ML20198N383

SER Accepting Second 10-yr Interval ISI Program Plan & Associated Request for Relief for Plant,Units 1 & 2

ML20198N383
Person / Time
Site:	Byron
Issue date:	12/30/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198N388	List: ML20198N383
References
NUDOCS 9901060116
Download: ML20198N383 (10)

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UNITED STATES 1

NUCLEAR REGULATORY COMMISSION s

WASHINGTON D.C. 30006 4001 k*****

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l

RELATED TO THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR RELIEF COMMONWEALTH EDISON COMPANY BYRON STATION. UNITS 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455 l

1.0 INTRODUCTION

The Technical Specifications (TS) for Byron Station, Units 1 and 2, state that the inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (Code) and applicable addenda as required by 10 CFR 50.55a(g),

except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the Byron Station, Units 1 and 2, second 10-year ISI interval is the 1989 Edition.

By a letter dated February 23,1996, Commonwealth Edison Company (Comed, the licensee) submitted its Second 10-Year Interval inservice inspection Plan, Revision 0, and associated requests for relief and by letter dated August 12,1996, the licensee submitted Revision 1 of its s

second 10-year ISI program plan for Ej on Station, Units 1 and 2. Additionalinformation vtas provided by the licensee in its letter dated December 16,1996.

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[ l 2.0 EVALUATION l

l The staff, with technical assistance from its contractor, the Idaho National Engineering and i

Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its Second 10-Year ISI Program Plan, Revisions 0 and 1 for Byron Station, Units 1 and 2. Based on the results of the review, the staff adopts the contractor's conclusions anc' recommendations presented in the Technical Letter Report (TLR) attached.

The requests for relief from the ASME Code requirements that the licensee has submitted for the second 10-year inspection interval are evaluated in the fellowing sections.

Request for Relief No.12R-01 (Revision 2): ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, item B1.11 requires 100 percent volumetric examination, as defined by Figure IWB-2500-1, of all circumferential shell welds. Item B1.21 requires volumetnc examination, as defined by Figure IWB-2500-3, of the accessible length of all circumferential head welds.

Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested from performing the volumetric examination, to the extent required by the Code, for reactor pressure vessel (RPV) circumferential shell Weld RPVC-WR29 and lower head circumferential head Weld RPVC-WR16. In addition, the licensee proposed an alternative to the augmented RPV examination required by 10 CFR 50.55a(g)(6)(ii)(A).

The Code requires 100 percent examination of the subject RPV welds. However, examination for each of these welds is restricted by physical obstructions that preclude the 100 percent volumetric examination. For the RPV lower disk-to-dutchman weld (RPVC-WR16), the examination is restricted by 58 instrument tubes that lim lt the examination to 81 percent of the Code-required volume. For the lower shell course-to-dutchman weld (RPVC-WR29), access for examination is restricted by six core barrel locating lugs that allow examination of only 57 percent of the Code-required volume. These restrictions make the Code coverage requirements impractical for the subject welds. To meet the Code requirements, the RPV would require design modifications to allow access for complete examination. Imposition of this requirement would create a considerable burden on the licensee.

The licensee can examine a significant portion of these welds and can completely examine the

- remaining RPV welds. In addition, a VT-1 visual examination is being performed on the welds and heat-affected zones (HAZs) using a submersible robot. The combination of these examinations will detect any existing pattem of degradation and provide reasonable assurance of the RPV's structuralintegrity.

Regarding the one-time volumetric examination required by 10 CFR 50.55a(g)(6)(ii)(A), the licensee proposed an alternative for Unit 1 in a separate submittal that was evaluated and authorized in an NRC Safety Evaluation dated August 28,1998. Since the one-time augmented examination requirement has been satisfied at Byron Station, Unit 1, authorization of the licensee's proposed alternative for Unit i remains authorized. However, the augmented RPV examination has not been performed for Unit 2. The staff cannot evaluate the relief for

g this weld until the augmented volumetric examinations required by 10 CFR 50.55a(g)(6)(li) are satisfied for Byron, Unit 2, in as much as 10 CFR 50.55a(g)(6)(ii)(A)(1) revoked all previously granted reliefs for the extent of volumetric examinations under B1.10. Therefore, relief for the Section XI examination of RPV she!! Weld RPVC-WR29 for the second 10-year ISI interval can not be granted until the regulations addressing the augmented RPV examination have been satisfied for Unit 2. The sta*f determined that based on the impracticality of meeting l

the Code requirements and the reasonable assurance of structuralintegrity provided by the l

examinations that are being performed, the staff concluded that relief is granted pursuant to l

10 CFR 50.55a(g)(6)(i) with the exception of Weld RPVC-WR2g for Byron, Unit 2. Relief for the i

Unit 2 Weld RPVC-WR29 !s denied.

Requsst for Relief No.12R-02 (Revision 2): ASME Code,Section XI, Table IWB-2500-1, Exarnination Category B-D, Item.B3.90, requires 100 percent volumetric examination as defined by Figure IWB-2500-7 for all RPV nozzle-to-vessel Pursuant to 10 CFR 50.55a(g)(iii), relief is requested from performing the volumetric examinations to the extent required by the Code for Unit 1 and 2 RPV nozzle-to-vessel Welds RPVN-A, RPVN-D, RPVN-E, and RPVN-H.

The Code requires 100 percent volumetric examination for the subject nozzle-to-vessel welds.

However, complete examination is not possible due the extension lip that partially obstructs access to the weld. Therefore, the volumetric examination is impractical to perform to the extent required by the Code. To meet the Code requirements, the nozzle-to-vessel welds l

would require design modifications to allow access for examination. Imposition of this requirement would create a considerable burden on the licensee.

The licensee can complete a significant portion ( 31%) of the Code-required volumetric erunination. Therefore, existing pattems of degradation will be detected and reasonable assurance of the structuralintegrity of the subject nozzle-to-vessel welds will be provided.

The staff concluded that the Code requirements are impractical for the subject nozzle-to-vessel welds, and reasonable assurance of structuralintegrity of the subject welds is provided by performing the Code-required volumetric examination to the extent practical. Therefore, relief is granted pursuant to 10 CFR 50.56a(g)(6)(i).

Request for Relief No.12R 04: ASME Code,Section XI, Table IWB-2500-1, Examination Category B-G-1, item B6.10, requires a surface examination of all RPV closure head nuts each 10-year interval. Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to perform a VT-1 visual examination in lieu of the Code-required surface examination for the RPV closure head nuts, utilizing the acceptance criteria of IWB-3517, as delineated in the 1989 Edition of ASME Section XI.

The licensee has requested relief from performing the Code-required surface examination on the RPV closure head nuts. As an attemative, the licensee proposed to perform a VT-1 visual examination. It should be noted that all items in Examination Category B-G-1 with the r

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.T exception of the RPV closure head nuts and the closure studs (when removed), require VT-1 visual examination and volumetric examination (as applicable).

Typical conditions that would require corrective action prior to putting closure head nuts back into service include corrosion, deformed or sheared threads, deformation, and degradation mechanisms such as boric acid attack. The Code examination requirement for closure head nuts is a surface examination. Surface examination procedures are typically qualified for the detection of linear discontinuities (cracks, aligned pitting, and corrosion) with corresponding acceptance criteria for rejectable linear flaw lengths. Surface examination acceptance criteria are not provided for the 1989 Edition of the Code, item B6.10, as they were in the course of preps:ation. Without clearly defined acceptance criteria, relevant conditions that require corrective nieasures may not be adequately addressed.

Article IWB-3000. Acceptance Standards, IWB-3517.1, Visual Examination VT-1, describes relevant conditions that require corrective action prior to continued service for bolting and associated nuts. IWB-3517.1 requires crack-like flaws to be compared to the flaw standards of IWB-3515 for acceptance. Because the VT-1 acceptance criteria includes the requiremont for evaluation of crack-like indications and other relevant conditions requiring corrective action, such as deformed or sheared threads, localized corrosion, deformation of the part, and other degradation mechanisms, il can be concluded that the VT-1 visual examination provides a more comprehensive assessment of the condition of the closure head nut. The staff determined thst VT-1 visual examination provides an acceptable level of quality and safety. In addition, it is noted that the 1989 Addenda of Section XI changes the requirement for the subject reactor pressure vessel closure head nuts from surface to VT-1 visual examination and provides appropriate acceptance criteria.

The licensee has proposed, as an alternative to the Code-required surface examination of reactor pressure vessel closure head nuts, to perform a VT-1 visual examination. The staff determined that based on the comprehensive assessment that the VT-1 visual examination provides, and considering that the later editions of the Code require only a VT-1 visual examination on reactor pressure vessel closure head nuts, the licensee's proposed alternative provides an acceptable level of quality and safety by providing reasonable structuralintegrity.

Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Request for Relief 12R-03 (Revision 2): ASME Code,Section XI, Table IWB-2500-1, Examination Category B-D, items 83.110 and B3.120, require 100 percent volumetric examination, as defined by Figure IWB-2500-7, for all pressurizer nozzle-to-vessel welds and nozzle inside radius sections.

Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee has determined that conformance with the Code-required volumetric examination of the pressurizer surge nozzle-to-vessel"; eld and inside radius section would result in hardship or unusual difficulty without a compensating increase in quality and safety. As an alternative, the licensee proposes to perform the Code-required VT-2 examination during system pressure testing, per Category B-P, on the pressurizer each refueling outago to verify the leaktight integrity of these areas. In addition, the pressurizer

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surge nozzle-to-vessel weld and nozzle inside radius section will be volumetrically examined if i

the lower head insulation is removed for any reason.

The Code requires 100 percent volumetric examination of the pressurizer surge nozzie-to-vessel welds and IR sections. However, access to the vessel surface is obstructed by multi-layered, stainless steel mirror insulation that is difficult to remove. As an alternative, the licensee proposes to perform a VT-2 visual examination during the Code-required pressure teste.

The licensee attempted to gain access to the subject examination area by cutting away and lowering the insulation at Byron. Unit 2, but this did not work. To gain access for examination at the Byron plant, the insulation covering the pressurizer lower head has to be completely removed, which requires disconnection of the heater assemblies and is time and dose intensive. Disconnection of the heater assemblies could also cause damage to the heaters at the tenaination points. Damage to an insulator would require total heater replacement. Based on survey results made during a Braidwood outage, the radiation exposure to remove the insulation, prepare the surface, and perform a limited examination e the nozzles is estimated to be 150 man-rem. Therefore, the staff determined that compliance with the Code regtcirements would result in considerable hardship for the licensee.

The subject examination areas were examined during fabrication and prior to service and found to be acceptable. These examinations ensured the structuralintegrity of the pressurizer surge nozzles prior to service. Assurance of their inservice operational readiness is provided by the Code-required pressure tests of the pressurizer and interconnected piping. Additional assurance is provided by the examination of the adjacent nozzle-to-safe end weld and other Examination Category B-D nozzles in the pressurizer, reactor pressure vesscl, and steam generators. These other areas do not experience the same conditions and stresses as the pressurizer surge nozzle, but in many cases, the conditions and stresses are similar.

Therefore, examination of the other Class 1 nozzles can be used as an indicator of generic degradation that could occur in the pressurizer surge nozzle and provide reasonable assurance of the structuralintegrity of the surge nozzle.

The subject examination areas are covered with insulation that would have to be removed to gain access for examination. The staff determined that based on the burden associated with removing this insulation, the potential damage to the ceramic heater connections, the assurances of structuralintegrity provided by the examinatien of other Class 1 nozzles, and performance of the Code-required pressure tests, the staff concludes that compliance with this requirement would result in a hardship without a compensating increase in quality and safety.

Therefore, the staff concluded that the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

3.0 CONCLUSION

The staff concluded that in the Byron Station, Units 1 and 2, Second Interval Inservice Inspection Program Plan, the licensee's responses to the NRC's requests for additional j

information, and the recommendations for granting relief from the ISI examinations that can not

l be performed to the extent required by Section XI of the ASME Codes, no deviations from regulatory requirements or commitments were identified.

Pursuant to 10 CFR 50.55a(g)(6)(i), the staff concluded that certain inservice examinations can not be performed to the extent required by Section XI of the ASME Code. In the cases of Requests for Relief 12R 01, Revision 2 (except for Unit 2 Weld RPVC-WR29), and 12R-02, Revision 2, the licensee has demonstrated that specific Section XI requirements are impractical. Furthermore, the staff concluded that the licensee's proposed attemative provided reasonable assurance of structuralintegrity of the subject weld. Therefore, the staff concluded that relief is granted and the alternatives imposed pursuant to 10 CFR 50.55a(g)(6)(i) as requested, with the exception of Unit 2 Weld RPVC-WR29. For Unit 2 RPV circumferential i

shell Weld RPVC-WR29, relief can not be evaluated until the augmented RPV examination required by the regulations has been satisfied. The relief granted will not endanger life, property, or the common defense and security and is otherwise in the public ir,terest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the faciity.

l The staff concluded that the licensee's proposed attemative contained in Request for Relief 12R-04, Revision 2, provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant 10 CFR 50.55a(a)(3)(i) for the current interval.

The staff concluded for the licensee's proposed altemative contained in Request for Relief No.

12R-03, that Code compliance will result in hardship or unusual difficulty without a i

compensating increase in safety because the licensee's proposed alternative provides l

reasonable assurance of the structuralintegrity of the subject components. Therefore, the attemative contained in Request for Relief No.12R-03 is authorized pursuant to 10 CFR 50.55a(s)(3)(ii) for the current interval.

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Attachment:

Technical Letter Report Summary Table of Relief Requests Dated: December 30, 1998 e

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Technical Evaluation Report on the Second 10-year Interval inservice inspection Program Plan:

Commonwealth Edison Company, Byron Nuclear Power Station, Units 1 and 2, 4.

Docket Numbers 50-454 and 50-455 M. T. Anderson, C. T. Brown, E. J. Feige, A. M. Porter Published September 1998 Idaho National Engineering and Environmental Laboratorf Materials Physics Group Lockheed Martin Idaho Technologies Company ldaho Falls, Idaho 83415 Prepared for the Civil and Geosciences Branch Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 JCN No. J2229 (Task Order A16)

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ABSTRACT This report presents the results of the evaluation of the Byron Station, Units 1 and 2, SecondIntervalinservice Inspection Program Plan, Revision 1, dated August 12,1996, including the requests for relief from the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI requirements that the licensee has determined to be impractical. The Byron Station, Units 1 and 2, SecondIntervalInservice Inspection Program Plan, Revision 1, is evaluated in Section 2 of this report. The inservice inspection (ISI) program plan is evaluated for (a) compliance with the appropriate editionladdenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related j

commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.

This work was funded under:

U.S. Nuclear Regulatory Commission JCN No. J2229 Task Order A16 Technical Assistance in Support of the NRC Inservice Inspection Program ii

SUMMARY

l The licensee, Commonwealth Edison Company, prepared the Byron Station, Units 1 l

l and 2 SecondIntervalinservice Inspection Program Plan, Revision 1, to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME)

Code,Section XI. The second 10 year interval is effective September 1,1996, through September 15,2006, for Unit 1 and August 16,1998, through August 15, 2008, for Unit 2.

The information in the Byron Station, Units 1 and 2, SecondIntervallaservice Inspection Program Plan, Revision 0, submitted February 23,1996, was reviewed.

Included in the review were the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. As a result of this review, a request for additional information (RAll was prepared de' cribing the information s

and/or clarification required from the licensee in order to complete the review. The licensee provided the requested information in the submittal dated August 12,1996, which included the Byron Station, Units 1 and 2, SecondIntervalinservice inspection i

Program Plan, Revision 1. Additional clarification was required and was requested in a supplemental NRC RAl. This information was provided by the licensee in a December 16, 1996, submittal.

Based on the review of the program plan, the licensee's response to the Nuclear Regulatory Commission's RAI, and the recommendations for granting relief from the ISI examinations that cannot be performed tc the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified in the Byron Station, Units 1 and 2, Second IntervalInservice Inspection Program Plan, Revision 1.

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CONTENTS i

A B ST R A CT....................................................... il

SUMMARY

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1. I NT R O DU CTI O N................................................. 1

2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN................. 3 2.1 Documents Evaluated......................................... 3 2.2 Compliance with Code Requirements............................... 3 2.2.1 Compliance with Applicable Code Editions....................... 3 2

2.2.2 Acceptability of the Examination Sample........................ 4 2.2.3 Exemption Criteria........................................ 4 1

2.2.4 Augmented Examination Commitments......................... 5

2. 3 C oncl usi on................................................. 5 4

3. EVALUATION OF RELIEF REQUESTS

.................................6 3.1 Cla ss 1 Com ponents.......................................... 6 3.1.1 Reactor Pressure Vessel.................................... 6 3.1.1.1 Request for Relief No.12R 01 (Revision 2), Examination Category B A, items B1.11, and B1.21, Reactor Pressure Vessel (RPV) Welds. 6 3.1.1.2 Request for Relief No.12R 02 (Revision 2), Examination Category B-D, item B3.90, Reactor Pressure Vessel Nozzle to-Vessel Welds.. 9 3.1.1.3 Request for Relief No.12R-04, Examination Category B G 1, item B6.10, Reactor Pressure Vessel Closure Head Nuts........ 11 3.1.1.4 Request for Relief No.12R 18, Use of Code Case N 521, Alternative Rules for Deferral of Inspections of Nozzle to-vessel Welds, inside Radius Sections, and Nozzle-to Safe End Welds of a Pressurized

'Nater Reactor (PWR) Vessel............................ 12 3.1. 2 Pre ssuriz er............................................. 12 3.1.2.1 Request for Relief 12R-03 (Revision 2), Examination Category B D, items B3.110 and B3.120, Pressurizer Surge Nozzle-to-Vessel Weld and inside Radius Section.............................. 12 3.1.3 Heat Exchangers anc' Steam Generators........................ 15

- 3.1.4 Piping Pressure Boundary.................................. 15 3.1.5 Pom p Pressure Boundary................................... 15 3.1.6 Valve Pressure Boundary................................... 15 3.1. 7 ' G e ne r a l............................................... 1 5

3. 2 Class 2 Com ponents......................................... 15 3.2.1 Pressure Vessels

........................................15 3.2.1.1 Request for Relief No.12R-05 (Revision 2), Examination Category C-B, item C2.32, Residual Heat Removal Heat Exchanger (RHRHX)

Nozzle-to-Shell Weld.................................. 15 3. 2. 2 Pi ping................................................ 1 5 3.2.2.1 Request for Relief No.12R 07, Examination Categories C-F-1 and C F 2, items C5.11N, C5.12N, C5.51N and C5.52N, Class 2 Piping W e ld s............................................ 1 5 iv

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l 3.2.2.2 Request for Relief No.12R 08, Examination Category C-F-1 and i

C-F-1, items C5.12, C5.12N, C5.52 and C5.52N, Class 2 j

Longitudinal Piping Welds.............................. 15 3.2.2.3 Request for Relief No.12R-09, Examination Categories C-F-1 and C-F 1, items C5.41 and C5.81, Circumferential Branch Connection W e ld s............................................ 1 6 3. 2. 3 Pum p s................................................ 1 6 i

3.2.3.1 Request for Relief No.12R-06, Examination Category C-C, item C3.30, Integrally Welded Attachments to Pumps.............. 16 f

3. 2.4 Valve s................................................ 1 6

3. 2. 5 G e ne ral............................................... 1 6 3.3 Class 3 Components....................................... 16 l

3.4 Pressure Tests............................................. 16 3.4.1 Class 1 System Pressure Tests.............................. 16 3.4.2 Class 2 System Pressure Tests.............................. 16 l

3.4.2.1 Request for Relief No.12R 10, Examination Category C 's. Items C7.30, C7.40i C7.70 and C7.80, Pressure Testing of Containment Penetration Piping Attached to Non-classed Piping............. 16 3.4.3 Class 3 System Pressure Tests.............................. 17 3.4.3.1 Request for Relief No.12R-14, Examination Category D-8, item D2.10, Alternate Rules for System Pressure Testing of Diesel Generator Subsystem Components 17 3.4.3.2 Request for Relief No.12R 16, Use of Code Case N 498-1, Altemative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems for Class 3 Systems................... 17 3.4.4 General

..............................................17 l

3.4.4.1 Request for Relief No.12R 15, Use of Code Case N 416-1, Alternate Testing for Class 1, Class 2, and Class 3 Welded Repaired / Replaced Comp onen ts....................................... 1 7 3.5 General

..................................................17 3.5.1 Ultrasonic Examination Techniques............................ 17 3.5.2 Exempted Components.................................... 17

3. 5. 3 O the r................................................ 1 7 3.5.3.1 Request for Relief No.12R 11, Paragraph IWA 5242(a), VT-2 Visual Examination of Insulated Components..................... 17 3.5.3.2 Request for Relief 12R 12, IWA 5250(a)(2), Corrective Measures for Bolted Connections

..................................17 3.5.3.3 Request for Relief No.12R-13, IWF 5300, inservice Examination and Test Requirements for Snubbers......................... 18 3.5.3.4 Request for Reliof No.12R-17 (Revision 2), Use of Code Case N-509, 1

Altemate Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments,Section XI, Division 1 18 l

4. C O N C LU SI O N................................................. 19 1

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5. R E F E R E N C E S................................................. 2 0 V

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TECHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:

l COMMONWEALTH EDISON COMPANY, BYRON NUCLEAR POWER STATION, UNITS 1 AND 2, DOCKET NUMBERS 50-454 AND 50-455

1. INTRODUCTION Throughout the service life of a water-cooled nuclear power facility,its components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1,2, and 3 are required by 10 CFR 50.55a(g)(4) (Reference 1) to 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 successive 120-month inspection intervals 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 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements s1t forth in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein, and subject to Nuclear Regulatory Commission (NRC) approvv. The licensee, Commonwealth Edison Company (Comed), has prepared the Byron Station, Units 1 and 2, SecondIntervalinservice Inspection Program Plan, Revision 1 (Reference 3), to moet the requirements of the 1989 Edition of the ASME Code,Section XI. The second 10-year interval is September 1,1996, through September 15,2006, for Unit 1 and August 16,1998, through August 15,2008, for Unit 2. As allowed by Section XI, IWA-2400(c), the first interval for Byron Station Unit 1 was extended from September 15, 1995, to August 31,1996, and from August 22,1997, to August 15,1998, for Unit 2.

Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the Code requirements may be used when authorized by the NRC. The licensee must demonstrate either that the proposed alternatives provide an acceptable level of quality and safety, or that Code compliance would result in hardship or unusual difficulty without a compensating increase in quality and safety. Pursuant to 10 CFR 50.55a(g)(5)(iii),if the licensee determines that l

compliance with certain Code examination requirements is impractical for its facility, the i

licensee shall submit information to the NRC to support that determination. Pursuant to l

10 CFR 50.55a(g)(6)(i), the NRC will evaluate the licensee's determination that Code requirements are impractical. The NRC may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, 1

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or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were l

l imposed on the facility.

l The information in the Byron Station, Units 1 and 2, SecondIntervalinservice l

inspection Program P/an, Revision O (Reference 4), submitted February 23,1996, was reviewed, including the requests for relief from the ASME Code Section XI requirements j

that the licensee has determined to be impractical. The review of the Inservice Inspection (ISI) Program Plan was performed using the Standard Review Plans of NUREG-0800 (Reference 5), Section 5.2.4, " Reactor Coolant Boundary inservice Inspections and i

Testing," and Section 6.6, " Inservice Inspection of Class 2 and 3 Components."

in a letter dated June 13,1996, (Reference 6), the NRC requested additional information that was required to complete the review of the ISI Program Plan. The requested information was provided by the licensec in the " Response to Request for Information Related to the inservice inspection Program Plan" dated August 12,1996 (Reference 3). In this response, the licensee provided Revision 1 of the Program Plan, which included several revised requests for relief and fcur new requests. In response to a supplementary NRC request for additional information (RAl) dated October 16,1996 (Reference 7), additional information was provided by letter dated December 16,1996 (Reference 8). Completion of this report was delayed until the augmented reactor pressure vessel examinations required by 10 CFR 50.55a(g)(6)(ii)(A) were satisfied.

The Byron Station, Units 1 and 2, Second Intervalinservice Inspection Program Plan, 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) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISl-related commitments identified during the NRC's previous reviews.

The requests for relief are evaluated in Section 3 of this report. As noted in Section 3, some requests for relief for the second 10-year ISI interval were evaluated in separate NRC l

Safety Evaluation Reports (SER), as requested by the licensee, to expedite the review.

Unless otherwise statsd, references to the Code refer to the ASME Code,Section XI, 1989 Edition. 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 This evaluation consists of a review of the applicable program documents to determine l

whether or not they are in compliance with the Code requirements and any previous license conditions pertinent to ISI activities. This section describes the submittals reviewed and the results of the review.

2.1 Documents Evaluated l

Review has been completed on the fellowing information from the licensee:

Byron Station, Units 1 and 2, SecondIntervalinservice Inspection Program Plan, l

Revision 0, submitted February 23,1996 (Reference 4):

Response to RAI, including the Byron Station Units 1 and 2, Second /nterval I

/nservice /nspection Program Plan, Revision 1, dated August 12,1996 (Reference 3); and i

Response to supplementary RAI dated December 16,1996 (Reference 8).

e 2.2 Compliance with Code Requirements 2.2.1 Compliance with Applicable Code Editions inservice inspection program plans are to be basad on Section XI of the ASME Code defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting dates of July 1,1996, for Unit 1 and August 16,1998, for Unit 2, the Code applicable to the l

second interval ISI program is the 1989 Edition. As stated in Section 1 of this report, the l

licensee has prepared the Byron Station, Units 1 and 2, SecondIntervalinservice l

inspection Program P/an, Revision 1 to meet the requirements of 1989 Edition of the Code.

l In accordance with 10 CFR 50.55a(c)(3),10 CFR 50.55a(d)(2), and 10 CFR 50.55a(e)(2), ASME Code Ca::es may be used as alternatives to Code requirements. ASME Code Cases that have been found suitable for use by the NRC are listed in Regulatory Guide 1.147, /nse/v/ce Inspection Code Case Acceptab//ity, (Reference l

9) with any conditions the NRC may tave imposed. When used, these Code Cases must be implemented in their entirety. Pubished Code Cases awaiting approval and subsequent listing in Regulatory Guide 1.147 may be adopted only if the licensee requests, and the NRC authorizes, their use on a case-by-case basis.

The licensee's second 10 year ISI program includes the Code cases listed below.

These Code cases either have been approved for use in Regulatory Guide 1.147 or are included as requests for relief.

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o Code Case N-408-2 Altemative Rules for Examination of Class 2 Piping Code Case N 416-1 Altemative Pressure Test Requirement for Welded Repairs or Installation of Replacement items by Welding, Class 1, 2, and 3 (evaluated in NRC SER dated January 13,1998)

Code Case N 460 Alternative Examination Coverage for Class 1 and 2 Welds Code Case N 491 Alternative Rules for the Examination of Class 1, 2, and 3 MC Components and Supports of Light Water Cooled Power Plants Code Case N-498-1 Altemative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems (evaluated in NRC SER dated January 13,1998)

Code Case N 509 Altemative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments Code Case N 521 Alternative Rules for Deferral ofInspections of Nozzle to-Vessel Welds, inside Radius Sections, and Nozzle-to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessel (evaluated in NRC SER dated January 13,1998)

Code Case N-522 Pressure Testing of Containment Penetration Piping (evaluated in NRC SER dated January 13,1998)

Code Case N 524 Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping (evaluated in NRC SER dated January 13, 1998) 2.2.2 Acceptability of the Examination Sample 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 10 CFR 50.55a(b) except as altered by Code Case N 408-2 and Request for Relief 12R-07, end appear to be correct.

2.2.3 Exemption Criteria The criteria used to exempt components from examination shall be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, IWD 1220, and 10 CFR 50.55a(b). The exemption criteria have been applied by the licensee in accordance with the Code as discussed in the ISI Program Plan, and appear to be correct.

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l 2.2.4 Augmented Examination Commitments l

l In addition to the requirements specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:

(a) Volumetric examination of piping welds per NRC Branch Technical Position MEB 3-1 and Byron Station UFSAR Sections 3.6.1 and 3.6.2:

l (b) Examination of reactor coolant pump flywheels in accordance with NRC Regulatory Guide 1.14, Reactor Coolant Pump FlywheelIntegrity (Reference 10);

(c) Examination of reactor pressure vessel (RPV) closure studs per NRC Regulatory Guide 1.65, Materials and Inspections for Reactor Vessel Closure Studs (Reference 11)

(d) Ultrasonic examination of the RPV in accordance with NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice andInservice Examinations (Reference 12); and (e) Visual and surface examination in accordance with Byron Station UFSAR Section 10.2.3, Turbine Disk and Rotor Integrity.

The one-time, augmented reactor pressure vessel examination required by 10 CFR 50.55a(g)(6)(ii)(A) was performed during the previous 10-year interval and, therefore, is not required during the second 10-year ISI interval.

2.3 Conclusion Based on the review of the documents listed in Section 2.1, no deviations from regulatory requirements or cornmitments were identified in the Byron Station, Units f and 2, SecondIntervalinservice Inspection Program Plan, Revision 1. Note that this report

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does not include a review of the implementation of the augmented examinations,it merely records that the licensee has committed to perform them.

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3. EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has submitted for the second 10 year inspection interval are evaluated in the following sections.

3.1 Class 1 Components 3.1.1 Reactor Pressure Vessel

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3.1.1.1 Request for Relief No.12R-01 (Revision 2), Examination Category B-A, items 81.11, and B1.21, Reactor Pressure Vessel (RPV) Welds Code Requ/rement--Section XI, Table IWB-2500-1, Examination Category B-A, item B1.11 i

requires 100% volumetric examination, as defined by Figure IWB-25001, of all l

circumferential shell welds. Item B1.21 requires volumetric examination, as defined by i

Figure IWB-2500 3, of the accessible length of all circumferential head welds, i

Licensee's Code Relief Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested i

from performing the volumetric examination, to the extent required by the Code, for RPV i

circumferential shell Weld RPVC WR29 and lower head circumferential head Weld RPVC-WR16. In addition, the licensee proposed an alternative to the augmented RPV examination required by 10 CFR 50.55a(g)(6)(ii)(A).

Licensee's Basis for Requesting Relief (as stated)-

" Pursuant to 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a l

compensating increase in the level of quality and safety, and pursuant to 10 CFR EO.55a(g)(5)(iii), conformance with these Code requirements is impractical as conformance would require extensive structural modifications to the reactor vessel support structure."

" HISTORY:

" Byron Station, during Refuel Outage B1R07, conducted ultrasonic examinations of the Byron Unit 1 Reactor Vessel. This was the last refuel outage of the third period of the first inservice inspection interval and occurred in April through June 1996. Framatome Technologies Inc. (FTI) was contracted to perform the examinations with their state-of-the art 'URSULA' manipulator and their ' ACCUSONEX' UT system. The examinations were performed in accordance with the requirements in ASME Section XI, Article IWA-2232, NRC Regulatory Guide 1.150 and 10 CFR 50.55a(g)(6)(ii)(A).

"Previously granted relief request NR-1 for Byron Station Units 1 and 2 First inspection Interval for the Reactor Shell welds was subsequently revoked per l

10 CFR 50.55a(g)(6)(ii)(A) with respect to examination coverages. During the

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performance of the B1R07 examinations, physical obstructions and geometry prevented UT coverage in excess of 90% of the required volume for the above listed i

component numbers. Full 100% UT coverage was obtained for the reactor circumferential shell welds WR 7, WR 18, and WR 34 (Byron Station Reactor Vessel (s) 6 U

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I do not have any longitudinal shell welds). The limited amount of examination coverage attained for shell welds WR-16 and WR 29 is provided below in the Basis for Relief.

"FTl is contracted to perform the 10 year Reactor Examination on Byron Unit 2 during the late Fall of 1997 and, as the two Byron reactors are identical, similar coverage percentages are expected.

"A. REACTOR VESSEL CIRCUMFERENTIAL SHELL WELDS "The examination of the Unit 1 Lower Shell Course-to-Dutchman weld, RPVC WR29 is restricted by six (6) core barrellocating lugs welded to the inner surface of the vessel approximately 4 inches above the weld (see Figure 1 *). These lugs obstruct the automated UT inspection tool from examining the code required volume of the weld below each lug (156*). The FTl 'URSULA' tool has a 6 degree movement arm and the physical size of the lugs and the ' yaw' joint of the tool prevented scanning below the lugs back into the weld and surrounding base metal. All weld metal can be examined from both sides where access is available between the lugs (204'). Examination for parpendicular and parallel reflectors can cover areas accounting for 57% of the weld metal and heat affected zone (HAZ). Similarly,57% of the weld metal can be examined for transverse reflectors from two opposing directions.

"The examination of the Unit 2 lower Shell Course-to-Dutchman weld, PRVC WR29 is similarly obstructed (see Figure 1 *). All weld metal can be examined where access is available between the lugs (204*). Examinations for perpendicular and parallel l

reflectors can cover areas accounting for 57% of the weld metal and heat affected l

zone (HAZ), Similarly,57% of the weld metal can be examined for transverse reflectors from two opposing directions.

"B. REACTOR VESSEL LOWER HEAD CIRCUMFERENTIAL WELDS l

"The examination of the Unit 1 Lower Disk-to Dutchman weld, RPVC-WR16 is restricted by the 58 instrument tubes that penetrate the lower disk and physically obstruct the UT search unit and/or the search unit position device as shown in Figures 2 and 3*. Approximately 19% of the weld length cannot be examined. The weld and the HAZ received essentially 100% coverage for parallel reflectors from the Dutchman side F.nd for transverse reflectors in two opposing directions. Partial coverage is l

achieved for parallel reflectors in two opposing directions. Partial coverage for parallel l

reflectors from the disk side on the remainder of the weld resulting in an aggregate of all scan coverage of approximately 81%.

"The examination of the Unit 2 Lower Disk to-Dutchman weld, RPVC WR 16,is similarly restricted, see Figure 2 and 3. The weld and the HAZ will receive essentially 100% coverage for parallel reflectors from the Dutchman side and for transverse reflectors in two opposing directions. Partial coverage can be achieved for parallel reflectors from the disk side on the remainder of the weld resulting in an aggregate coverage of approximately 81%.

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• Figures provided by the licensee are not included in this report.

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"For welds referenced in Sections A and B above, the probability of a flaw occurring only in one of the areas not being examined is extremely small. Most future indications of significant size will be found by the examination of the weld as it is currently performed.

o "In addition, a VT-2 examination during system pressure testing per Category B-P is performed on the Reactor Vessel each refueling outage to verify leaktight integrity of these welds.

" Examination from the vessel exterior is obstructed by the concrete structure surrounding the reactor vessel (s). This reinforced concrete structure is the reactor vessel support structure. The vesselis suspended by 4 nozzle supports which distribute the loads through the surrounding concrete structure. The annulus tatween the vesselin the vicinity of the obstructed welds and the structure is approximataly 6".

No exterior access to these welds is possible from either above or below the vessel.

This concrete structure is the reactor vessel support structure and modifications to allow access to the vessel exterior are not practicable.

" Compliance with the applicable Code requirements can only be accomplished by redesigning and refabricating the Reactor Vessel and concrete support structure. Byron Station deems this course of action a hardship without a compensating increase in the level of quality and safety."

Licensee's Proposed Alternative Examination (as stated)~

"The ultrasonic examination of the Byron Unit 1 reactor vessel was performed to the maximum extent possible. No alternative volumetric examination is proposed to examine the areas not scanned due to obstructions or geometric constraints.

"A VT-1 examination was conducted on the shell welds (s) and HAZ(s) from the inside clad surface utilizing a submersible robot during the Byron Unit 1 Refuel Outage B1R07.

Additionally, a VT-2 examination during system pressure testing per Category B-P is performed on the Reactor Vessel each refueling outage to verify leaktight integrity of these welds.

"The ultrasonic examination of the Byron Unit 2 reactor vessel will also be performed to the maximum extent possible. The same obstructions and geometric constraints are expected to limit the examination. The visual examinations will be repeated for the Unit 2 RPV welds."

Evaluation-The Code requires 100% examination of the subject RPV welds. However, examination for each of these welds is restricted by physical obstructions that preclude the 100% volumetric examination. For the RPV lower disk-tc-dutchman weld (RPVC-WR16),

the examination is restricted by 58 instrument tubes that limit the examination to 81% of the Code-required volume. For the lower shell course-to-dutchman weld (RPVC WR29),

access for examination is restricted by six core barrellocating lugs that allow examination of only 57% of the Code required volume. These restrictions make the Code coverage requirements impractical for the subject welds. To meet the Code requirements, the RPV would require design modifications to allow access for complete examination. Imposition of this requirement would create a considerable burden on the licensee.

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._m The licensee can examine a significant portion of these welds and can completely examine the remaining RPV welds. In addition, a VT-1 visual examination is being i

performed on the welds and HAZ(s) using a submersible robot. The combination of these

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examinations will detect any existing pattern of degradation and provide reasonable assurance of the RPV's structuralintegrity.

Regarding the one-time volumetric examination required by 10 CFR 50.55a(g)(6)(ii)(A),

l the licensee proposed an alternative for Unit 1 in a separate submittal that was evaluated and authorized in an SER dated August 28,1998. Since the one time augmented examination requirement has been satisfied at Byron Station, Unit 1, authorization of the i

licensee's proposed alternative is not required for Unit 1. However, the augmented RPV l

examination has not been performed for Unit 2. Therefore, relief for the Section XI examination of RPV shell We!d RPVC-WR29 for the second 10 year ISI interval cannot be granted until the regulations addressing the augmented RPV examination have been satisfied for Unit 2.

Conclus/on-Based on the impracticality of meeting the Code requirements, and the l

reasonable assurance of structuralintegrity provided by the examinations that are being i

performed,it is concluded that relief should be granted pursuant to 10 CFR 50.55a(g)(6)(i) with the exception of Weld RPVC-WR29 for Byron Unit 2. Relief for this weld should not be granted until the augmented volumetric examinations required by 10 CFR 50.55a(g)(6)(ii) are satisfied for Byron Unit 2.

3.1.1.2 Request for Relief No.12R-02 (Revision 2), Examination Category B-D, item B3.90, Reactor Pressure Vessel Nozzle to Vessel Walds Code Requ/rement-Section XI, Table IWB 25001, Examination Category B-D, item B3.90, requires 100% volumetric examination as defined by Figure IWB 2500-7 for all RPV nozzle-to-vessel welds.

l L/censee's Code Relie/ Request-Pursuant to 10 CFR 50.55a(g)(iii), relief is requested from performing the volumetric examinations to the extent required by the Code for Unit 1 and 2 RPV nozzle-to vessel Welds RPVN A, RPVN-D, RPVN-E, and RPVN H.

Licensee's Basis for Requesting Relief (as stated)-

" Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested on the basis that compliance with the Code requirements is impractical as conformance would require extensive structural modifications to the reactor vessel support structure and surrounding concrete structure.

"The Reactor Vessel Outlet Nozzles were designed and fabricated with an extension lip as shown in Figure 1

• such that the Nozzle-to-vossel Welds are partially obstructed for the Code required volumetric exarnination. Tids obstruction limits the circumferential i

scan for reflectors transverse to the weld such that 19% (aggregatc) of the required l

volume cannot be examined.

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' Figures provided by the licensee are not included in this report.

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" Based on the above information, reasonable assurance of the continued inservice structural integrity of the subject welds is achieved without performing a complete

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Code examination. In addition, a VT-2 examination during system pressure testing per l

Category B-P is also performed on the Reactor Vessel each refueling outage to verify leaktight integrity of these welds.

1 JUSTIFICATION I

"The Code required volumetric examination has been completed to the maximum

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extent practical using ultrasonic examination techniques for Byron Unit 1. The RPV i

examinations are conducted using an automated techniques from the ID of the vessel.

Examination from the vessel exterior is obstructed by the reinforced concrete structure j

surrounding the reactor vessel (s) and the reactor vessel nozzle supports. The vesselis i

suspended by 4 nozzle supports which distribute the loads through the surrounding i

concrete structure. The annulus between the vessel, in the vicinity of the nozzle-to-2' vessel welds, and the structure is approximately 6" or less. No exterior access to these nozzle-to-vessel welds is possible from either above or below the nozzles. This j

concrete structure is the reactor vessel support structure and modifications to allow access to the nozzles exterior are not practicable.

l "VT-1 inspection was conducted on the weld (s) and HAZ(s) from the inside cisd surface utilizing a submersible robot during the Byron Unit 1 Refuel Outage B1R07.

j Additionally, a VT 2 examination during system pressure testing per Category B-P is performed on the Reactor Vessel each refueling outage to verify leaktight integrity of 1

j these welds.

"The ultrasonic examination of the Byron Unit 2 reactor vessel will also be performed to the maximum extent possible. The same obstructions and geometric constraints are l

expected to limit the examination. The visual examinations will be repeated for the Unit 2 RPV welds.

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" Compliance with the applicable Code requirements can only be accomplished by e

i redesigning and refabricating the Reactor Vessel (s) and/or building a structure surrounding the vessel (s). Byron Station believes this course of action is a hardship without a compensating increase in the lovel of quality and safety."

l Licensee's Proposed Altemative Examination (as stated)-

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"None; the Code required volumetric examination will be completed to the maximum j

extent practical using ultrasonic examination techniques. FTl is contracted to perform tha 10 year Reactor Examinations on Byron Unit 2 during the late Fall of 1997 and, as

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the two Byron Reactors are identical, similar coverage percentages are expected."

l Evaluation-The Code requires 100% volumetric examination for the subject nozzle-to-vessel welds. However, complete examination is not possible due the extension lip that partially obstructs access to the weld. Therefore, the volumetric examination is impractical to perform to the extent required by the Code. To meet the Code requirements, the nozzle-to-vessel welds would require design modifications to allow access for examination. Imposition of this requirement would create a considerable burden on the licensee.

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4 The licensee can complete a significant portion (~81 %) of the Code-required volumetric examination. Therefore, existing pattems of degradation will be detected and reasonable assurance of the structural integrity of the subject nozzle-to vessel welds will be provided.

Conclus/on-Considering the impracticality of the Code' requirements for the subject nozzle-to-vessel welds, and the assurance provided by performing the Code-required volumetric examination to the extent practical, it is recommended that relief be granted pursuant to l

10 CFR 50.55a(g)(6)(i).

j 3.1.1.3 Request for Relief No.12R-04, Examination Category B-G-1, item B6.10, Reactor Pressure Vessel Closure Head Nuts i

Code #squ/remont-Section XI, Table IWB 25001, Examination Category B-G-1, item B6.10, requires a surface examination of all RPV closure head nuts each 10-year interval.

l Licensee's Proposed Alternative Exam /nat/on--Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to perform a VT-1 visual examination in lieu of the Code-required j

surface examination for the RPV closure head nuts, utilizing the acceptance criteria of lWB-3B17, as delineated in the 1989 Edition of ASME Section XI.

Licensee's Basis for Proposed A/temative (as stated)-

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" Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.

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" Table IWB 2500-1 of the 1989 Edition of ASME Section XI requires a surface i

examination to be performed on the reactor vessel closure head nuts. However, Table j

IWB 25001 does not provide the corresponding ' Examination Requirements / Figure Number' and ' Acceptance Standard'. These provisions were stillin the course of preparation.

"The 1989 Edition of ASME Section XI, Category B-G 1, employs a VT-1 visual examination for nuts associated with Heat Exchangers, Piping, Pumps, and Valves (Item Numbers B6.140, B6.170, B6.200, and 86.230, respectively). These Category B-G-1 requirements also provide an Acceptance Standard, IWB 3517, for the t!T-1 i~

examinations. Accordingly, these rules are deemed by Byron Station as an acceptable j

and complete set of rules to assure the integrity of reactor vessel closure nuts.

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" Based on the above, Byron Station requests relief from the requirements specified in Table IWB-2500-1 of the 1989 Edition of ASME Section XI for reactor vessel closure head nuts".

1 Eva/ust/on-The licensee has requested relief from performing the Code-required surface j

examination on the RPV closure head nuts. As an alternative, the licensee proposed to perform a VT-1 visual examination. It should be noted that all items in Examination J

Category B-G-1 with the excegion of the RPV closure head nuts and the closure studs (when removed), require VT-1 visual examination and volumetric examination (as applicable).

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Typical conditions that would require corrective action prior to putting closure head nuts back into service include corrosion, deformed or sheared threads, deformation, and degradation mechanisms such as boric acid attack. The Code examination requirement for closure head nuts is a surface examination. Surface examination procedures are typically qualified for the detection of linear discontinuities (cracks,~ aligned pitting, and corrosion) with corresponding acceptance criteria for rejectable linear flaw lengths. Surface examination acceptance criteria are not provided for the 1989 Edition of the Code, item B6.10, as they were in the course of preparation. Without clearly defined acceptance criteria, relevant conditions that require corrective measures may not be adequately addressed.

Article IWB-3000, Acceptance Standards, IWB-3517.1, Visual Examination, VT-1, describes relevant conditions that require corrective action prior to continued service for bolting and associated nuts. IWB-3517.1 requires crack-like flaws to be compared to the l

flaw standards of IWB 3515 for acceptance. Because the VT-1 acceptance criteria l

includes the requirement for evaluation of crack-like indications and other relevant conditions requiring corrective action, such as deformed or sheared threads, localized corrosion, deformation of the part, and other degradation mechanisms, it can be concluded that the VT-1 visual examination provides a more comprehensive assessment of the condition of the closure head nut. As a result, the INEEL staff believes that VT-1 visual examination provides an acceptable level of quality and safety. In addition, it is noted that the 1989 Addenda of Section XI changes the requirement for the subject reactor pressure i

vessel closure head nuts from surface to VT-1 visual examination and provides appropriate acceptance criteria.

l Conclusion-The licensee has proposed, as an alternative to the Code required surface l

examination of reactor pressure vessel closure head nuts, to perform a VT-1 visual examination. Based on the comprehensive assessment that the VT-1 visual examination provides, and considering that the later editions of the Code require only a VT-1 visual examination on reactor pressure vessel closure head nuts, an acceptable level of quality and safety is provided. Therefore, it is recommended that the licensee's proposed I

alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

3.1.1.4 Request for Relief No.12R-18, Use of Code Case N 521, Alternative Rules for Deferral ofinspections of Nozzle to-vessel Welds, inside Radius Sections, and Nozzle-to Safe End Welds of a Pressurized Water Reactor (PWR) Vessel Note: Request for Relief 12R 18 was evaluated and authorized in an NRC SER dated January 13,1998.

l 3.1.2 Pressurizer 3.1.2.1 Request for Relief 12R-03 (Revision 2), Examination Category B-D, items B3.110 and 83.120, Pressurizer Surge Nozzle to Vessel Weld and inside Radius Section Code Requirement-Section XI, Table IWB 2500-1, Examination Category B-D, items B3.110 and B3.120, require 100% volumetric examination, as defined by Figure IWB-2500-7, for all pressurizer nozzle-to-vessel welds and nozzle inside radius sections.

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L/censee's Proposed A/temat/ve Examination-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee has determined that conformance with the Code-required volumetric examination of the pressurizer surge nozzle-to vessel weld and inside radius section would result in

. hardship or unusual difficulty without a compensating increase in quality and safety. As an alternative, the licensee proposes to perform the Code-required VT-2 examination during system pressure testing, per Category B-P, on the pressurizer each refueling outage to verify the leaktight integrity of these areas. In addition, the pressurizer surge nozzle-to-vessel weld and nozzle inside radius section will be volumetrically examined if the lower head insulation is removed for any reason.

Licensee's Basis for Proposed A/temative (as stated)-

" Pursuant to 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that the specified 1

requirements would result in hardship or unusual difficulty without a compensating i

increase in the level of quality and safety.

"The Byron Unit 1 and 2 Pressurizers include a single surge nozzle which is welded to the lower head as shown in Figure 1 *. In order to perform the code required volumetric examinations on the nozzle-to-vessel weld and the nozzle inside radius section, the outer surface of the lower vessel head must be accessible for proper surface preparation and ultrasonic scanning.

"The lower head is normally covered by a 4 inch thick, multi-layered stainless steel insulation which was not designed for removal. In order to remove the insulation, the 78 heater penetration cables shown in Figure 2' would have to be disconnected. In addition, each of the 78 convection stops, which are riveted to the insulation would have to be cut to facilitate the insulation removal per Figure 3*.

"The radiation exposure to plant personnel for the insulation removal, surface preparation, and examination is estimated to be 150 person rem, based on an area j

dose rate of 0.5 R/ hour. Even with the insulation removed, full volurnetric examination coverage of the nozzle-to-vessel weld can not be achieved. The surge nozzle geometry a

j limits ultrasonic transducer contact, and thus scanning on the nozzle side of the weld is impractical. On the vessel side of the weld, tue heater penetrations obstruct scanning such that only a small percentage of the weld volume could be captured.

"Very limited volumetric examination of the nozzle inner radius section is achievable from the outside surface of the pressurizer with the insulation removed. The blend region would not be accessible to allow for an adequate surface preparation and i

examination. A limited exam would be possible if scanning was performed from the nozzle side; however, due to the complex geometry of tne nozzle, the resulting

's coverage would provide very limited data from which to assess the condition of the inner radius.

" Volumetric examination of the nozzle-to-head weld and nozzle inner radius section is also not practical from the vesselinside surface. The inside surface is accessible only by removing the manway. The radiation exposure for the removal and reinstallation of the manway is estimated to be approximately 2 person-rem. In addition, the internal

• Figures provided by the licensee not included in this report.

13 r-

baffle plates wou'd obstruct access to the debris screen and surrounding inside surface of the nozzle, thus prohibiting a meaningful visual VT-1 examination.

" Based on the above information, the code required volumetric examination of the pressurizer nozzle-to-vessel lower head weld and associated nozzle inner radius section

.I is deerned impractical. Even partial compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety the personnel radiation hazards associated with limited data obtained by partial volumetric examination is not justified. Reasonable assurance of the continued inservice structuralintegrity of the subject welds can be achieved by performing a VT-2 examination during system pressure testing per Category B-P."

Evaluation-The Code requires 100% volumetric examination of the pressurizer surge nozzle to-vessel welds and IR sections. However, access to the vessel surface is obstructed by multi-layered, stainless steel mirror insulation that is difficult to remove. As an Elternative, the licensee proposes to perform a VT-2 visual examination during the Code-required pressure tests.

The licensee attempted to gain access to the subject examination area by cutting away and lowering the insulation at Byron, Unit 2, but this did not work (Reference 13). To gain access for examination at the Byron plant, the insulation covering the pressurizer lower head has to be complotely removed, which requires disconnection of the heater assemblies and is timo and dose intensive. Disconnection of the heater assemblies could also cause damage to the heaters at the termination points. Damage to an insulator would require total heater replacement. Based on survey results made duri%2 o &ddwood outage, the radiation exposure to remove the insulation, prepare the surface, :md perform a limited examination on the nozzles is estimated to be 150 man-rem. Therefore, compliance with the Code requirements would result in considerable hawhip for the licensee.

The subject examination areas were examined during fabrication and prior to service and found to be acceptable. These examinations ensured the structuralintegrity of the pressurizer surge nozzles prior to service. Assurance of their inservice operational readiness is provided by the Code-required pressure tests of the pressurizer and interconnected piping. Additional assurance is provided by the examination of the adjacent nozzle-to-safe end weld and other Examination Category B-D nozzles in the pressurizer, reactor pressure vessel, and steam generators. These other areas do not experience the same conditions M stresses as the pressurizer surge nozzle, but in many cases, the conditions and stresses are similar. Therefore, examination of the other Class 1 nozzles can be used as an indicator of generic degradation that could occur in the pressurizer surge nozzle and provide reasonaise assurance of the structuralintegrity of the surge nozzle.

Conclus/on-The subject examination areas are covered with insulation that wcald have to be removed to gain access for examination. Based on the burden associated with removing this insulation, the potential damage to the ceramic heater connections, the assurances provided by the examination of other Class 1 nozzles, and performance of the Code-required pressure tests, the INEEL staff concludes that cornpliance with this requirement would result in a hardship without a compensating increase in quality and safety. Therefore,it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

14

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4 1

3.1.3 Heat Exchangers and Steam Generators No relief requests, j

3.1.4 Piping Pressere Boundary i

No relief requests.

j 3.1.5 Pump Pressure Boundary l

No relief requests.

3.1.6 Valve Pressure Boundary No relief requests.

i 3.1.7 General 3

No relief requests.

I l

3.2 Class 2 Components 1

3.2.1 Pressure Vessels 3.2.1.1 Request for Relief No.12R-05 (Revision 2), Examination Category C-B, item C2.32, Residual Heat Removal Heat Exchanger (RHRHX) Nozzle-to-Shell Wold Note: Request for Relief 12R 05 (Revision 2) was evaluated and authorized.n an NRC SER dated January 13,1998.

3.2.2 Piping 3.2.2.1 Request for Relief No.12R-07, Examination Categories C-F-1 and C-F-2, items C5.11N, C5.12N, C5.51N and C5.52N, Class 2 Piping Welds Note: Request for Relief 12R-07 was evaluated and authorized in an NRC SER dated January 13,1998.

3.2.2.2 Request for Relief No.12R-08, Examination Category C-F-1 and C-F-2, items C5.12, C5.12N, C5.52 and C5.52N, Class 2 Longitudinal Piping Wolds Note: Request for Relief 12R-08 was evaluated and authorized in an NRC SER dated January 13,1998.

15 y

i 1

3.2.2.3 Request for ReIIef No.12R-09, Examination Categories C-F-1 and C-F 2, items C5.41 and C5.81, Circumferential Branch Connection Welds Note: Fiequest for Relief 12R-09 was evaluated and granted in an NRC SER dated January 13,1998.

3.2.3 Pumps 3.2.3.1 Request for Relief No.12R-06, Examination Category C-C, item C3.30, Integrally Welded Attachments to Pumps No:e: Request for Relief 12R-06 was evaluated and authorized in an NRC SER dated January 13,1998.

3.2.4 Valves No relief requests.

3.2.5 General 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 Request for Relief No.12R-10, Examination Cetegory C-H, items C7.30, C7.40, l

C7.70 and C7.80, Pressure Testing of Containment Penetration Piping Attached to Non-classed Piping j

Note: Request for Relief 12R-10 was evaluated and authorized in an NRC SER dated i

January 13,1998.

16 4

4 4

3.4.3 Class 3 System Pressure Tests i

3.4.3.1 Request for Relief No.12R 14, Examination Category D B, item D2.10, Alternate Rules for System Pressure Testing of Diesel Generator Subsystem Components s

Note: In response to the NRC RAl, Request for Relief 12R 14 was withdrawn by the l

licensee in the August 12,1996, submittal.

t l

3.4.3.2 Request for Relief No.12R 16, Use of Code Case N-498-1, Altemative Rules for l

10 Year System Hydrostatic Testing for Class 1, 2, and 3 Systems for Class 3 Systems Note: Request for Relief 12R-16 was evaluated and authorized in an NRC SER dated January 13,1998.

3.4.4 General 3.4.4.1 Request for Relief No.12R 15, Use of Code Case N4161, Altemate Testing for Class 1, Class 2, and Class 3 WeldedRepaired/ Replaced Components Note: Request for Relief 12R 15 was evaluated and authorized in an NRC SER dated January 13,1998.

3.5 General 3.5.1 Ultrasonic Examination Techniques No relief requests.

3.5.2 Exempted Com;:ensnis No relief requests.

3.5.3 Other 3.5.3.1 Request for Relief No.12R-11, Paragraph IWA 5242(a), VT-2 Visual Examination of Insulated Components l

Note: Request for Relief 12R-07 was evaluated and authorized for Class 1 systems only in I

an NRC SER dated January 13,1998.

3.5.3.2 Request for Relief 12R 12, IWA 5250(a)(2), Corrective Measures for Bolted Connections l

Note: Request for Relief 12R 12 was evaluated and authorized in an NRC SER dated July 22,1996.

l 17

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e 3.5.3.3 Request for Relief No.12R 13, IWF-5300, inservice Examination and Test Requirements for Snubbers Note: Request for Relief 12R 13 is considered part of the inservice Test (IST) Program and therefore, is not included in this evaluation. It will be evaluated by the Mechanical Engineering Branch of the NRC.

l 3.5.3.4 Request for Relief No.12R-17 (Revision 2), Use of Code Case N 509, Altemate Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments,Section XI, Division 1 Note: Request for Relief 12R-15 was evaluated and authorized in an NRC SER dated January 13,1998.

e 18

e

4. CONCLUSION Pursuant to 10 CFR 50.55a(g)(6)(i),it has been determined that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code.

I in the cases of Requests for Relief 12R-01, Revision 2 (except for Unit 2 Weld RPVC-WR29), and 12R 02, Revision 2, the licensee has demonstrated that specific Section XI requirements are impractical;it is therefore recommended that relief be granted as requested. The granting of relief will not endanger life, property, or the common defense and security and is otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. For Unit 2 RPV circumferential shell Weld RPVC-WR29, relief cannot be granted during the second 10 year interval until the augmented RPV examination required by the Regulations has been satisfied.

Pursuant to 10 CFR 50.55a(a)(3),it is concluded that for Requests for Helief 12R-03, l

Revision 2, and 12R-04, either the licensee's proposed attematives will provide an i

acceptable level of quality and safety, or Code compliance will result in hardship or unusual difficulty without a compensating increase in safety. In these cases, it is recommended 1

that the proposed alternatives contained in these requests be authorized.

Request for Relief 12R-12 was previously evatusted in an NRC SER dated July 22.

1996. Requests for Relief 12R-05, Revision 2; 12R-06; 12R-07; 12R-08; 12R-09; 12R-10; 12R-11; 12R-15; 12R 16; 12R 17, Revision 2; and 12R-18 have been evaluated in a separate i

NRC SER dated January 13,1998. Request for Relief 12R-13 will be evaluated elsewhere by the Mechanical Engineering Branch of the NRC. Request for Relief IR-14 was withdrawn by the licensee in the August 12,1996, submittal.

l This technical evaluation has not identified any practical method by which the licensee can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the Byron Nuclear Power Station, Units 1 and 2. Compliance with all of the Section XI examination requirements would necessitate radesign of a significant number of plant systems, procurement of replacement components, installation of the new components, and performance of baseline examination for these components. Even after redesign, complete compliance with the Section XI eemination requirements probably could not be achieved. Therefore,it is concluded that the public interest is not served by imposing certain provis*oas of Section XI of the ASME Code that have been determined to be impractical.

The licenses should continue to monitor the development of new or improved examination techniques. As improvements in these areas are achieved, the licenses should incorporate these techniques in the ISI program plan examination requirements.

9 Based on the review of the Byron Station, Units 1 and 2, SecondIntervalInservice Inspection Program Plan, the licensee's responses to the NRC's requests for additional information, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified.

19 l

l

0 9

5. REFERENCES 1

i 1.

Code of Federal Regulations, Title 10, Part 50.

1 2.

American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1,1989 Edition.

3.

Letter, oated August 12,1996, K. L. Kofron (Comed) to Document Control Desk, containing response to NRC's request for additionalinformation and the Byron Station, Units 1 and 2, Second IntervalInservice Inspection Program Plan, Revision 1.

4.

Byron Station Units 1 and 2 Second Interval inservice inspection Program Plan, Revision 0, February 23,1996.

5.

NUREG-0800, Stancard Review Plans, Section 5.2.4, " Reactor Coolant Boundary inservice Inspection and Testing," and Section 6.6, " Inservice Inspection of Class 2 and 3 Components," July 1981.

6.

Letter, dated June 13,1996, from George F. Dick (NRC) to D. L. Farrar (Comed),

containing request for additional information.

7.

Letter, dated October 16,1996, from George F. Dick (NRC) to D. L. Farrar (Comed), containing supplementary request for additionalinformation.

8.

Letter, dated December 16,1996 M. T. Lesniak (Comed) to Document Control Desk, containing response to NRC's supplementary request for additional information.

9.

NRC Regulatory Guide 1.147, inservice Inspection Code Case Acceptability, Revision 11, October 1994.

10.

NRC Regulatory Guide 1.14, Reactor Coolant Pump flywheellntegrity, Revision 1, dated August 1975.

Il.

NRC Regulatory Guide 1.65, Materials and Inspections for Reactor Vessel Closure Studs, October 1973.

12.

NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice andInservice Examinations. Revision 1, February 1983.

13.

Letter, dated May 3,1996, R. A. Capra (NRC) to D. L. Farrar (Comed) containing NRC SER.

20

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'A l NRC Form 335 U.S. Nuclev Repidory Camnia

1. REPORT NUMBER NPCM i102 (Amigned by NRC, Add Vol., Supp, Rev, and 320s,32 2

"#I BIBLIOGRAPHIC DATA SHEET

_ INEE1/ EXT-97-01285

2. "'TTLE AND SUBTITLE

3. DATE REPORT PUBLISHED Technical Evaluation Report on tbc Secor.J 10-year Interval Month Year Inservice Inspection Program Plan:

Commonwealta Edison Compsey, September 1998 il Byron Nuclear Power Station, Units I med 2,

4. FIN OR GRANT NUMBER Docket Numbers 50-454 and 50-d55 JCN J2229 (Task Order A16)

5. Al.TINOR(S)

6. TYPE OF REPORT e

M.T. Anderson Technical C. T. Brown E. J. Feige

7. PERIOD COVERED (Inclusive Data)

A. M. Porter R. PERFORMING ORGANI7AT10N NAME AND ADDRESS (tf NRC, provide Division.Omce or Region. U.S. Nuclear Regulatory Conunission, and mailing address; ifoontractor,;rovide na me and mailing addren)

Idaho National Engineering and Enwonmental Laboratory Mat rialsPhysics lockheed Idaho Technologies Company Idaho Falls, Idaho 83415

9. SPONSORING ORGANIZATION NAME AND ADDRESS (If NRC, typeSame as above',ifcontrador, provide NRC Division, Office or Region,U.S.

Nuclear Regulatory Comnumion, and madmg nd has)

Civil and Geosciences Branch Division ofEngmeermg Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wa=Wgtan D.C. 20555

?0. SUPPLEMENTARY NOTTS

11. ABSTRACT (200wordsorless)

This report presents the results of the evaluation of the Byron Station, Units I and 2, Second Literval Inservice Inspection Program Plan, Revision I, dated August 12,1996, including the requests for relief from ihe American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI requirements that the licensee has deterrrdned to be impractical. The Byron Station, Units I and 2, Second Interval laservice Inspection Program Plan, Revision 1 is evaluated in Section 2 of this report. The inservice ' spection (ISI) program plan is m

evaluated for (a) compliance with the appropriate edition / addenda of Section XI,(b) acceptability of examination sample,(c) correctness of the application of sysu:m or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commissico reviews. The reque.sts for relief are evaluated in Section 3 of this report.

12. KEY WORDS/DESCRIPTORS (1.ist words or phra.ia that will anist researchers in locating the repor0

13. AVAIL.ABIL.rIY STATEMENT Unlimited

14. SECURITY CLASSIFICATION a

(This page)

Unclassified

('this reparo Unclassified

15. NUMBER OF PAGES

16. PRICE

o 1

i.

e BYRON NUCLEAR POWER STATION, UNITS 1 AND 2 Page 1 of 3 Second 10-Year ISI interval TABLE 1

SUMMARY

OF RELIEF REQUESTS Relief Ucensee Proposed ReEef Request Request System or Exam item Required Method Altemative -

Status Number Componerit Category No.

Votome or Area to be Examined -

t 12R-01, RPV B-A B1.11 Circumferential shell and head Volumetric Exam Volumetric exam to the Granted, except i

lRev.2 B1.21 Welds RPVC-WR29 and PRVC-extent practical Unit 2 Weld

+

WR16 RPVC-WR29 I

12 R-02, RPV B-D B3.90 Nozzle-to-vessel Welds RPVN-A, Volumetric Exam Volumetric exam to the Granted Rev.2 RPVN-D, RPVN-D, RPVN-E and extent practical RPVN-H J2 R-03, Pressurizer B-D B 3.110 Surge nozzle-to-vessel weld and Vaksmetric None. Codc-required VT-2 Authorized Rev.2 33.120 inside radius section and volumetric exam to extent practicalif insulation removed 12R-04 RPV B-G-1 86.10 Closure Head Nuts Surface Exam VT-1 Visual Exam Authorized i

F 12 R-05, RHRHX C-B C2.32 Nozzle-to-shell welds Volumetric Exam Volumetric exam to the Authorized in NRC Rev.2 extent psactical; surface SER dated exam each period January 13,1998 l

r I2R-06 Class 2 C-C C3.30 Integral attachment we:ds Surface exam Surf ace on accessible Authorized in NRC l

Pumps portions, VT-1 on SER dated inaccessible portions of weld January 13,1998 k

2 m

o, h

e t

4 BYRON NUCLEAR POWER STATION, UNITS 1 AND 2 Page 2 of 3 Second 10-Year ISI interval TABLE 1

SUMMARY

OF RELIEF REQUESTS r

Relief Number Component Category No.

Volume or Area to be Examined Licensee Proposed Relief Request l

Request

. System or Exam item.'

. Required Method

. Altemative :

Status

[

O-12R-07 Class 2 C-F-1, C5.11N Circumferential and longitudinal -

Surface and Shift surface and volumetric Authorized in NRC Piping C-F-2 C5.12N jwelds volumetric exams exams on to examine 7.5%

SER dated l

C5.51N

• on 7.5% of welds of piping welds less than January 13.1998 C5.52N

> 3/8 nominal 3/8" but greater can 1/4*

wall thickness nominal wall thickness d

j 12R-08 Class 2 C-F-1.

C5.12 Longitudinal welds Surface and Surf ace and volumetric Authorized in NRC l

Piping C-F-2 C5.12N volumetric exam of exams on portion of SER dated i

C5.52 2.5t from circ.

longitudinal weld January 13,1998 C5.52N welds intersecting circumferential f

welds i

12 R-09 Class 2 C-F-1, C5.41 Branch connection welcs Surface exam Surface exam of saddle Granted in NRC SER Piping C-F-2 C5.81 welds dated January 13,1998 12 R-10 Class 2 C-H C7.30 Containment penetration piping System pressure Appendix J leakage test at Authorized in NRC Piping C7.40 attached to non-classed piping testing no less than peak calculated SER dated C7.70 containment pressure.

January 13,1998 C7.80 t

i2R-11 Borated IWA-5242(a)

Insulated systems borated for the VT-2 Visual exam VT-2 Visual exam with 4-Authorized for Systems purpose of controlling reactivity with insulation hour hold time with Class 1 system only 7

removed insulation in place and direct in NRC SER dated j

viscal with insulation January 13,1998 removed dunne shutdown i

i

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V i

i BYRON NUCLEAR POWER STATION, UNITS 1 AND 2 Page 3 of 3 Second 10-Year ISI interval TABLE 1

SUMMARY

OF RELIEF REQUESTS Relief Request System or Exam item Licensee Proposed Relief Request Number Component Category No.

Volume or Area to be Examined Required Method Altemative Status i

12R-12 Bolted IWA-5250(a)

Authorized in NRC Connections SER dated July 22,1996 i

12R-13 Snubbers IWF-53OO Evaluated elsewhere 12R-14 Diesel D-B D2,10 Withdrawn l

Generator 12R-15 Class 1,2, IWA-4700 Repair / replacement of pressure System Code Case N-416-1 Authorized in NRC and 3 retaining boundary Hydrostatic Test SER dated i

Systems January 13,1998 12R-16 Class 1, 2, Pressure retaining systems System Code Case N-498-1 Authorized in NRC i

and 3 Hydrostatic Test SER dated Systems January 13,1998 i

12R-17, Class 1,2, integrally welded attachments Surface or Code Case N-509 with Authorized in NRC Rev.2 and 3 volumetric exam minimum 10% examination SER dated Systems sample January 13,1998 12R-18 RPV B-D B3.90 Nozzle-to-vessel welds and inside 25-50% of the Code Case N-521 Authorized in NRC B3.100 radius sections nozzles examined SER dated i

B-F B5.10 by the end of the January 13,1998 first period

-