ML20198Q282

Safety Evaluation Granting Relief for Second 10-yr Interval ISI Program Plan & Associated Requests for Plant

ML20198Q282
Person / Time
Site:	Wolf Creek
Issue date:	10/24/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198Q266	List: L-96-026, Tech Evaluation Rept on Second 10 Year Inservice Insp Program Plan:Wolf Creek Nuclear Operating Corp ML20198Q262 ML20198Q282
References
NUDOCS 9711120162
Download: ML20198Q282 (20)

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't NUCLEAR REZULATORY COMMISSION WASHINGTON, D.C. *=8 "1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SECOND TEN YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR REllEF WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50 482

1.0 INTRODUCTION

The Technical Specifications (TS) for the Wolf Creek Generating Station state that the inservice inspection 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 (B&PV) Code and applicable l addenda as required by 10 CFR 50.55a. The Code of Federal Regulations, l 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph l (g) may be used, when authorized by the NRC, if (1) the proposed alternatives would arovide an acceptable level of quality and safety or (ii) compliance l with tie specified requirements would result in hardship or unusual difficulty l 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 I

(including supports) shall meet the requirements, except the design and access provisions and the )re 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 ten-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) twelve 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 Wolf Creek Generating Station second ten-year inservice inspection (ISI) interval is the 1989 Edition.

Pursuant to 10 CFR 50,55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME

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Code requirement. After ' <aluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(1) 1 Commission may grant relief and may impose alternative re .nat are determined to be authorized by law, will not endanger life,quiremr proper 6, or the comon defense and security, and are otherwise in the public interest, 'iving due consideration to the burden upon the licensee that could rest; , if the requirements were imposed.

In a letter dated August 30, 1995, Wolf Creek, Nuclear Operating Corporation (licensee), submitted to the NRC its second ten year inservice inspection l interval program plan and associated requests for relief for the Wolf Creek Generating Station. Additional information was provided by the licensee in its letters dated September 20, 1995, November 17, 1995, May 3, 1996.

December 9, 1996, January 23, 1997, and May 6, 1997, i 2.0 EVALUATION The staff, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its second ten-year inservice inspection program plan and associated requests for relief for the Wolf Creek Generating' Station. Based on the information submitted, the staff adopts the contractor s conclusions and recommendations ) resented in the attached Technical Evaluation Report (TER), as modified :elow.

Request for Relief 12R 03: Section XI. Table IWB-2500-1. Examination Category B-A, Item Bl.30, Reactor Vessel Shell-to Flange Weld of the Code requires 100%

volumetric examination of the reactor vessel shell-to-flange weld as defined in Figure IWB 2500-4. Pursuant to 10 CFR 50,55a(g)(5)(111), the licensee requested relief from 100% volumetric examination of the reactor vessel shell-to flange weld.

The Code requires 100% volumetric examination of the reactor vessel shell-to-flange weld. However. he flange taper precludes scanning for 100% coverage of the required examination area, thus making the Code-required volumetric examination to the extent required by the Code impractical. To obtain com)lete volumetric coverage, modifications or replacement of the component wit 1 one of a design providing for complete coverage would be required.

imposition of this requirement would cause a considerable burden on the licensee.

The licensee provided the following information in their response to the NRC's request for additional information.

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Automated Examination Coverage  ;

Clockwise Counterclockwise Scan Angle Direction Direction _

Breakdown for Parallel Scans 45' 51.3% 51,3%

60* 51.3% 51.3%

j 70' 65.2% 65.2%

Breakdown for Perpendicular Scans 3 45' 48.9% 64.5%

60' 76.8% 42.4%

70' 19.1% 78%

" Manual examination from the flange face was 90% complete for the first interval."

l The licensee proposed no additional examinations. However, based on the i percentage of volumetric coverage that has been obtained from the shell side

and the 90% coverage obtained from the flange face during the first interval as noted in the table above, it is reasonable to conclude that during the second interval inspections, degradation, if present, will be detected.

Therefore, reasor,able assurance of continued structural integrity is provided, The flange taaer makes the Coae-required volumetric examination to the extent required by t1e Code impractical. The staff concludes-that based on the significant amount of weld coverage obtained, reasonable assurance of

structural integrity is provided. Therefore, relief is granted pursuant to

10 CFR 50.55a(g)(6)(1).

Request for Relief 12R 04: Section XI, Table IWB-2500-1. Examination Category B A Item Bl.21. Reactor Vessel Closure Head Circumferential Weld of the Code

requires 100% volumetric examination of reactor vessel closure head

- circumferential weld; : defined in Figure IWB-2500-3.

l Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from 100%

volumetric examination of reactor vessel closure head weld CH-103-101. The Code requires volumetric examination of the reactor vessel closure head weld.

4 However, the location of the cooling duct ring and three lifting lugs restrict scanning making the Code-required 100% volumetric examination impractical. To

obtain complete volumetric coverage, modifications or replacement of the k'

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component with one of a design providing for complete coverage would be j required. Im 1

the licensee. position of this requirement would cause a considerable burden on The licensee is performing the Code examination to the maximum extent 3racticable. Based on the significant percent of volumetric coverage that has

>een obtained it is reasonable to conclude that degradation. if present, would be detected. Therefore, reasonable assurance of continued structural

integrity is provided. The staff concludes that component configuration makes the Code required volumetric examination impractical. Furthermore, based on the significant percent of weld volume that was examined. reasonable assurance

of structural integrity is provided, Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(1).

Request for Rel'ef 12R 10: Section XI. Table IWB 2500-1. Examination Categor B-G-1. Item B6.10. Reactor Vessel Closure Head Nuts requires of the Code 100%y i

surface examination of the reactor vessel closure head nuts. Pursuant to 4

10 CFR 50.55a(a)(3)(1). the licensee proposed an alternative to performing 4

100% surface examination of the reactor vessel closure head nuts. The

! licensee proposed as an alternative examination to perform a VT-1 visual examination of the surface of all reactor closure head nuts, utilizing the 4 acceptance criteria of IWB-3517. as delineated in the 1989 Edition of ASME

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Section XI. All Items in Examination Category B-G-1 except the reactor pressure vessel closure head nuts and the closure studs (when removed) require VT-1 visual examinations and/or volumetric examination (as applicable).

Typical conditions that would require corrective action prior to putting closure head nuts back into service would include corrosion, deformed or sheared threads. deformation, and degradation (i.e.. boric acid attack). The Code requires a surface examination for closure head nuts. Surface examination procedures are typically qualified for the detection of linear flaws (cracks) and have acceptance criteria specifying only rejectable linear flaw lengths. Acceptance criteria are not provided in the 1989 Edition of the Code. Item B6.10. as they were in the course of preparation when the Code was published. Without clearly defined acceptance criteria, conditions that require corrective measures may not be adequately addressed. The 1989 Addenda of Section XI addresses these problems by changing the requirement for the subject reactor pressure vessel closure head nuts from surface to VT-1 visual examination and providing appropriate acceptance criteria.

Article IWB-3000. Acceptance Standards. IWB-3517.1 Visual Examination. VT-1.

-describes conditions that require corrective action prior to continued service for bolting and associated nuts. One of these requirements is to compare crack-like flaws to the flaw standards of IWB-3515 for acceptance. Because the VT-1 visual examination acceptance criteria include evaluation of crack-like indications and other conditions requiring corrective action, such as deformed or sheared threads localized corrosion, deformation of 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 e - 4= s .aw+e-.n~_.. ,--w,. . . , . v -*4 mae m,--, w o e,m ._a

closure head nut. As a result the staff determined that a VT-1 visual examination provides an acceptable level of quality and safety.

The staff concludes that based on the comprehens1ve assessment that the VT-1 visual examination provides, and considering that the 1989 Addenda and 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. Therefore, the licensee's proposed alternative. VT 1 visual examination is authorized pursuant to 10 CFR 50.55a(a)(3)(1).

Request for Relief 12R 06: Section XI. Table IWB-2500-1. Examination Categor B D. Item B3.110. Pressurizer Nozzle-to Vessel Weld of the Code requires 100%y

, volumetric examination of pressurizer nozzle-to-vessel welds as defined in l Figure IWB 2500 7. Pursuant to 10 CFR SC 55a(g)(5)(iii), the licensee requested relief from performing 100% volumetric examination of pressurizer nozzle to vessel Welds TBB0310B-C-W and TBB0310B D W.

The Code requires volumetric examination of 3ressurizer nozzle to vessel welds. However, the tapers of the weld and 3ase metal preclude achieving 100%

coverage of nozzle-to-vessel welds TBB03 10B C W and TBB03-10B D W. These tapers cause the transducer to " lift off". making the Code-required 100%

volumetric examination impractical. To obtain complete volumetric coverage, modifications or replacement of the component with one of a design providing for complete coverage would be required. Im cause a considerable burden on the licensee. position of this requirement would The licensee is performing the Code examination to the maximum extent practicable. Based on the significant amount of volumetric coverage that has been obtained (= 85% of each scan) it is reasonable to conclude that degradation, if present, will be detected. Therefore, reasonable assurance of continued inservice structural integrity is provided.

The staff concludes that the component configuration restriction makes tne Code required volumetric examination impractical. Furthermore, based on the significant amount of weld coverage obtainable, reasonable assurance of continued structural integrity is provided. Therefore, relief is granted, pursuant to 10 CFR 50.55a(g)(6)(1).

Request for Relief 12R 07: Section XI. Table IWB-2500-1. Examination Category B-D Item B3.110. Pressurizer Surge Nozzle-to-Bottom Head Weld of the Code requires 100% volumetric examination of all pressurizer nozzle-to bottom head welds as defined in Figure IWB-2500-7(b). Pursuant to 10 CFR 50.55a(g)(5)(111), relief is requested on the basis that compliance with the specified requirements is impractical. Conformance with the inservice inspection requirements would necessitate a design modification to the pressurizer surge nozzle and pressurizer heater to remove obstructions that preclude 100% examination coverage of the subject weld.

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s The Code requires volumet?ic examination of the pressurizer nozzle to bottom head weld. However, the geometric configuration of the nozzle to head area 4 precludes 100% coverage of nozzle to bottom head Weld TBB03 10A-W and makes the Code required volumetric examination impractical. To obtain corplete volumetric coverage, modifications or replacement of the component with one of a design providing for complete coverage would be required. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee is performing the Code examination to the maximum extent possible. The weld and adjacent base metal on the head side were examined in three of the four Code-required directions. This represents a significant portion of the Code required examination volume, and-therefore degradation. if present, should be detected. Therefore, reasonable assurance of continued structural integrity is provided.

The staff concluded that the component configuration restriction makes the Code required volumetric examination imaractical. Based on the extent of volumetric coverage achievable, reasonaale assurance of structural integrity is provided. Therefore, relief is granted, pursuant to 10 CFR 50.55a(g)(6)(1).

Request for Relief 12R 08: Section XI. Table IWB-2500 1. Examination Category l

I B D. Item B3.120. Pressurizer Surge Nozzle Inner Radius Section of tne Code requires 100% volumetric examination of all sections as defined in Figure IWB 2500 7(b). pressurizer Pursuant tonozzle 10 CFRinner radius 50.65a(g)(5)(111), the licensee requested relief from 100% volumetric examination of pressurizer surge nozzle inner radius section TBB03-10A-IR.

The Code requires volumetric examination of the pressurizer nozzle inside radius sections. However, volumetric examination of the surge nozzle inner radius is restricted by heater penetrations and the nozzle's geometric configuration. These obstructions make the volumetric examination impractical to perform to the extent required by the Code. To meet the Code recuirements.

the surge nozzle and adjacent obstructions would have to be modifiec to allow access for examination. Imposition of this requirement would create a considerable burden on the licensee.

Approximately 30% of the pressurizer surge nozzle inside radius section can be examined. In addition, other pressurizer nozzle inner radius sections are receiving com)lete volumetric examination. Therefore, significant patterns of degradation s1ould be detected by the examinations that are being performed, and reasonable assurance of the structural integrity is provided.

The staff concluded that the pressurizer heater penetrations and the nozzle configuration design make the Code-required 100% volumetric examination of the subject pressurizer surge nozzle inner radius section impractical to complete.

Furthermore, based on the portion of the surge nozzle inner radius that can be examined, in conjunction with the Code examination of other pressurizer nozzle inner radius sections, reasonable assurance of structural integrity is provided. Therefore, relief is granted, pursuant to 10 CFR 50.55a(g)(6)(1).

Request fo Relief 12R 09: Section XI. Table IWB 25001. Examination Category B-F. Itet 85.40. Pressurizer Dissimilar Metal Nozzle to Safe End Welds of the Code requires 100% volumetric and surface examination of the pressurizer dissimilar metal nozzle to-safe end welds as defined in Figure IWB-2500-8. --

Pursuant to 10 CFR 50.55a(a)(3)(11), the licensee proposed an alternative in lieu of the Code requirements of performing 100% volumetric examination of the =

following pressurizer nozzle-to-safe end welds, because the specified Code. L Section XI requirements would result in hardship or unumal difficulty without a compensating increase in the level of quality and safety.

Weld Number Weld Description Percent Not Examined TBB03-4 W Relief Nozzle-to-Safe 20% - 60' axial scan End 45% - 45' axial scan _

TBB03-3 A-W Safety Nozzle-to Safe 50% - 60' axial scan l

End 3% - 45' axial scan i TBB03-1-W Surge Nozzle-to-Safe 15% - 60* axial scan End 40% 45' axial scan TBB03-3-B W Safety Nozzle-to-Safe 55% 60' axial scan [ -

End 40% - 45* axial scan TBB03-2-W Spray Nozzle-to Safe 10% - 60' axial scan End 40% - 45* axial scan TBB03-3-C-W Safety Nozzle-to-Safe 20% - 60 axial scan End 40% - 45 axial scan The Code requires volumetric and surface examination of the pressurizer dissimilar metal nozzle-to safe end welds. However, the configuration of the subject nozzle-to-safe end welds precludes achieving 100% volumetric examination.

The licensee anticipates that the improved examination techniques that will be used during the second interval will improve coverage of the subject welds.

Therefore, the staff recommends that the licensee resubmit for relief (if necessary) after performing these examinations. Further, the staff recommends that the licensee review NRC Information Notice No. 90-30. " Ultrasonic Inspection Techniques for Dissimilar Metal Welds" for guidance.

The staff concludes that since increased coverage will be obtained this interval due to enhanced examination techniques, the licensee's alternative is denied at this time. The licensee should resubmit after completion of these examinations if complete coverage (>90%) is not obtained and determine whether the requirement is impractical under 10 CFR 50.55a(g)(5)(iii).

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.- . j B-Request for Relief 12R 05: Section XI. Table IWB 2500 1. Examination Category B B Item B2.40, Steam Generator Tubesheet to Channel Head Weld of the Code requires 100% volumetric examination of Oe steam generator tubesheet to channel head weld as defined in Figura IWB 2500 6. Pursuant to 10 CFR 50.55a(g)(5)(iii). the licensee requested relief from 100% volumetric examination of steam generator tubesheet-to channel head weld EBB 010 SEAM 1 W.

The Code recuires volumetric examination of the steam generacor tubesheet-to-channel heac weld. However, the obstruction caused by the four support legs and steam generator design geometry make the Code-required 100% volumetric examination of tubesheet-to channel head weld EBB 010 SEAM 1 W. impractical to complete. To perform a complete volumetric examination. modifications or replacement of the component with one of a design providing for complete coverage would be required. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee is performing the Code examination to the maximum extent practicable. Based on the significant amount of volumetric coverage that is obtainable, it is reasonable to conclude that a pattern of degradation if present, will be detected. Therefore, reasonable assurance of continued structural integrity is provided.

The staff concludes that the component configuration and su make the Code required volumetric examination impractical p) ort restrictions r urthermore, based on the significant amount of weld coverage obtainable, reasonable assurance of structural integrity is provided. Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(1).

Recuest for Relief 12R.18: Section-XI, Table IWB 2500-1, Examination Category Bs. Item B9.31. Reactor Coolant System Branch Connection Welds of the Code '

requires 100% volumetric and surface examaation of piping branch connection welds as defined in Figure IWB 2500 9. Pursuar.t to 10 CFR.50.55a( )(5)(111),

the licensee requested relief- from 100% volumetric examination of iping branch connection Welds BB 01-S101-7. BB-01-S302-3, and BB 01-S402 3.

The Code requires volumetric and surface examinations of the subject piping branch connection welds. However, the configuration and metallurgical properties preclude achieving 100% coverage of pipe branch connection welds

-BB 01-S101-7, BB 01-S302-3. and BB-01-S402-3.

The pipe branch connection geometry makes the Code-required 100% volumetric examination impractical. To obtain complete volumetric coverage, modification or. replacement of the branch connections with connections of a design providing for complete coverage would be required. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee is performing the Code examinttion to the maximum extent practicable. Extended beam paths were considered and found to be impractical due to the metallurg4.al properties of centrifuga11y cast stainless steel.

However, based on 1' !gnificant amount of the volume that will be examined (75%). it is rease .e to conclude that degradation. if present, will be detected. Therefore, reasonable assurance of continued structural integrity will be provided.

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• 9 The staff concludes that the component configuration restriction makes the Code-required volumetric examination impractical. Based on the significant amount of weld coverage obtainable, reasonable assurance of structural integri y is Therefore, relief is granted pursuant to 10 CFR 50,55a( )(6)(provided.

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Request for Relief 12R 14: Code Case N 491. Examination Category F-A, Item F1,40. Examination of Reactor Vessel Supports requires 100* VT-3 visual examination of all Class 1, 2, and 3 supports other than piping supports.

Code Case H 491 is approved in Regulatory Guide 1,147. Revision 10 for general use by licensees.

Pursuant to 10 CFR 50.55a(g)(5)(iii), the 11cen & e requested relief from aerforming 100% of the Code required VT 3 visual examinations contained in Code Case N 491 for the reactor vessel supports. Code Case N 491. Table 2500 1. Examination Category F-A. Item F1.40 requires 100% VT-3 visual examination of the reactor vessel supports. However, due to access restrictions, the support design, and high local radiation levels, the licensee proposed to )erform a limited visual examination. The support assembly at each of tie reactor vessel nozzles consists of nozzle weld '

buildup, shoe plate, air cooled box, and steel support structure. The steel support structure is embedded in the primary shield wall, making the VT-3 visual examination of the reactor vessel support impractical to perform to the extent required by the Code. The licar..;ee's proposed alternative, to perform a VT-3 visual examination of the entire nozzle weld buildup and shoe plate and approximately 30% of the air-cooled box, will provide reasonable assurance of cnntinued structural integrity.

The staff concludes that the design of the reactor vessel supports makes it impractical to perform the VT-3 visusl examination to the extent required by Code. The licensee's alternative will provide reasonable assurance of structural integrity: therefore, relief is granted and the alternative imposed pursuant to 10 CFR 50.55a(g)(6)(1).

Request for Relief 12R 12: Section XI. Table IWB 2500-1. Examination Category C F 1, items C5.12. C5.22, and C5.42, Class 2 Longitudinal Piping Welds in Austenitic Piping Systems, requires that surface and volumetric examination of Class 2 longitudinal piping welds. Item C5.42 requires surface examination of Class 2 longitudinal piping welds. The examination volume / surface area includes 2,5t at the intersection with circumferential welds required to be examtned. Pursuant to 10 CFR 50.55a(a)(3)(1), the licensee proposed an alternative to performing 100% of the Code-required surface and volumetric examinations on Class 2 longitudinal piping welds in austenitic piping systems to the extent required by Code.

The licensee requested relief from performing the surface and volumetric examinations, to the extent required by Code, of the longitudinal welds in Class 2 piping. The licensee proposes to examine the potentially critical portions of the longitudinal welds (the portions that intersect circumferential welds) in conjunction with examination of the circumferential welds. The licensee's alternative is based on the position that, due to

fabrication controls and lack of susceptibility to the conditions that lead to failure, longitudinal welds are unlikely to fail. The potentially critical portions of the longitudinal welds are the portions that intersect circumferential welds these regions will be examined in conjunction with the circumferential welds. The staff determined that the licensee's proposed alternative examines tN most critical area of the longitudinal weld, the overall cuality of longitudinal welds, and the extent of examinations ,

performec provides itn acceptable level of quality and safety.

The staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50,55a(a)(3)(1).

Request for Relief I2R 13: Section XI. Table IWC-5250-1. Examination Category C H, Items C7.30. C7.40. C7.70, and C7.80, Pressure Testing of Class 2 Components at Containment Penetrations requires system leakage testing of Class 2 piping and valves once each inspection period.

Furthermore. IWC-5210(b) requires that where air or gas is used as a testing medium, the test procedure shall include methods for detection and location of through-wall leaks in system components. Because an Appendix J. Type C, pneumatic test uses air versus gas as a testing medium, the licensee's test 3rocedure should meet the above requirement for the CIV's and pipe segments

)etween the CIVs.

Pursuant to 10 CFR 50,55a(a)(3)(1), the licensee proposed an alternative to performing the Code-required system leakage test for the following Code Class 2 piping and valves at containment penetrations where the balance of the system is outside the scope of Section XI (non-ASME Class). Ir. addition, the licensee's proposed alternative also includes the provision that where air or gas is used as a testing medium, the test 3rocedure shall include methods for

- detection and location of through-wall leats in system components and, when Section XI pressure testing credit is to be taken, Ty)e C tests, the peak calculated containment pressure as defined by the Tec1nical Specifications, will be used as the test pressure.

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-D-Table 1 Line Number Penetration Description BB-103-HCB-1" P-62 H-12BB02 BL-028-HCB-3" P-25 .M-12BL01 BM-053-HBB-3" P-78 M-12BM01 EC-067-HCB 6" P-53 M-12EC02 EC-072-HCB-6" P-54 H-12EC02 EC-081-HCB 3" P-55 H-12EC02 EM-071-BCB-3/4" P-92 M-12EM01 GP-003-HBB-1" P-51 H-12GP01 .

GP-005-HBB-1" P-51 M-12GP01 GS-025-HBB-6" P-65 M-12GS01 GT-007-HBB-36" V-160 M-12GT01 '

CT-004-HBB-36" V-161 M-123T01 GT-029-HBB-18" V-161 M-12GT01 GT-034-HBB-18" V-160 M-12GT01 GT-033-HBB-18" V-160 M-12GT01 GT-030-HBB-18" V-161 M-12GT01 HB-015-HCB-3" P-26 M-12HB01 HB-025-HBB-3/4" P-44 M-12HB01 HD-015 HBB-2" P-43 M-12HD01 KA-244-HCB-14" P-30= M-12KA01 KA-259-HCB-18" P M-12KA01 KA-051-HBB-4" P M-12KA02 KA-261-HBB-1" P-63 H-12KA02' KA-732-HBB-1" N/A- H-12KA05 KA-733-HBB-1" N/A M-12KA05 KB-001-HCB-2" P-98 M-12KA05 KC-560-HBB-4" P-67 M-12KC02

'LF-842-HCB-5" P-32 M-12LF09

- SJ-002-BCB-1" P-64 M-12SJ01

- SJ-003-ECB-1" P-95 M-12SJ01

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The licensee has proposed, as an alternative to Code requirements to perform 10 CFR Part 50. Appendix J 1eakage. Option B. Performance-Based tests on the primary rentor containment penetration lines listed in the table above and on their associated valves as required by 10 CFR Part 50. etc. WCGS Technical Specification 3/4.6. Appendix J. Option B. Performance-Based recuiremer.ts ensure that (a) leakage through these containments or systems anc components penetrating these containments does not exceed allowable leakage rates specified in the Technical Specification: and (b) integrity of the containment

- structure is maintained during its service life. Option B of Appendix J ,

identifies the performance-based requirements cnd criteria for preoperational and subsequent periodic leakage-rate testing. Specific guidance concerning a 3erformance-based leakage-rate test methods, procedures, and analyses that may L >e used to implement these requirements and criteria are provided in l Regulatory Guide 1.163. " Performance Based Containment Leak-Test Program."

dated September 1995, in addition, the licensee will also perform the Appendix J leak test at peak calculated containment design pressure and including'a method for detection and locction of leakage.

The Code system leakage test required by Examination Category C-H provides periodic verification of the leak-tight integrity of Class 2 piping systems or segments once every 40 months. Pipe segments from non-Code class systems that penetrate containment are designed and examined as Class 2 piping for the sole purpose of verifying the -leak-tight integrity of containment. The Appendix J pressure testing provides periodic verification of the leak-tight integrity of the primary reactor containment, and of systems and components that penetrate containment: and provides assurance that the containment pressure boundary is being maintained at an acceptable level while monitoring for deterioration of seals.- valves, and piping. Use of Appendix J. Option B results in tests being performed at intervals not exceeding 60 months versus 40 months as required by the Code. The Staff has determined that these containment testing frequencies are acceptable. therefore they should also be considered acceptable for the subject piping. Each of the Code Class 2 lines listed in the above table are tested during an Appendix J leakage test at a pressure of 48.1 psi versus the design basis loss o coolant accident peak calculated containment internal pressure of 47.3 psig per TS 3/4.6 and TS 6.8.4.1.

The system leakage test required by Examination Category C-H provides periodic verification of the leak-tight integrity of Class 2 piping systems or segments once every 40 months. Pipe segments from non-Code class systems that penetrate containment are designed and examined as Class 2 pipe to protect the integrity of containment. The Appendix J pressure testing provides periodic

. verification of the leak-tight integrity of the primary reactor containment, and of. systems and components that penetrate containment. The 10 CFR Part 50.

Appendix J. Option B test frec;aency provides assurance that the containment

. pressure boundary is being maintained at an acceptable level while monitoring for deterioration of seals, valves, and piping.

-The Class 2 containment isolation valves (CIVs) and connecting pipe segments must withstand the peak calculated containment internal pressure related to the maximum design containment pressure. The containment penetration piping is classified as Class 2 because it is part of the containment pressure

boundary, and because containment integrity is the only safety-related function performed by this piping. Therefore, it is logical to test the penetration )iping portion of the associated system to the Appendix J criteria. T1e staff finds that the pressure-retaining integrity of the CIVs and connecting piping, and their associated safety functions, may be verified with an Appendix J. Type C test if it is conducted at the peak calculated containment pressure.

IWC-5210(b) requires that where air or gas is used as a testing medium, the test procedure shall include methods for detection and location of through-wall leaks in system components. Because ar Apaendix J. Type C, pneumatic test uses air versus gas as a testing medium, t1e licensee's test procedure should meet the above requirement for the CIVs and pipe segments setween the CIVs. The licensee's proposed alternative includes both performing the Appendix J leak test at peak calculated containment design pressure and includes a method for detection and location of leakage and where air or gas is used as a testing medium, and the test )rocedure will include methods for detection and location of through-wall leats in system components.

The staff concludes that compliance with Appendix J provides an acceptable level of quality and safety for the subject Class 2 piping that penetrates containment, where the balance of the piping system is non-Code class.

Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(1).

Request for Relief 12R 17: Section XI. Table IWB-2500-1. Examination Category D-B, Item D2.10 requires system leakage testing of Class 3 piping and valves once each inspection period and hydrostatic testing once each interval.

Pursuant to 10 CFR 50.55a(a)(3)(1), the licensee proposed an alternative to performing the Code-required system leakage test for the Code Class 3 diesel generator subsystems.

The Code requires system leakage testing of Class 3 piping and valves once each inspection period. In-lieu of the Code-required testing, the licensee has proposed to perform leak or flow testing each month on the emergency diesel generator subsystems. Performing leak and/or flow testing 3rovides an indirect verification of system integrity. Each system receives tiese tests every thirty days, which is a more rigorous testing schedule than that required by the Code. Each test has aressure or flow indicators that are monitored during the test. Each of t1ese indicators has an associated maximum / minimum allowable value, which will alert an operator if a leak exists. The cause of the unallowable indication is then located and repaired if necessary.

The staff believes that the licensee *s proposed alternative to the Code-recuired periodic pressure tests will provide an acceptable level of quality anc safety.

The staff concludes that the licensee's proposed alternative will detect leakage, if 3 resent, providing an acceptable level of quality and safety.

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

Request for Relief 12R 02: Section XI. IWA-5250(a)(2). System Pressure Test Corrective Measures Code Requirement-lWA-5250(a)(2) states that if leakage occurs at a bolted connection during a system pressure test. then all bolting must be removed end a VT-3 visual examination performed to detect corrosion.

Pursuant to 10 CFR 50.55a(a)(3)(1) the licensee proposed an alternative to performing the Code-required removal and VT-3 visual examination of bolting if leakage occurs during a system pressure test of Class 1, 2 and 3 systems.

In accordance with the 1989 Edition of the Code, when leakage occurs ~at bolted connections, all bolting is required to be removed for VT-3 visual 3 examination. In lieu of the Code-required removal of bolting to perform a VT-3 visual examination, the licensee has proposed to perform an evaluation of the bolted connection to determine the susceptibility of the bolting to corrosion and the potential for failure as noted below:

1. Bolted materials

2. Corrosiveness of the process fluid

3. Leakage location

4. Leakage history or other system components

5. Visual. evidence of corrosion at connection (connection assemblies)

6. Service age of bolting materials.

This alternative allows the licensee to utilize a systematic approach and sound engineering judgment: provided, as a minimum, all six evaluation factors listed above, are considered. Furthermore, if the initial evaluation

-indicates the need for a more in-depth evaluation, the bolt closest to the source of leakage shall be removed. VT-3 examined, and evaluated in accordance with IWA-3100(a) by the license 2. With this provisiu . the licensee's alternative to the Code-required removal of bolting at a joint when leakage occurs will provide an acceptable level of quality and safety, as the integrity of the joint will be maintained.

The staff concludes that the licensee's proposed alternative, to use a systematic approach and sound engineering judgement, will provide an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10-CFR 50.55a(a)(3)(1), considering that if the initial evaluation indicates the need for a more in depth evaluation the bolt closest to the source of leakage is removed. VT-3 examined, and evaluated in accordance with IWA-3100(a).

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-Request for Relief 12R 01: Section XI. IWA-5242(a). System Pressure Tests for Insulated Com)onents in Class 1 Borated Systems states that for systems borated for tie purpose of controlling reactivity insulation shall be removed from pressure-retaining bolted connections for a direct VT-2 visual examination. Pursuant to 10 CFR 50.55a(a)(3)(1), the licensee proposed an alternative to the Code-required removal of insulation for VT-2 visual examinations of bolted connections in Class 1 borated systems.

Paragraph IWA-5242(a) requires the removal of all insulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. The licensee's alternative includes a 4-hour hold time at test conditions (pressure and temperature) prior to the VT-2 visual examination

• without removal of insulation, and during the refueling outage, the connection will not be required to be at operating conditions for performance of the VT-2 visual examination with the insulation removed. Any evidence of leakage will be evaluated in accordance with IWA-5250. Requiring removal and installation of insulation, and disassembly of scaffolding at Class 1 bolted connections when the plant is at operating conditions for the sole purpose of the VT-2 visual examination is a >ersonnel safety hazard. The staff determined that based on the review of tie licensee's basis for relief and proposed alternative, the licensee's approach to the Code-required insulation removal is acceptable because any degradation due to boric acid at a bolted connection would be detected.

The staff concludes that compliance would result in hardship without a comeensating increase in the level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

Request for Relief I2R 19: The licensee pursuant to 10 CFR 50.55a(a)(3)(1) requested authorization to use ASME Code Case N-498-1. Alternative Rules for 10-Year System Hydrostatic Testing for Class 1. 2. and 3 Systas.Section XI.

Division 1 in lieu of the Code of the Code-required hydrostatic testing requirements.

The system hydrostatic test, as sti)ulated in Section XI. is not a test of the structural integrity of the system aut rather an enhanced leakage test.2 Hydrostatic testing only subjects the piping components to a small increase in pressure over the design pressure: therefore, piping dead weight, thermal expansion, and seismic loads present far greater challenges to the structural integrity of a system. Consequently, the Section XI hydrostatic pressure test is primarily regarded as a means to enhance leak detection during the examination of components under pressure, rather than as a method to determine the structural integrity of the components. In addition, industry experience indicates that leaks are not being discovered as a result of hydrostatic test pressures causing a preexisting flaw to propagate through the wall, in most S. H. Bush and R. R. Haccary. Development of In-Service Inspection Safety Philosophy for U.S.A. Nuclear Pcver Plants. ASHE.1971.

pressures causing a preexisting flaw to propagate through the wall in most cases leaks are being found when the system is at normal operating pressure.

Code Case N-498. Alternative Rules for 10-Year System Hydrostatic Testing for Class 1 and 2 Systems, was previously approved for general use on Class 1 and 2 systems in Regulatory Guide 1.147. Rev. 9. For Class 3 systems. Revision N-499-1 specifies requirements identical to those for Class 2 components (requirements for Class 1 and 2 systems in N-498-1 are unchanged from N 498).

In lieu of hydrostatic pressure testing at or near the end of the 10-year interval. Code Case N 498-1 requires a VT-2 visual examination at nominal operating pressure in conjunction with a system leakage test performed in accordance with paragraph IWA-5000 of the 1992 Edition of Section XI.

Class 3 systems do not normally receive the amount and/or type of nondestructive examinations that Class 1 and 2 systems receive. While Class 1 and 2 system failures are relatively uncommon. Class 3 leaks occur more frequently and are caused by different failure mechanisms. Based on a review of Class 3 system failures requiring repair during the last 5 years.2 the most common causes of failures are erosion-corrosion (EC), microbiologically-induced corrosion (HIC), and general corrosion. In general, licensees have implemented programs for the prevention detection, and evaluation of EC and HIC: therefore Class 3 systems receive inspection commensurate with their functions and expected failure mechanisms.

System hydrostatic testing entails considerable time, radiation dose, and financial resources. The safety assurance provided by the enhanced leakage gained from a slight increase in system pressure during a hydrostatic test may be offset or negated by the necessity to gag or remove Code safety ano/or relief valves (placing the system, and thus the plant, in an off-normal state), erect temporary supports in steam lines, and expend resources to set up testing with special equipment and gages. Therefore, performance of system hydrostatic testing at 3ressures greater than operating pressure represents a considerable burden witlout a compensating increase in quality and safety.

The staff concludes that, considering the minimal amount of increased assurance provided by the elevated pressure of a hydrostatic test versus the pressure for the system leakage test, tne excessive burden associated with performing the hydrostatic test, and that Code Case N-498-1 provides reasonable assurance of operational-readiness of the subject systems, the staff finds that compliance with the Section XI hydrostatic testing

- requirements results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, the use of Code Case N-498-1 for Code Class 1. 2. and 3 systems is authorized for the second interval pursuant to 10 CFR 50.55a(a)(3)(ii), until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

' Documented in Licensee Event Reports and the Nuclear Plant Reliability Data System databases.

. o Request for Relief 12R 16: The Code Requirement-IWA-2311(b) Qualification of Nondestructive Examination Personnel for Ultrasonic Examination requires that the training, qualification, and certification of ultrasonic examination personnel comply with the requirements specified in Appendix VII. The licensee requested relief from implementation of Appendix VII until the performance demonstration requirements of Appendix VIII are fully implemented.

Appendix VII was incorporated into the Code in 1988 to enhance the qualifications of ultrasonic examiners. This appendix ) laces controls on a wide variety of class oom and laboratory training. Altlough Appendices VII and VIII are both designed to improve flaw detection confidence, their concurrent implementation is not necessary.

The staff finds that the licensee has been given sufficient time to develop an A)pendix VII )rogram. Appendix VII has been generally adopted tirough-out tie industry. Although Appendix VIII will further improve flaw detection confidence, its implementation is not required in conjunction with Appendix VII. An Appendix VII program will increase quality and safety and is not considered impractical, l The staff concludes that sufficient technical justification has not been provided and, therefore, the licensee's request is denied.

Request for Relief 12R 11: Section XI, IWB-2420(b,c) and IWC-2420(b.c).

Successive Inspections. In the May 3, 1996, response to the NRC's RAI, the licensee withdrew the relief.

Request for Relief 12R 20: Pursuant to 10 CFR 50.55a(a)(3)(1), the licensee requested in lieu of the Code requirements to use Code Case N-509, Alternative Rules for the Selection and Examination of Class 1, 2. and 3 Integrally-Welded AttachmentsSection XI Division 1. The Code requires examination of integrally-welded attachments as specified for Examination C6tegories B H, B-K, C-C D-A, D-B and D-C. The Code stipulates volumetric or surface-examinations, as appropriate, and the e" tent of examinations, Non-mandatory Code cases may be used for ISI after general acceptance by the NRC staff and incorporation into Regulatory Guide 1.147. Pursuant to 10 CFR 50.55a, Code cases not incorporated into Regulatory Guide 1.147 may be used provided specific NRC authorization is obtained.

The licensee has proposed, as an alternative to the Code requirements, to apply the requirements of Code Case N-509 for the examination of integrally-welded attachments on Class 1, 2. and 3 piping and com>onents. Code Case N-509 provides alternative sampling requirements for tie examination of Class 1, 2, and 3 integral attachments. Review of this Code Case indicates that there is an ambiguity in the notes of the examination tables that would allow the selection of a 10% sample of the integrally-welded attachments imm the percentage of component su) ports selected for examination under the rums of the Code (specifically, Su)section IWF of the 1990 Addenda). -This coulo

potentially reduce the examination sample to an insignificant amount, or to no integral attachments at all. The staff finds that Code Case N 509 should be augmented to ensure that this does not occur. Considering that most of the Code examination requirements are based on sampling to ensure the detection of service-induced degradation extending the sampling philosophy to the integral attachment welds will provide an equivalent level of quality and safety for those welds. The licensee' has proposed in its alternative that 10% of the

-total population of non-exempt attachments will be examined when using Code Case N 509.

The staff concludes that licensee's proposal to use Code Case N-509 and to examine a minimum of 10% of the total number of all nonexempt Class 1, 2. and 3 piping, pump, and valve integral attachments provides an ecceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i). ~ Use of the Code Case as proposed is authorized for the second interval at Wolf Creek Generating Station, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

l Request for Relief 12R 15: Section XI. 1989 Edition. Subsection IWF-5000, t

references the first addenda to ASME/ ANSI OH-1987. Part 4 (0M Part 4) for visual examination and functional testing of snubbers. Pursuant to 10 CFR 50.55a(a)(3)(1). the licensee pro)osed to use the snubber inspection program that is currently in the Updated rinal Safety Analysis Report (USAR) as an alternative to the OM Part 4 program.

The first addenda to 0M Part 4 contains a visual examination schedule which was recommended for removal from Technical Specifications (TS) by Generic Letter (GL) 90-09 " Alternative Requirements for Snubber Visual Ins)ection Intervals and Corrective Actions." dated December 11, 1990. This G_ was issued to reduce the burden placed upon utilities by the then overly restrictive visual examination schedule. In May 1991. Amendment 44 was approved by the NRC. That amendment consisted of changes to the WCGS TS which now contain the alternate schedule recommended by the GL. In a letter dated November 17, 1995, the licensee clarified that the snubber inspection program was relocated from the TS to Chapter 16 of the USAR as a result of the TS relocation effort at Wolf Creek and was approved by the staff in Amendment No.

89 to WCGS Operating License No. NPF-42 dated October 2. 1995. Therefore, the licensee recuested that the snubber inspection program in Chapter 16 of the USAR be usec in lieu of the requirement of= Section XI of the ASME Code. 4 The staff has reviewed the licensee's submittal and finds that WCGS has comprehensive programs-for visual examination. repair and replacement and functional testing of all safety-related snubbers. The program scope encompasses all snubbers within Code Class 1, 2 and 3 boundaries and provides reasonable assurance of snubber operability. The WCGS snubber program previously approved by the staff is similar to the program required by 0M Part

4. In addition, the staff considered the licensee's program to be sufficient without duplication of the requirements in OM Part 4.

Based on the information provided, the staff finds that the licensee has presented adequate justification for its relief request from the requirements of ASME Code 1989 Edition.Section XI. Subsection IWF (which references OM Part 4), with regard to visual examination and functional testing of snubbers.

This is based on the fact that acceptable alternatives already exist in Chapter 16 of the WCGS USAR. The staff, therefore, concludes the )roposed alternative provides an acceptable level of quality and safety. Tierefore, the licensee's proposed alternative for visual examination, repair and re)lacement. and functional testing of snubbers is authorized pursuant to 10 CF1 50.55a(a)(3)(1).

3.0 CONCLUSION

The staff concludes that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code. In the cases of Requests for Relief 12R-03. 12R-04, 12R-05. I2R-06. 12R-07. 12R-08. 12R-14. and 12R-18, the licensee has demonstrated that specific Section XI requirements are impractical. Therefore, the staff concludes that, pursuant to 10 CFR 50.55a(g)(6)(1), relief is granted and the alternatives imposed 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 consideratie-to the burden upon the licensee that could result if the requirements were imposed on the facility.

The staff concludes that, for Requests for Relief I2R-01. 12R-02. 12R-10, 12R-12.12R-13,12R-15,12R-17. and 12R-20. the licensee's proposed alternatives provide an acceptable level of quality and safety in lieu of the Code required examinations and the proposed alternatives are authorized pursuant to 10 CFR 50.55a(a)(3)(1).

The statf concludes that, for Request for Relief 12R-19 the licensee has demonstrated that specific Section XI requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(11).

Request for Relief 12R-11 was withdrawn by the licensee, and deleted from the ISI Program Plan, by letter dated May 3,1996, in response to the NRC's request for additional information.

For Requests for Relief 12R-09 and 12R-16, it is concluded that the licensee has not provided sufficient justification to support the determination that in the case of the alternative contained in Request for Relief 12R-09.

compliance with the Code requirement would result in hardship or that, for Request for Relief 12R-16 the Code requirements are impractical. Therefore, in these cases, the alternative and relief is denied respectively.

Based on the review of the Wolf Creek Generating Station. Second 10-Year Interval Inservice Inspection Program Plan, 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, except those noted in the evaluation of Requests for Relief 12R-09 and 12R-16. Furthermore. this report does not include the review of how a licensee implenents augmented examination commitments: it merely notes if tht licensee has committed to the augmented examinations.

Attachment:

Technical Evaluation Report Principal Contributors: T. McLellan Y. C. Li Date: October 24, 1997 1

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