ML20106F797

Technical Evaluation Rept on Third 10-Yr Interval Inservice Insp Exam Plan:Northern State Power Co,Prairie Island Nuclear Generating Plant,Unit 2

ML20106F797
Person / Time
Site:	Prairie Island
Issue date:	12/31/1995
From:	Mary Anderson, Feige E, Hall K IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC (Affiliation Not Assigned)
Shared Package
ML20106F792	List: L-95-059, Technical Evaluation Rept on Third 10-Yr Interval Inservice Insp Exam Plan:Northern State Power Co,Prairie Island Nuclear Generating Plant,Unit 2
References
CON-FIN-L-2556 INEL-95-0592, INEL-95-592, NUDOCS 9602260311
Download: ML20106F797 (24)

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INEL-95/0592 December 1995 i

i i Idaho i National }

Eng/neering Laboratory Technical Evaluation Report on the Third 10-Year Interval inservice i Inspection Examination Plan:

Northern States Power Company, i i

Prairie Island Nuclear Generating Plant, i Unit 2, Docket Number 50-306

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M. T. Anderson l E. J. Feige K. W. Hall 4

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i i e Attachment to NRC SE Dated:

LOC /f/leed February 22, 1996 l Idaho Technologies Company

, 9602260311 960222 DR ADOCK0500g6

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INEL 95/0592 i ,

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i Technical Evaluation Report on the j

, Third 10-Year interval inservice inspection Examination Plan:

l Northern States Power Company,  !

Prairie Island Nuclear Generating Plant, Unit 2,  ;

Docket Number 50-306

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) M. T. Anderson E. J. Folge 4

K. W. Hall i

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, Published December 1995 t

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idaho National Engineering Laboratory Materials Physics Department j Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415 J.

Prepared for the

,. Division of Engineering I 4

Office of Nuclear Reactor Regulation  ;

U.s. Nuclear Regulatory Commission '

Washington. D.C. 20555 l

! Under DOE Idaho operations Office _

Contract DE-AC07-941D13223 FIN No. L2558 (Task Order 61)  !

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ABSTRACT This report presents the results of the evaluation of the Prairie Island Nuclear Generating Plant, Unit 2, Third 10-Year Interval Inservice Inspection Examination Plan, Revision 0, submitted November 15, 1994, 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 impractica1. The Prairie Island Nuclear Generating Plant, Unit 2, Third 10-Year Interval Inservice Inspection Examination Plan,

Revision 0 is evaluated in Section 2 of this report. The inservice inspection (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 i exc1usion criteria, and (d) compliance with ISI-related commitments identified during the Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.

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This work was funded under:

U.S. Nuclear Regulatory Commission FIN No. L2556, Task Order 61 Technical Assistance in Support of the NRC Inservice Inspection Program 4

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SUMMARY

The licensee, Northern States Power Company, has prepared the Prairie Island 1

Nuclear Generating Plant, Unit 2, Third 10-Year Interval Inservice Inspectfon

!. Examination Plan, Revision 0, to meet the requirements of the 1989 Edition of ,

I the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel i Code,Section XI. The third 10-year interval began December 21, 1994, and .

} ends December 20, 2004.

The informat'1on in the Prairie Island Nuclear Generating Plant, Unit 2, Third 10-Year Interval Inservice Inspection Exanination Plan, Revision 0, submitted l November 15, 1994, was reviewed. Included in the review were the requests for

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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 s additional information (RA!) was prepared describing the information and/or i f clarification required from the licensee to complete the review. The licensee j provided the requested information in the submittal dated July 13, 1995. As a j result of the review of the licensee's response, a conference call was held j August 21, 1995, to discuss the information required from the licensee to complete the review. This information was provided in a letter dated j October 5, 1995.

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Based on the review of the Prairie Island Nuclear Generating Plant, Unit 2, Thirti 10-Year Interval Inservice Inspection Examination Plan, Revision 0, 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 to the extent required by Section XI of the ASME Code, no deviations j from regulatory requirements or comitments were identified in the Prairie i

Island Nuclear Generating Plant, Unit 2, Third 10-Year Interval Inservice

-Inspection Examination Plan, Revision D.

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CONTENTS ABSTRACT

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.- 1. INTRODUCTION ............................. 1

2. ' EVALUATION OF INSERVICE INSPECTION PROGRAN PLAN . . . . . . . . . . . .4 2.1 Documents Evaluated ......................... 4 2.2 Compliance with Code Requirements ................

4 2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . .. 4 2.2.2 Acceptability of the Examination Sample ........... 5 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . .

5 j 2.2.4 Augmented Examination Commitments .............. 5 1

2.3 Conclusions ............................ 5 i

3. EVAttlATION OF RELIEF llEQUESTS . . . . . . . . . . . . . . . . . . . . 6 3.1 Class 1 Components . . . . . . . . . . . . . . . . . . . . . . . . 7

i. 3.1.1 Reactor Pressure Vessel (No requests for relief) j 3.1.2 Pressurizer (No requests for relief) 3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . 7 i 3.1.3.1 Request for Relief No. 5, Rev. 2, (Part 1),

Examination Category B-B, Item B3.160, Regenerative Heat Exchanger Circumferential Head Weld . . . . . . . . . 7 i

! 3.1.3.2 Request for Relief No. 5, Rev. 2, (Part 2),

! Examination Category B-D, Item B3.160, Regenerative

Heat Exchanger Nozzle Inner Radius Sections ....... 8 3.1.4 Piping Pressure Boundary (No requests for relief) 3.1.5 Pump Pressure Boundary . . . . . . . . . . . . . . . . . . . . 10 3.1.5.1 Request for Relief No. 3, Examination Categories B-L-1 and B-L-2, Items B12.10 and B12.20, Pump Casing Welds and Pump Casing ..................... 10 3.1.6 Valve Pressure Boundary (No requests for relief)  !

i 3.1.7 General (Norequestsforrelief) i iv l l

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. 3.2 Class 2 Components (No request for relief) 3.3 Class 3 Components (No requests for relief) 3.4 Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.4.1 Class 1 System Pressure Tests . . . . . . .......... 10 3.4.1.1 Request for Relief No. 02, Examination Category B-P, Item B15.50, Pressure Testing of the .oressurizer Auxiliary Spray Line . . . . . . . . . . . . . . . . . . . 10 3.4.2 Class 2 System Pressure Tests . . . . . . .......... 11 3.4.2.1 Request for Relief No.1 Table IWC-2500, Examination Category C-H, Class 2 Pressure Retaining Piping Non-Isolable from Class 1 Piping . . . . . . . . . . . . . . . 11 3.4.2.2 Request for Relief No. 4 Table IWC-2500, Examination Category C-H, Class 2 Steam Generator Hydrostatic Testing ......................... 11 ,

l 3.4.3 Class 3 System Pressure Tests . . . . . . .......... 11 1

3.4.3.1 Request for Relief No. 6, Examination Category D-C, Ites D3.10, Hydrostatic Testing of Class 3 Pressure Retaining Components in the Cooling Water Systen; . . . . . 11 3.4.4 General (No requests for relief) 3.5 General ............................. 13 3.5.1 Ultrasonic Examination Techniques . . . . .......... 13 3.5.1.1 Request for Relief No. 7, ASME Section XI, Appendix I, Paragraphs I-2100 and I-2200, Calibration Block Requirements for Piping Welds . . . . .......... 13

3.5.2 Exempted Components (No requests for relief) l 3.5.3 Other (No requests for relief)

4. CONCLUSION ............................. 16 1

5. REFERENCES ............................. 18 k

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. I TEcl#1ICAL EVALUATION REPORT ON THE  !

THIRD 10-YEAR INTERVAL INSERVICE INSPECTION EXANINATION PLAN: l NORTHERN STATES P0WER COMPANY,

' PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2, D0CKET NUMBER 50-306 i l -

l l- 1. INTRODUCTION  ;

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) Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including i supports) that are classified as American Society of Nechanical Engineers f (ASNE) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the l l requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code Section XI, " Rules for 1 l Inservice Inspection of Nuclear Power Plant Components," (Reference 2) to the l l extent practical within the limitations of design, geometry, and materials of

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i construction of the components. This section of the regulations also requires

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l that inservice examinations of components and system pressure tests conducted

! during successive 120-month inspection intervals comply with the requirements j in the latest edition and addenda of the Code incorporated by reference in l 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month I

inspection interval, subject to the limitations and modifications listed

! therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code 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) approval.

The licensee, Northern States Power Company, has prepared the Prairie Island Nuclear Generating Plant, Unit 2, ThIni 10-Year Interval Inservice Inspection Examination Plan, Revision 0, (Reference 3) to meet the requirements of the 1989 Edition, except that the extent of examination- for Class 1, Examination Category 8-J, has been determined by the requirements of the 1974 Edition through Summer 1975 Addenda (74S75) as permitted by 10 CFR 50.55a(b). The third 10-year interval began December 21, 1994, and ends December 20, 2004.

As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain Code examination requirements are impractical and requests relief from them, i 1

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4 j the licensee shall submit information and justifications to the NRC to support '

that determination.

i Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's determination that Code requirements are impractical to implement. The NRC j may grant relief and may impose alternative requirements that are determined  :

j to be authorized by law, will not endanger life, property, 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 imposed on the facility.

1 Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the.

! licensee's determination that either (1) the proposed alternatives provide an acceptable level of quality and safety, or (ii) Code compliance would result  !

in hardship or unusual difficulty without a compensating increase in safety.

Proposed alternatives may be used when authorized by the NRC. '

l The information 1n the Prairie Island Nuclear Generating Plant, Unit 2, Thirti 10-Year Interval inservice Inspection Examination Plan, Revision 0, submitted November 15, 1994, was reviewed, including the requests for relief from the

ASME Code Section XI requirements that the licensee has determined to be

! impractical. The review of the Inservice Inspection (ISI) Examination Plan was performed using the Standard Review P1'ans of NUREG-0800 (Reference 4),

i Section 5.2.4, " Reactor Coolant Boundary Inservice Inspections and Testing,"

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

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In a letter dated May 30,1995 (Reference 5), the NRC requested additional

! information that was required to complete the review of 'the ISI Examination i Plan. The requested information was provided by the licensee in the " Response l

! to Request for Additional Information on Unit 2, 3rd 10-year Interval  ;

j Inservice Inspection Examination and Associated Request for Relief (TAC No.

,- M90187)" dated July 13,1995(Reference 6). In this response, the licensee, I

Northern States Power Company, withdrew Request for Relief No. 3 and revised

[ No. 5. The licensee submitted Request for Relief No. 7 and withdrew Nos. 1, 2, and 4 by letter dated April 19, 1995 (Reference 7). As a result of this i

submittal, a conference call was held August 21, 1995, to discuss the I

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, information required from the licensee to complete the review. This information was provided in a letter dated October 5,1995 (Reference 8).

The Prairie Island Nuclear Generating Plant, Unit 2, Third 10-Year Interval

, Inservice Inspection Examination Plan, Revision 0, 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 ISI-related commitments identified during the NRC's previous reviews.

The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI, 1989 Edition. Specific inservice test programs for pumps and valves are being evaluated in other reports.

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2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN j , This evaluation consisted of a review of the applicable program documents to determine whether they are in compliance with the Code requirements and any j 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 following information from the licensee-  !

i i (a) Prairie Island Muclear Generating Plant, Unit 2, Third 10-Year \

l Interval Inservice Inspection Examination Plan, Revision 0, 1 J

(Reference 3);

, (b) Response to Request for Additional Information on the 3rd 10-year  ;

1 Interval Inservice Inspection Program and Associated Request for

! Reifef (TAC #o. #90187), dated July 13, 1995 (Reference 6);

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l (c) Request for Relief for the 3rd 10-year Interval Inservice Inspection

, Programs, dated April 19, 1995 (Reference 7); and l (d) Request for Relief for the 3rd 10-Year Interval Inservice Inspection Programs, dated October 5,1995 (Reference 8).

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l i 2.2 Como11ance with Code Reauirements i

2.2.1 Conoliance with Acolicable Code Editions

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l The Inservice Inspection Program Plan shall be based on the Code l

editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of December 21, 1994, the Code applicable to the third interval ISI program is the 1989 Edition. As stated in Section 1 of this report, the licensee has prepared the Prairie

! Island Nuclear Generating Plant, Unit 2, Thini 10-Year Interval Inservice Inspection Examinstion Plan, Revision 0, to meet the requirements of 1989 Edition of the Code, except that the extent of j examination for Class 1, Examination Category B-J has been determined

{ by the requirements of the 1974 Edition through Summer 1975 Addenda (74575) as permitted by 10 CFR 50.55a(b).

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, 2.2.2 Accentability of the Examination Samole  ;

Inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1, 2, and 3 components and their

, supports using sampling schedules described in Section XI of the ASME

, Code and 10 CFR 50.55a(b). Sample size and weld selection have been ,

implemented in accordance with the Code and 10 CFR 50.55a(b) and i i appear to be ccrrect.

2.2.3 Exemotion Criteria 3

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

j 2.2.4 Auamented Examination Commitments 4

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

(a) Reactor vessel examinations in accordance with the requirements i

of NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations,

Revision 1 (Reference 9);

i (b) Volumetric examination of the reactor coolant pump flyuheels in accordance with NRC Regulatory Guide 1.14, Reactor Coelant Pump Flywheel Integrity (Reference 10); and l

(c) Examination of the portions of high energy lines specified in '

Supplemental Reply to a Notice of Deviation, NRC inspection

• Report Mos. 282/92008 and 306/92008 Final Safety Analysis Report Comattaent for Inservice Examination of High Energy line Piping, ,

(Reference 11).

2.3 Conclusions i

Based on the review of the documents listed above, no deviations from '

regulatory requirements or commitments were identified in the yrafrie 5

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, Island Nuclear Generating Plant, Unit 2, Third 10-Year Interval Inservice Inspection Examination Plan, Revision D.

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- 3. EVALUATION 0F RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the third 10-year inspection interval are evaluated in the following sections.

j 3.1 Class 1 Components 1

3.1.1 Reactor Pressure Vessel (No requests for relief) 3.1.2 Pressurizer (No requests for relief) 4

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1 3.1.3 Heat Exchanaers and Steam Generators i

3.1.3.1 Reauest for Relief No. 5. Rev. 2 (Part 1). Examination Cateaory B-B. Item B2.51. Reaenerative Heat Exchanaer

$ Circumferential Head Weld 4

i Code Reauirement: Examination Category B-B, Item 82.51 requires i a volumetric examination of heat exchanger circumferential head welds, as defined in Figure IWB-2500-20(f), using acceptance standard IWB-3510.

Licensee's Code Relief Reauest: The licensee requested relief from using acceptance standard IWB-3510 for the Regenerative Heat Exchanger Circumferential Head Weld.

Licensee's Basis for Reauestino Relief (as stated):

"The acceptance standard is for ferritic vessels; the

, regenerative heat exchanger is austenitic.

4 "The regenerative heat exchanger is more pipe like than vessel like. The head is a 6 inch pipe cap; made of cast stainless; 0.375 inches thick. The integral tubesheet. is a 6 inch forging; 4

stainless; 0.500 inches thick at the weld."

i Licensee's Proposed Alternative Examination (as stated):

"The acceptance standard used for limited volumetric examination of the regenerative heat exchanger circumferential head weld will 7

i be IWB-3514.1, IWB-3514.3 and Table-3514-2 for austenitic

,' piping."

l Evaluation: The Code requires volumetric examination of heat i exchanger circumferential ' head welds, applying the acceptance i criteria of IWB-3510. However, the licensee proposed to use <

l acceptance standards IWB-3514.1, IWB-3514.3, and Table-3514-2,

which are for austenitic piping welds.

. The required acceptance standard uses several tables which are

, explicitly for ferritic steels. However, the regenerative heat

. exchanger is not constructed of ferritic steel-it consists of a i

cast stainless pipe cap welded to an integral tubesheet stainless forging. Therefore, the acceptance standards used for this component should be ones designated for stainless steels.

j Furthermore, the design of the heat exchanger, being fabricated of piping components, suggests the use of the acceptance i standards for piping. IWB-3514.1, IWB-3514.3, and Table-3514-2

are required to be used for Examination Category B-J piping welds i and are identified for use with austenitic steel components.

Based on this review, it has been determined that the use of these acceptance standards, as requested by the licensee, will provide an acceptable level of quality and safety.

Conclusion:

For welds on the regenerative heat exchanger, the proposed use of IWB-3514.1, IWB-3514.3,. and Table-3514-2 in lieu l

of acceptance standard IWB-3510 will provide an acceptable level

of quality and safety. Therefore, it is recommended that the

proposed alternative be authorized pursuant to l 10CFR50.55a(a)(3)(1).

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3.1.3.2 Reauest for Relief No. 5. Rev. 2 (Part 2). Examination Cateaory B-D. Item B3.160. Recenerative Heat Exchanaer Nozzle Inner Radius Sections Code Reauirement: Examination Category B-D, Item B3.160 requires i volumetric examination of heat exchanger nozzle inside radius i

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- sections as dafined in Figure IWB-2500-7(a) through (d), as l applicable.

l Licensee's Code Relief Reauest: The licensee requested relief l from performing the Code-required volumetric examinations on the j nozzle inner radius sections.

Licensee's Basis for Reauestina Relief (as stated):

"The integral tubesheet component includes the 2 inch nozzles.

The forged nozzle ends are socket welded to the inlet and outlet pipe. This arrangement results in complex geometry and limited 4

accessible scan area. Therefore, a volumetric examination would provide no meaningful information. Since the heat exchangers are in a locked radiation area, ALARA considerations would recommend not performing an examination which provides no significant benefit."

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Licensee's Proposed Alternative Examination (as stated):

l "When insulation is removed for the regenerative heat exchanger circumferential head weld examination, the nozzle area will be j visually inspected."

Evaluation: The Code requires 100% volumetric examination of j

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nozzle inside radius sections of Class I heat exchangers.

However, as stated by the licensee, the size and geometry of the regenerative heat exchanger nozzles preclude volumetric examination of the inside radius sections. The licensee's drawing' confirms that, due to the design of the subject nozzles, ultrasonic examination of the inside radius sections is impractical. To meet the Code requirement, the regenerative heat exchanger nozzles would have to be redesigned and replaced. This would represent a considerable burden for the licensee.

The visual examination of the subject area proposed by the licensee and the Code-required volumetric examination of the associated vessel-to-head weld should detect any significant inservice degradation that say. occur.

• Ilot included in this report.-

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Conclusion:

Volumetric examination is impractical for the

regenerative heat exchanger nozzle inside radius sections.

, However, the examinations being performed will provide reasonable ,

assurance of operational readiness. Therefore, it is recommended i

that relief he granted pursuant to 10 CFR 50.55a(g)(6)(1).

3.1.4 Pinino Pressure Boundary (No requests for relief) 3.1.5 Pumn Pressure Boundarv 3.1.5.1 Reouest for Relief No. 3. Examination Cateaorfes B-L-1 and B-L-2.

Items B12.10 and B12.20. Pumo Casina Welds and Pumo Casina HQIE: The licensee withdrew Request for Relief No. 3, by letter dated July 13, 1995, in response to the NRC's request for s

additional information.

J 3.1.6 Valve Pressure Boundary (No requests for relief) 3.1.7 General (No requests for relief) 1 3.2 Class 2 Components (No request for relief)

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3.3 Class 3 Components (No requests for relief) 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests

,3.4.1.1 Reauest for Relief No. 2. Examination Cateaory B-P. Item B15.50.

d Pressure Testino of the Pressurizer Auxiliary Sorav Line HQIE: In a letter dated April 19, 1995, Request for Relief No. 2 was withdrawn by the licensee based on the decision to use Code Case N-498, which was approved for general use in Revision 9 to NRC Regulatory. Guide 1.147. I

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3.4.2 Class 2 System Pressure Tests

[ 3.4.2.1 Reaue'st for Relief No. 1 Table IWC-2500. Examination

, Cateaory C-H. Class 2 Pressure Retainina Picina Non-Isolable from i, Class 1 Pinina l M: In a letter dated April 19, 1995, Request for Relief No. 1 l was withdrawn by the licensee based on the decision to use Code

Case N-498, which was approved for general use in Revision 9 to

NRC Regulatory Guide 1.147.

3.4.2.2 Reauest for Relief No. 4. Table IWC-2500. Examination

! Cateaory C-H. Class 2 Steam Generator Hydrostatic Testina j

4 E: In a letter dated April 19, 1995, Request for Relief No. 4 l was withdrawn by the licensee based on the decision to use Code

Case N-498, which was approved for general use in Revision 9 to NRC Regulatory Guide 1.147.

3.4.3 Class 3 System Pressure Tests

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3.4.3.1 Reauest for Relief No. 6. Examination Cateaory D-C. Item D3.10.

Hydrostatic Testina of Class 3 Pressure-Retainino Components in j the Coolina Water System Code Reautrement: Table IWD-2500-1, Examination Category D-C, j Item D3.10, requires a system hydrostatic test as specified by

IWD-5223. IWD-5223-states that the system hydrostatic test t

pressure shall be at least 1.10 times the system pressure for systems with design temperatures of 200*F or less, and at least 1.25 times the system pressure for systems with design

.. temperatures greater than 200*F.

l Licensee's Code Relief Reauest: The licensee requested relief from performing the Code-required system hydrostatic tests of the Class 3 Cooling Water System.

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i Licensee's Basis for Reauestino Relief (asstated):

"The cooling water system is designed such that Unit I and Unit 2 i

safeguards equ!pment is supported from both sides of the cooling water system header. Consequently, the entire supply and return header must be in operation at all times to meet operating license requirements."

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Licensee's Pronosed Alternative Examination (as stated):

i "The cooling water system will be visually examined every one-third interval for conditions adverse to system operation.

Additionally, the system is in constant operation and any leaks would be-immediately known. Portions that are isolatable from j

the main headers will be pressure tested in accordance with the applicable requirements."

Evaluation: The Code requires a system hydrostatic pressure test for Class 3 pressure-retaining components. The licensee stated that the subject lines are the only source of cooling water for the Cooling Water Supply and Return headers and would have to be h taken out of service to conduct the Code-required hydrostatic

) test. Since the plant operating license requires cooling water 3

flow during both operation and refueling modes, taking these lines out of service would be a violation of the operating j license. Therefore, the Code requirements are impractical for

, these lines. In lieu of this requirement, the licensee proposed to perform a system inservice test with an associated VT-2 visual examination, providing reasonable assurance of the system's operational readiness.

Conclusion:

The hydrostatic tes', of the subject portions of the

cooling water system is impractical at Prairie Island, Unit 2.

The licensee's proposed alternative will provide reasonable assurance of the system's operational readiness. Therefore, it i

is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).

' 3.4.4 General (No requests for relief) 1 4

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3.5 General j

, 3.5.1 Ultrasonic Examination Techniaues 3.5.1.1 Reauest for Relief No. 7. ASME Section XI. Anoendix I.

l Paraaraohs 1-2100 and I-2200. Calibration Block Reauirements for Pinina Welds i

61 Code Reauirement

Paragraph I-2100 requires ultrasonic examination of vessel welds greater than 2-in. thick in accordance with Article 4 of Section V.Section V, Article 4, Paragraph T-441.1.2.1 requires the calibration blocks to be fabricated from one of the following:

"(a) Nozzle drop out from the component; (b) A component prolongation; (c) Material of the same material specification, product 2

form, and heat treatment condition as of the materials being joined."

Paragraph I-2200 requires ultrasonic examination of vessel welds less than or equal to 2-in. thick in accordance with j Appendix III. Appendix III, Paragraph III-3411 requires:

"(a) The calibration block for similar metal welds shall be l fabricated from one of the materials specified for the piping being joined by the weld.

J (c) Where examination is to be performed from only one  ;

side of the joint, the calibration block material  !

shall be of the same specification as the material on  !

that side of the joint.

(d) If material of the same specification is not )

I j available, material of similar chemical analysis, tensile properties, and metallurgical structure may be used."

Licensee's Code Relief Reauest: Relief is requested from the

material requirements of Section V, Article 4, ParagraphT-441.1.2.I(a,c,d),andSectionXI,AppendixIII, l i

Paragraph 111-3411 for existing calibration blocks. 4 1

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i Licensee's Basis for Reauestina Relief (as stated):

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" Documentation requirements existing at the time of fabrication did not require traceability to the material's chemical or l- physical certifications. Existing calibration blocks certification is verified through appropriate P-number grouping.

The P-number grouping provides adequate assurance that' the blocks i

will establish the proper ultrasonic calibration and sensitivity.

' Using P-number grouping to choose calibration blocks was allowed by the 1971 ASME B&PV Code Section III, Paragraph IX-3431.

"It would be impractical to fabricate a new set of calibration I

j blocks in order to satisfy the documentation requirements of the l current Code. Existing records, indicate the appropriate.

P-number grouping, thereby providing adequate assurance that the

). blocks will establish the proper ultrasonic calibration and sensitivity."

Licensee's Proposed Alternative Examination (as stated):

l

" Existing calibration blocks will be used as is.

"Any calibration blocks obtained in the future will be obtained with documentation to demonstrate compliance with the material

specification requirements of ASME Code Section V Article 4 or i Section XI, Appendix III, as applicable."

In a letter dated October 5, 1995, the following statements were added.  !

l l

! " Existing calibration blocks greater than 1" thick have been i

verified to require no correction for attenuation differences.

" Additionally, when using existing calibration blocks less than 1" thick that lack the appropriate documentation and when an i indication is detected, a comparison will be made between the i attenuation of the calibration block and the material being ,

examined."  !

Evaluation: Section XI, Appendix I, Paragraphs I-2100 and I-2200  :

require that calibration blocks be of the same material or a j material of similar chemical, tensile, and metallurgical i properties as the weld being examined. However, the calibration blocks at Prairie Island were constructed to the 1971 Edition of l

, Section III, which only required that they be of the same l P-number grouping. I 1

14  !

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s .

4 The use of existing calibration blocks, fabrf ated as required by the original construction Code, will provide an acceptable method l of establishing the proper ultrasonic calibration and sensitivity, provided the acoustical properties are similar to those of the examination volume. Furthermore, requiring the licensee to replace all calibration blocks would impose a considerable burden.

The existing blocks have been proven satisfactory for performing calibrations. Therefore, any increase in plant safety that might occur with new blocks would not compensate for the burden placed on the licensee to fabricate new calibration blocks to satisfy the current Code requirements.

However, when using the existing calibration blocks without the j appropriate documentation, a comparison should be made between the acoustical properties (i.e., velocity and attenuation) of the

{

calibration block and the material being examined. This comparison should be done once, prior to the use of the calibration block, to ensure that the sensitivity is sufficient to find existing flaws in corresponding examination volumes.

Conclusion:

Requiring the licensee to fabricate new calibration l blocks would result in a burden without a compensating increase f in the level of quality and safety. The existing calibration i

blocks will provide an acceptable examination sensitivity i provided the acoustical properties are verified to be similar to those of the area being examined. Therefore, it is recommended ,

that the proposed alternr.tive be authorized pursuant to l 10 CFR 50.55a(a)(3)(ii), with the above condition.

3.5.2 Exemoted Comoonents (No requests for relief)

3.5.3 Other (No requests for relief) 15

- _ _ - . _ __ _ ~ _- -- _.

i

4. CWCLUSION Pursuant to 10 CFR 50.55a(g)(6)(1), it has been determined that certain inservice examinations cannot be performed to the extent required by

, Section XI of the ASME Code. In the case of Requests for Relief Nos. 5,

Rev. 2 (Part 2) and 6, 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 I

interest, giving due consideration to the burden upon the licensee that could 1

result if the requirements were imposed on the facility.

i

) Pursuant to 10 CFR 50.55a(a)(3)(1), it is concluded that for Request for Relief No. 5, Rev. 2 (Part 1), the licensee's proposed alternative will provide an acceptable level of quality and safety in lieu of the Code-required acceptance standard. In this case, it is recommended that the proposed alternative be authorized.

Pursuant to 10 CFR 50.55a(a)(3)(ii), it is concluded that for Request for i i Relief No. 7 the licensee has demonstrated that specific Section XI requirements would result in hardship er unusual difficulty without a l compensating increase in the level of quality and safety. In this case, it is i

recommended that the proposed alternative be authorized only if the licensee satisfies the conditions stated in the above request for relief evaluation.

Requests for Relief Nos.1, 2, 3, and 4 were withdrawn by the licensee, and deleted from the ISI Examination Plan, by letters dated April 19, 1995, and July 6,1995, in response to the NRC's request for additional information.

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 existing Prairie Island Nuclear Generating Plant, Unit 2. Compliance with all of the Section XI examination requirements I would necessitate redesign of a significant number of plant systems, procurement of replacement components, installation of the new components, and performance of baseline examinations for these components. Even after the redesign efforts, complete compliance with the Section XI examination 16

k requirements probably could not be achieved. Therefore, it is concluded that the public interest is not served by imposing certain provisions of Section XI ,

of the ASME Code that have been determined to be impractical.

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

Based on the review of the Prairie Island Nuclear Generating Plant, Third

10-Year Interval Inservice Inspection Ex' amination Plan, Revision 0, the I licensee's response to the NRC's request for additional information, and the recommendatior;s 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.

1 Y

i i

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f j

i

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. . . i l

l

, 5. REFERENCES l 1. Code of Federal Regulations, Title 10, Part 50.

.~ 2.

American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section KI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 1:

1989 Edition j) 1974 Edition Summer 1975 Addenda i 3. Prairie Island Nuclear Generating Plant, Third 10-Year Interval Inservice i

)

Inspection Examination Plan, Revision 0, dated November 15, 1994.

4. NUREG-0800, Standani Review Plan for the Review of Safety Analysis i

Reports for Nuclear Power Plants, Section 5.2.4, " Reactor Coolant

, Boundary Inservice Inspection and Testing," and Section 6.6, " Inservice j Inspection of Class 2 and 3 Components," July 1981.

5. Letter, dated May 30, 1995, Charles R. Thomas (NRC) to Roger 0. Anderson 1

(Northern States Power), containing request for additional information on j the third 10-year interval ISI program plan.

6. Letter, dated July 13,1995, Roger 0. Anderson (Northern States Power) to Document Control Desk (NRC), containing the response to NRC request for additional information.

7. Request for Relief for the 3ni 10-year Interval Inservice Inspection Programs, dated April 19,1995 (Reference 7);

. 8. Request for Relief for the 3rd 10-Year Interval Inservice Inspection l Programs, dated October 5,1995.

i l 9. NRC Regu1atory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservfce Examinstfans, Revis1on 1, February 1983.

10. NRC Reguintory Guide 1.14, Reactor Coolant Pump Flywheel Integrity, j Revision 1, August 1975.

^

11. Supplemental Reply to a Notice of Deviation NRC Inspection Report i

i Nos. 282/92008 and 306/92008 Final Safety Analysis Report Comaitment for Inservice Examination of High Energy line Piping, dated

{ February 10, 1993.

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Prairie Island Nuclear Generating Plant, Unit 2 Dec~ ember 1 199$..a Docket Number 50-306

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10. SUPPLEMENTARY NOTE 5

11. A33TR ACT (200 =wum ar emJ This report presents the results of the evaluation of the Prairie Island Nuclear Generating Plant, Unit 2, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted November 15, 1994 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 Prairie Island Huclear Generating Plant, Unit 2, Third 10-Year Interval Inservice Inspection Program Plan, is evaluated in Section 2 of this report. The Inservice Inspection (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 ISI-related comitments identified during previous Nuclear Regulatory Comission reviews. The requests for relief are evaluated in Section 3 of this report.

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