1.4.1 Reauest for Relief No. RR-1. r.evision 0. Examination Cateaorv S-J. Item B9.11. Pressure' r tainina Class 1 Circumferential e
Pioina Welds in Containment Penetrations Code Requirement: Section XI, Table IWB-2500-1, Examination-Category B-J, Item 89.11 requires both 100% volumetric and surface examinations of the circumferential welds in' Class 1 piping 4 inches and greater as defined by Figure IWB-2500-8.
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Licensee's Code Relief Reauest: Relief is requested from performing the Code-required surface and volumetric examinations of the Class I circumferential piping welds within o
the following Type A penetrations:
X-7A, B, C, O X-12 X-8 X-13A, B X-9A, B X-16A, B X-10 X-17 X-11 Licensee's Proposed Alternative Examination: The Licensee states that the first circumferential pressure retaining weld outside containment on each of the subject penetrations will be examined during the inspection interval.
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nr Licensee's Basis for Reauestina Relief: The Licensee: states-that the design'of the Brunswick plant was finalized befora the access requirements of Section XI were published. These p
penetrations-were fabricated in such a manner that there is no--
access to these welds to perform the required examinations.
The number of welds affected represents only a small number of' the total number of Category B-J welds'(15 of approximately-
'l 470).. Adequate as:urance of the weld: integrity is verified 1 during the VT-2 examination performed on the affected piping systems each refueling outage ~durihg system leakage or hydrostatic pressure tests, q
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~ Evaluation: The sketch included in the relief request ~shows
' that the' subject welds are located inside containment
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. penetrations and, therefore, are completely inaccessible for the surface and volumetric examinations. The combination of.
the volumetric and surface examinatie of the first I
circumferential pressure retaining weld outside containment on~
J each of the suoject penetrations and the VT-2: examination for leakage during the system leakage or hydrostatic pressure tests ensure an acceptable level of inservice structural integrity.
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Conclusions:==
Based on' the above evaluation, it is concluded that compliance with'the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Therafore, it is recommended that relief be granted as requested.
.3.1.5 pumo Pressure Boundary 3.1.5.1 Reauest for Relief No. RR-4. Revision 0. Examination Cateaory B-L-2. Item B12.20. Internal Surfaces of Class 1 Pumo Casinos Code Requirement: Section XI Table IWB-2500-1, Examination Category B-L-2, Item B12.20 (pump casings) requires a visual 9
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examination (VT-3) of the internal surfaces of at least one pump in each group of pumps performing similar functions in the system. This examination may be performed on the same pump selected for volumetric examination of welds. This examination may be performed at the end of the 10-year interval.
Licensee's Code Relief Recuest: The Licensee requests relief from performing the Code-required visual examination (VT-3) of internal surfaces of the pump casings of the two reactor coolant recirculation pumps in each of Brunswick Units 1 and 2.
Licensee's Proposed Alternative Examination: None. The Ltcensee states that, whenever a recirculation pump is disassembled for maintenance, an examination of the internal pressure bound; y surfacec will be performed.
Licensee's Basis for Recuestina Relief: The Lf.censee states that the basis for this relief request is predicat9d on the following two points:
(a) To ccmplete the subject examination, large expenditures of I
man-hours and man-rem uposure are required with essentially no compensating increase in plant safety; and i
(b) The structural integrity afforded by the pump casing material will not significantly degrade over the lifetime of the pump.
Based on data cr'npiled from an actual recirculation pump
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disassembly, it is expected that approximately 1000 man-hours and 50 man-rem exposure would be required to disassemble, inspect, and reassemble one pump. Performing this visual examination under adverse conditions such as high dose rate (30-40 R/hr) and poor as-cast surface condition, realistically, provides little additional information as to the pump casing integrity.
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The recirculation pump casing material, cast stainless steel (ASTM A351-CF-8), is widely used in the nuclear industry and has performed extremely well. The presence of 'some delta ferrite (typically 5% or more) imparts substantially increased resistance to intergranular stress corrosion cracking. The delta ferrite also results in improved pitting corrosion resictance in chloride containing environments.
In addition, the Licensee reports that the Reactor Recirculation System is subjected to a VT-2 examir.ation whenever the drywell is deinerted in accordance with Generic Letter 84-11 (Reference 14).
Evaluation: The visual examination'is to determine whether unanticipated severe degradation of the ctsing is occurring due to phenomena such as eresion, corrosion, or cracking. However, previous experience during examination of pumps at other plants 1
has not shown any significant degradation of pump casings. The concept of visual examination when the pump is disassembled for maintenance is acceptable. The disassembly of the pumps for the sole purpose of inspection is a major effort and, in addition to the possibility of additional wear or damage to the internal surfaces of the pumps, could result in personnel receiving large amounts of radiation exposure. However, if the pumps are disassembled for maintenance, the internal surfaces would be examined, in which case relief would not be required for those particular pumps.
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Conclusions:==
Based on the above evaluation, it is concluded that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Therefore, it is recommended that:
(1) The Licensee's proposal to perform the visual examination (VT-3) of the internal surfaces of the pumps, whenever they are made accessible due to disassembly for maintenance, should be accepted; and (2) Relief 11
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should be granted at the end of the interval if one of the subject pumps, for which a visual examination is required, has not been disassembled for maintenance.
3.1.6 y_alve Pressure Boundary
~ 3.1.6.1. Raouest for Relief No. pR-3. Revision 0. Examination Cateaory B-M-2. Item B12.50. Internal Surfaces of Class 1 Valve Bodies Code Pecuirement: Section XI, Table IWB-2500-1, Examination Category B-M-2, Item B12.50 (Valve Body, Exceeding 4 in.
Nominal Pipe size) requires a visual examination (VT-3) of the internal surfaces of valve bodies. The examinations are limited to one valve within each group of valves that are of the same constructional design, such as globe, gate, or check valve, and manufacturing method, and that perform similar:
functions in the system, such as containment isolation and rystem overpressure protection. The examination may be performed on the same valve selected for volumetric examination. This examination may be performed at the end of the 10-year interv:1.
Licensee's Code Relief Reauest: Relief is requested from performing the Code-required visual (VT-3) examination of the 56 Class 1 valves greater than 4-inch nominal pipe size that are listed in the relief request.
Licensee's Prooosed Alternative Examination: None. The Licensee states that an examination of the internal pressure boundary surfaces will be performed, to the extent practical, each time a valve is disassembled for maintenance.
Licensee's Basis for Reauestina Relief: The Licensee states that the requirement to disassemble primary system valves for l
the sole purpose of performing a visual examination of the internal pressure boundary surfaces has only a very small 12 i
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potential of increasing plant' safety margins and a very disproportionate impact on expenditures of plant manpower and radiation exposure.
The Licensee also states that performing these visual examinations, under adverse conditions such as high dose rates (10 R/hr) and poor as-cast surface condition, realistically, provides little additional information as to the valve casing integrity. For approximately 20% of these valves, the reactor vessel core must be completely unloaded and the vessel drained to permit disassembly for examination.
A more practical approach that would essentially provide an equivalent sampling program and significantly reduced radiation exposure to plant personnel is to inspect the internal pressure boundary of only those valves that require disassembly for maintenance. This would still provide a reasonable sampling of primary system valves and give adequate assurance that the integrity of these components is being maintained.
Evaluation: The visual examination is to determine whether unanticipated severe degradation of the valve body is occurring due to phenomena such as erosion, corrosion, or cracking.
t However, previous experience during examination of valves at other plants has not shown any significant degradation of valve bodies. The concept of visual examination if the valve" is disassembled for maintenance is acceptable. The disassembly of the valves for the' sole purpose of inspection is a major effort and, in addition to the possibility of additional wear or damage to the internal surfaces of the valves, could result in personnel receiving large mornts of radiation exposure.
However, if the valves are disassembled for maintenance, the internal surfaces would be examined, in which case relief would not be required for those particular valves.
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Conclusions:==
Based on the above evaluation, it is concluded that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Therefore, it is recommended that:
(1) The Licensee's proposal to perform the visual examination (VT-3) of the internal.
surfaces of the valves, whenever they are made accessible due to disassembly for maintenance, should be accepted; and (2)
Relief should be granted at the en'd of the interval if one of the subject valves, for which a visual examination is required, has not been disassembled for maintenance.
3.1.7 General (No relief requests) 3.2 Class 2 Components (No relief requests) 3.3 Class 3 Components (No relief requests) 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests (No relief requests) 3.4.2 Class 2 System Pressure Tests 3.4.2.1 Reauest for Relief No. RR-7. Revision 0. Hydrostatic Test of the Class 2 Portion of the Main Steam System Code Requirement: SectionXI,SubarticleIWC-5222(a) requires a hydrostatic pressure test of this Class 2 portion of the Main Steam System at a pressure of 1105 psig.
Licensee's Code Relief Recuest: Relief is requested from performing the Code-required hydrostatic pressure test of the Class 2 portion of the Main Steam System at the test pressure of 1105 psig.
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Licensee's Pronosed Alternative Examination: The Licensee states that the MSIVs will be left open during the Class I hydrostatic pressure test to facilitate both Class 1 and 2 Main Steam to be subject to a combined VT-2 examination at 1088 psig.
Licensee's Basis for Recuestina Relief: The Licensee states that the Main Steam Isolation Valves (MSIVs) are the only isolation valves between the Class 1 and 2 portions of this system. These valves are unidirectional, and normally see pressure from the reactor side. Performance of the hydrostatic pressure test would result in 1105 psig of pressure being placed against the outboard valve seats, which are not designed for this pressure being applied from the opposite direction.
Gagging these valves would result in potential stem damage.
Evaluation: Because the system's design does not permit pressurizing the sections of piping to the Code-required pressure from the Class 2 piping side, the performance of the Code-required hydrostatic pressure test from the Class 2 piping side is impractical. The required visual examination of the piping during the alternative pressure test as well as the other required NDE of the welds in the system will provide reasonable assurance of the integrity of the piping. The difference in the required test pressure and that proposed by the Licensee does not warrant imposition of the Code requirement.
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Conclusions:==
Based on the above evaluation, it is concluded that the Code requirement is impractical and that the alternative test proposed by the Licensee, in conjunction with the other NDE requirements, will ensure an acceptable level of inservice structural integrity. Compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
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.a 3.4.3 Class 3 System Pressure Tests i
3.4.3.1 Recuest for Relief No. RR-6. Revision 0. Hydrostatic Test of the Class 3 Hiah Pressure Portion of the Hiah Pressure Coolant In.iection System Code Requirement: Section XI, Subarticle IWD-5223(a) requires a hydrostatic pressure test of this Class 3 portion of the High Pressure Coolant Injection (HPCI) System at a test pressure of at least 1.10 times the system pressure for systems with Design Temperature of 200*F or less, and at least 1.25 times the system pressure for systems with Design Temperature above 200*F. The system pressure shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested.
Licensee's Code Relief Recuest: Relief is requested from performing the Code-required hydrostatic pressure test of the Class 3 portion of the HPCI System (approximately two-foot section of pipe between F035 and F059) at the required test pressure.
Licensee's Proposed Alternative Examination: The Licensee states that this high pressure portion of piping will be included in the hydrostatic test of the low pressure portion of the system.
Licensee's Basis for Recuestino Relief: The Licensee states that the piping between E41-F059 and E41-PCV-F035 (line number 11-2-605) has a design rating of 460 psig, whereas the piping downstream of E41-PCV-F035 has a design pressure rating of 150 psig. The E41-PCV-F035 valve is a pressure control valve providing a pressure drop across the valve, thus preventing overpressurization of the low pressure portion of the system. As this valve is designed with holes in the valve 16
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< disc, it cannot'be used as a boundary valve' for the hydrostatic-pressure test, thus making the. hydrostatic test requirement impractical.
Evaluation: The system's design.does not permit pressurizing this section of piping to the Code-required pressure without either extensive temporary valve modifications'or overpressurizing the low pressure portion of the syftam.
Because of this, the' Code-required test. pressure for this y
two-foot section of piping is impractical. :The visual inspection of the two-foot section of piping during the alternative pressure test will provide a reasonable assurance
.of the integrity of the piping.:
==
Conclusions:==
' Based on the above evaluation, it is concluded that' the Code requirement. is impractical and that the alternative test proposed by the Licensee will ensure an acceptable level of inservice structural integrity. Compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recomended that relief be granted as requested.
l 3.4.4 Genera'i (No relief requests) 3.5 General 3.5.1 Ultrasonic Examination Techniaues (No relief requests).
3.5.2 Exemoted Components (No relief requests) i 17 l
9 l
3.5.3 Q.tflitt 3.5.3.1 Reauest for Relief No. RR-5. Revision 0. Comoonent Suonorts or the Service Water Pioino for the Four Diesel Fuel Storace.-
)
Chambers I
Code Requirement: Section XI, Table IWF-2500-1 requires a 100%
visual (VT-3) examination of the component supports as defined by Figure IWF-1300-1.
Licensee's Code Relief Reauest: Relief is requested from removing the fire retardant from the component supports on the Service Water piping for the four Diesel Fuel Storage Chambers in order to perform the required visual examinations.
Licensee's Proposed Alternative Examination: The Licensee states that these supports will be visually examined onca each interval with the fire retardant Pyrocrete 102 remaining on the supports. They will be examined for any condition which might indicated that the support has been structurally degraded (i.e., severely cracked or missing Pyrocrete, support entirely detached from the component, etc.).
Licensee's Basis for Reauestina Relief: The Licensee states that the component supports on these lines are buried in fire retardant Pyrocrete 102 as mandated per 10 CFR 50, Appendix R, Fire Prutection Program for Nuclear Power Facilities prior to January 1,1979, and Branch Technical Position APCSB 9.5-1, Appendix A, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976. This material has been applied to cover the entire support from the component to the l
ceiling / wall for all component supports.
l (a) This material (Pyrocrete 102) is an extremely hard, rigid material. When applied, this material is considered to be a permanent feature of the system to endure through the l
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e.
life span of the facility as it is in this case. To remove this material from these supports and components would require chipping the material from the support / component. This poses potential damage to the supports and/or components.
(b) The subject supports represent approximately five percent of the total Service Water System enmponent supports, and thus do not compose a large percentage of the Service Water System supports.
(c) With the exception of the two component supports at the south end of each header line (lines 2-SU-235-18-157 and 1-SW-217-18-157) all the remaining supports are basic U-bolt type supports. These U-bolt type supports have no moving parts, thus eliminating the problems associated with supports having moving parts and reducing the possible problem.e which these supports might encounter.
The remaining two supports are rigid strut type supports.
Evaluation: Removal of the permanently installed fire retardant of the type described above in order to perform the visual examination of the subject component supports is impractical because the possibility exists for damage to the supports and components. Because the Pyrocrete is a hard, j
rigid material, any movements of the supports should cause noticeably visible damage to the Pyrocrete. Therefore, the Licensee's proposed visual examination of these suoports with the fire retardant in place is an acceptable alternative to the i
Code requirement.
==
Conclusions:==
Based on the above evaluation, it is concluded that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a 1
compensating increase in the level of quality and safety.
Therefore, it is recommended that relief be granted as requested.
19 9
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3.5.3.2 Reauest for Relief No. RR-8. Revision 0. Functional Testino of Snubbers N_01E: The functional testing of snubbers is not included in this evaluation.
Functional tests are not within the scope of this document and will be evaluated elsewhere.
20
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r 4.
CONCLUSION Pursuant to 10 CFR 50.55a(g)(6), it has been determined that certain Section XI required inservice examinations are impractical.
In these cases, the Licensee has demonstrated that either the proposed alternatives would provide an acceptable level of quality and safety or that' compliance with the requirements would result in hardships or unusual difficulties without a -
compensating increase in the level of quality and safety.
This technical evaluation has not identified any practical method by which the existing Brunswick Steam Electric Plant, Unit 1, can meet all the specific inservice inspection requirements of Section XI of the ASME Code.
Requiring compliance with all the exact Section XI required inspections would require redesign of a significant number of plant systems, acquisition of sufficient replacement components, installation of the new components, and a baseline examination of these components. The reactor pressure vessel and a number of the piping and component support systems would require redesign to meet the specific inservice examination provisions. Even after the redesign efforts, complete compliance with the Section XI examination requirements probably could not be achieved. Therefore, it is concluded that the public interest is not served by imposing certain provisions of Section XI of the ASME Code that have been determined to be impractical.
Pursuant to 10 CFR 50.55a(g)(6), relief is allowed from these requirements which are impractical to implement.
The development of new or improved examination techniques will continue to be monitored. As improvements in these areas are achieved, the NRC may require that these techniques be incorporated in the next inspection interval ISI program plan examination requirements.
Based on the review of the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan, through Revision 1, the Licensee's responses to the NRC's Request for Additional Information, and the recommendations for granting relief from the ISI examination requirements that have been determined to be impractical, it has been concluded that the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval Inservice 21
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Inspection Program Plan, through Revision 1, with the exception of Request for Relief RR-8 which is not evaluated in this report, is acceptable and in compliance with 10 CFR 50.55a(g)(4).
I' 1
22 4
m.-_____________.
__._._._.m____.____
L 5.
REFERENCES 1.
Code of Federal Regulations, Volume 10, Part 50.
2.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1:
1980 Edition through Winter 1981 Addenda 1974 Edition through Summer 1975 Addenda Code Cases - Nuclear Components 3.
Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval Inservice Inspection Program Plan, Revision 0, dated November 10, 1986.
4.
NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant Boundary Inservice Inspectioh and Testing," ar.d Section 6.6, " Inservice Inspection of Class 2 and 3 Components," July 1981.
5.
Letter, dated April 3, 1987, E.D. Sylvester (NRC) to E.E. Utley
[ Carolina Power and Light Company (CP&L)], " Request for Additional Information on Second 10-Year Interval Inservice Inspection Program Plan for Brunswick Steam Electric Plant, Units 1 and 2."
6.
Letter, dated June 4, 1987, S.R. Zimmerman -(CP&L) to NRC, " Response to Requests for information."
7.
Letter, dated August 19, 1987, S.R. Zimmerman (CP&L) to NRC, " Inservice Inspection Program Additional Information."
8.
Regulatory Guide ~1.147, Revision 5, " Inservice Inspection Code Case Acceptability, ASME Section XI Division 1," dated August 1986.
9, NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," Revision 1, dated July 1980.
- 10. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," dated Nove:nber 1980.
j 23
i
]
3
- r..
2 NRC Insp$ction and Enforcement Bulletin 80-13, " Cracking in Core Spray 11.
Spargers," dated May 12, 1980.
- 12. NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," August 1981.
- 13. Regulatory Guide 1.150, " Ultrasonic Testing of Reactor Vessel Welds-During Preservice and Inservice Examinations," Revision 1, dated February 1983.
- 14. Generic Letter 34-11 " Inspections of BWR Stainless Steel Piping,"
April 19, 1984.
24
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E"!F BIBUOGRAPHIC DATA SHEET EGG-SD-7914
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J. virk..mo.werirL.
3 L.v.. Lama Technical Evaluation Report on the First 10-Year Interval Inservice Inspection Program Plan: Carolina Power and Light Company, Brunswick Steam Electric Plant,
. o."
o" a"'u"o Unit 1, Docket Number 50-325 l
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Januarv 1988
.o.n...on.uw.o B.W. Brown, J.D. Mudlin l
January 1988
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EGhG Idaho, Inc.
P. O. Box 1625 Idaho Falls, ID 83415-2209 FIN-06022 (Project 5)
,.o.....o......r,o.........L..........,,.c, i i. r.. o,..m.1 Materials Engineering Branch Technical Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
,3.w,.L..mr..vNov.5 Y.43r..cr 4200 egres.c ses This report presents the results of the evaluation of the Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval Inservice Inspection (ISI) Program Plan, through Revision 1, submitted August 19, 1987, including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements which the Licensee has determined to be impractical. The Brunswick Steam Electric Plant, Unit 1, Second 10-Year Interval ISI Program Plan is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous PSI and ISI reviews. The requests for relief from the ASME Code requirements which the Licensee has determined to be impractical for the second 10-year inspection interval are evaluated in Section 3 of this report.
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