ML20100H605

Technical Evaluation Rept on Third 10-Year Interval Inservice Insp Program Plan:Northern States Power Co, Prairie Island Nuclear Generating Plant,Unit 1

ML20100H605
Person / Time
Site:	Prairie Island
Issue date:	11/30/1995
From:	Mary Anderson, Hall K, Porter A IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC (Affiliation Not Assigned)
Shared Package
ML20100H547	List: L-95-056, Technical Evaluation Rept on Third 10-Year Interval Inservice Insp Program Plan:Northern States Power Co, Prairie Island Nuclear Generating Plant,Unit 1
References
CON-FIN-L-2556 INEL-95-0564, INEL-95-564, NUDOCS 9602270375
Download: ML20100H605 (25)

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- INEL-95/0564

, November 1995

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I idaho National j Engineering Laboratory Technical Evaluation Report on the Third 10-year Intervalkservice -

! Inspection Program Plan:

I Northern States Power Company,

! Prairie Island Nuclear Generating Plant, i Unit 1, l Docket Number 50-282

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! M. T. Anderson K. W. Hall A. M. Porter l

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Attachment to NRC SE dated:

YlOChheed

/daho Technologies Compery February 22, 1996 9602270375 960222

, PDR ADOCK 05000282

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4-INEL-95/0564 . ,

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l' Technical Evaluation Report on the . i Third 10-year interval inservice inspection Program Plan:

Northem States Power Company,

Prairie Island Nuclear Generating Plant, Unit 1, j Docket Number 50-282 e

! l 4-4 M. T. Anderson

K. W. Hall i A. M. Porter V

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, Published November 1995 i

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Idaho National Engineering Laboratory l Materials Physics

' Lockheed Idaho Technologies Company idaho Falls, Idaho 83415 i-l l Prepared for the l..

Division of Engineering Office of Nuclear Reactor Regulation

- U.s. Nuclear Regulatory Commission Washington, D.C. 20555

- Under doe Idaho operations Office Contract DE AC07 944D13223 FIN No. t.2556 (Task order 57) l l _ _ _ _ _ - _ _ _ _ _ _ _ _

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. ABSTRACT This report presents the results of the evaluation of the Prairie Island Nuclear Generating Plant, U' n f t 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision I, including the requests for relief from the American

' Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI requirements that the licensee has determined to be impractical.

The Prairie Island nuclear Generating Plant, Unit 1, 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 4

examination sample, (c) correctness of the application of system or component examination excluston criteria, and (d) compliance with ISI-related connitments identified during previous Nuclear Regulatory Commission (NRC) 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 1 FIN No. L2556, Task Order 57 Technical Assistance in Support of the NRC Inservice Inspection Program i

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SUMMARY

1 j- The licensee, Northern States Power Company, has prepared the Prairie Island j Muclear Generating Plant, Unit 1, Third 10-Year Interval Inservice Inspection l

. Program Plan, Revision 0, to meet the requirements of the 1989 Edition of the l

American Society of Mechanical Engineers Boiler and Pressure Vessel Code, ,

Section XI. The third 10-year interval began December 17, 1993 and ends December 16, 2003.

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l i The information in the Prairie Island Nuclear Generating Flant, Unit 1, Third 1 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted j August 5 1994, was reviewed. Included in the review were the requests for l 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 l

[ additional information (RAI) was prepared describing the information and/or l l clarification required from the licensee in order to complete the review. The i licensee provided the requested information in a submittal dated '

i March 28, 1995. As a result of a telephone conversation with the licensee on i Nay 30, 1995, the licensee submitted Prairie Island Nuclear Generating Plant,

Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision 1, I dated July 6, 1995. As a result of the review of Revision 1, a conference

call was held August 21, 1995, to discuss the information required from the

! licensee in order to complete the review. This information was provided in a letter dated October 5,1995.

Based on the review of the Prairie Island Nuclear Generating Plant, Unit 1, l Third 10-Year Interval Inservice Inspection Program Plan, Revision 1, 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 from regulatory requirements or commitments were identified in the Prairie

Island Nuclear Generating Plant, Unit 1, Thirti 10-Year Interval Inservice i I Inspection Progran Plan, Revision 1.

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CONTENTS ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-i_

SUMNARY . . . ... . . . . . . . . .. . . . . . . . . . . . . . . . . . . . iii.

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

2. ' EVALUATION OF. INSERVICE INSPECTION PROGRAN PLAN . . . . . . . . . . . 4 i 2.1 Documents Evaluated . . . . . . . . . . . . . . . . . . . . . . . . 4 i

2.2 Compliance with Code Requirements ...............'.

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2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . 4 2.2.2 Acceptability of the Examination Sample ........... 5 9 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . 5

2.2.4 Augmented Examination Commitments .............. 5 2.3 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 lr 3. EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . 7 ,

n 3.1 Class 1 Components . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.1 Reactor Pressure Vessel (No requests for relief) 3.1.2 Pressurizer (No requests for relief) '

3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . 7 3.1.3.1 Request for Relief No. 05, Rev. 2, (Part 1),

' Examination Category B-B, Item B3.160, Regenerative i

Heat Exchanger Circumferential Head Weld . . . . . . . . . 7  :

3.1.3.1 Request for Relief No. 05, Rev. 2, (Part 2),

i Examination Category B-D, Iten B3.160, Regenerative

Heat Exchanger Nozzle Inner Radius Sections ....... 9 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 Pum i and Pump Casing . . . . . . . . .,. . p Casing Welds

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3.1.6 Valve Pressure Boundary (No requests for relief) '

3.1.7 General (No requests for relief) i

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3.2 Class 2 Components (No request for relief) 3.3 f. lass 3 Components (No requests for relief)  :

3.4 Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.4.1 Class 1 System Pressure Tests (No requests for relief) 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 >

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

. 3.4.3.1 Request for Relief No. 2 Examination Category D-C, Item D3.10, Hydrostatic Testing of Class 3 Pressure Retaining Components in the Cooling Water System . . . . . 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. 6, ASME Section XI, Appendix I, l Paragraphs I-2100 and -2200, Calibration Block Requirements for Piping Welds . . . . . . . . . . . . . . 13 1

3. 5. 2 Exempted Components (No requests for relief) 3.5.3 Other (No requests for relief)

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

5. REFERENCES ............................. 18 l

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TECWICAL EVALUATION REPORT ON THE 4

THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAN PLAN:

NORTHERN STATES POWER COMPANY,

, PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 D0CKET NUNBER 50-282 I~

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1. INTRODUCTION 1

Throughout the service life of a water-cooled nuclear power facility, .

i 10-CFR 50.55a(g)(4)'(Reference ~1) requires that components (including

, supports) that are classified as An.erican Society of Mechanical Engineers j (ASNE) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASNE Code Section XI, Rules for l Inservice Inspection of Nuclear Power Plant Components (Reference 2), to the j' extent practical within the limitations of design, geometry, and materials of

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

. that inservice examinations of components and system pressure tests conducted

during successive 120-month inspection intervals shall comply with the l

requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the l 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set

! forth in subsequent editions and addenda of this Code that are incorporated by l reference in 10 CFR 50.55a(b) subject to the limitations and modifications ,

l listed therein, and subject to Nuclear Regulatory Commission (NRC) approval.

i The licensee, Northern States Power Company, has prepared the Prairie Island l Nuclear Generating Plant, Unit 1, Third 10-Year Interval Inservice Inspection l Program Plan, Revision 0 (Reference 3), to meet the requirements of the 1989 Edition, except that the extent of examination for Class 1, Examination Category B-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 e

third 10-year interval began December 17, 1993 and ends December 16, 2003.

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

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the licensee shall submit information and justification 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 '

f may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common i 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.

l Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the licensee's determination that either (i) 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.

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

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. The information in the Prairie Island Nuclear Generating Plant, Unit 1, Thini 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted i

August 5, 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) Program Plan was l performed using the Standard Review Plans of NUREG-0800 (Reference 4),

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

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

3 In a letter dated January 20, 1995 (Reference 5), the NRC requested additional l f information that was required in order to complete the review of the ISI Program Plan. The requested information was provided by the licensee in the

" Response to Request for Additional Information on the 3rd 10-year Interval Inservice Inspection Program and Associated Request for Relief (TAC No.

. M90186)" dated March 28, 1995 (Reference 6). In this response, the licensee, 4

Northern States Power Company, withdrew 2 requests for relief, revised 2 requests for relief, and committed to revise the plan to include small bore High Pressure Injection piping. As a result of a telephone converstt on with i

the licensee on May 30, 1995, Request for Relief No. 3 was withdrawn, No. 5 was revised, No. 6 was submitted, and small bore High Pressure Injection 2

, piping was included in the re-submittal by the licensee of the Prairie Island Nuclear Generating Pinnt, Unit 1, Third 10-Year Interval Inservice Inspection

, Program Plan, Revision 1, on July 6,1995, (Reference 7). A conference call was held August 21, 1995 to request clarification on information in Request for Relief No. 6. This clarification was provided by the licensee in a letter dated October 5, 1995 (Reference 8).

The Prairie Island Nuclear Generating Plant, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision 1, is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with 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 (IST) programs for pumps and valves are being evaluated in other reports.

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.. 2. EVALUATION 0F INSERVICE INSPECTION PROGRAN PLAN .

2 This evaluation consists of a review of the applicable program documents to

) determine whether they are in compliance with the Code requirements and any

previous liccasa conditions pertinent to ISI activities. .This section describes the submittals reviewed and the results of the review.

1 j 2.1 Documents Evaluated -

1 b l l Review has been completed on the following information from the licensee:

i (a) Prairie Island Nuclear Generating Plant, Unit 1, Third 10-Year

i. Interval Inservice Inspection Examination Plan, Revision 0 l, (Reference 3);

l (b) Response to Request for Additional Information on the 3rd 10-year i Interval Inservice Inspection Program and Associated Request for l

Relief (TAC No. M90186) dated March 28, 1995 (Reference 6):

(c) Prairie Island Nuclear Generating Plant, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revis1on 1 (Reference 7):

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i (d) Request for Reifef for the 3rd 10-Year Interval Inservice Inspection Programs dated October 5, 1995 (Reference 8). j

2.2 Como11ance with Code Reauirements

) 2.2.1 Como11ance with Aeolicable Code Editions The Inservice Inspection Program Plan shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of December 15, 1993, 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 1, Thini 10-Year Inservice Inspection Program Plan, i Revision 1, to meet the requirements of 1989 Edition, except that the j extent of examination for Class 1, examination category B-J has been determined by the requirements of the 1974 Edition through Summer 1975 l

} 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 i 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 appear to be correct. -

2.2.3 Exemotion Criteria '

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The criteria used to exempt components from examination shall be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, IWD-1220, and 10 CFR 50.55a(b). The exemption criteria have been applied by the )

licensee in accordance with the Code as discussed in the ISI Program l Plan, and appear to be correct. I I

, 1 2.2.4 Auamented Examination Commitments

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

i (a) Reactor vessel examinations in accordance with the requirements of NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, Revision 1, (Reference 9);

(b) Volumetric examination of the reactor coolant pump flywheels satisfying NRC Regulatory Guide 1.14, Reactor Coolant Pump flywheel Integrity, (Reference 10); and (c) Examination of the portions of high energy lines specified in Supplemental Reply to a Notice of Deviation, NRC inspection Report Nos. 282/92008 and 306/92008 Final Safety Analysis Report Commitment for Inservice Examination of High Energy Line Piping, (Reference ll).

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2.3 Conclusion l Based on the review of the documents listed above, no deviations from

! regulatory requirements or commitments were identified in the Prairie i

Island Nuclear Generating P1artt, Third 10-Year Interval Inservice

Inspection Progran Plan, Revision 1. '

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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.

3.1 Class 1 Comoonents 3.1.1 Reactor Pressure Vessel (No requests for relief) -

3.1.2 Pressurizer (No requests for relief) 3.1.3 Heat Exchanaers and Steam Generators 3.1.3.1 Reauest for Relief No. 05. Rev. 2. (Part 1). Examination Cateaory B-B. Item B2.51. Reaenerative Heat Exchanaer Circumferential Head Weld

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l Code Reauirement: Examination Category B-B, Item B2.51 requires a volumetric examination of heat exchanger circumferential head I welds as defined in Figure IWB-2500-20(f), using acceptance l standard IWB-3510.

l 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.

"The regenerative heat exchanger is more pipe like than vessel

' like. The head is a 6 inch pipe cap; made of cast stainless; O.375 inches thick. The integral tubesheet is a 6 inch forging; l stainless; 0.500 inches thick at the weld."

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Licensee's Proposed Alternative Examination (as stated):

"The acceptance standard used for limited volumetric examination ,

of the regenerative heat exchanger circumferential head weld will

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

! piping."

l Evaluation: The Code requires volumetric examination of heat f exchanger circumferential head welds using IWB-3510 for acceptance criteria. However, the licensee proposed to use the  :

{ acceptance standards IWB-3514.1, IWB-3514.3 and Table-3514-2 for austenitic piping welds.

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. The required acceptance standard uses several tables which are i

explicitly for ferritic steels. However, the regenerative heat

, exchanger is not constructed of ferritic steel. The regenerative heat exchanger head consists of a cast stainless pipe cap and an

) integral tubesheet stainless forging. Therefore, the acceptance

standards used for this component should be designated for j stainless steels. Furthermore, the design of the heat exchanger, l being fabricated of piping components, suggests the use of the

, acceptance standards for piping. IWB-3514.1, IWB-3514.3 and

Table-3514-2 are required to be used by Examination Category B-J l piping welds and are identified for use with austenitic steel components. Use of these acceptance standards, as requested by j the licensee, will provide an acceptable level of quality and safety.

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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 j 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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. 3.1.3.1 Reauest for Relief No. 05. Rev. 2. (Part 2). Examination i Cateaory B-D.' Item B3.160. Recenerative Heat Exchanaer Nozzle Inner Radius Sections l Code Reauirement: Examination Category B-D, Item B3.160 requires a volumetric examination of heat exchanger nozzle inside radius

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

Licensee's Code Relief Reauest: The licensee requested relief l from performing the Code-required volumetric examinations on the l

nozzle inner radius sections.

t Licensee's Basis for Reauestina Relief (as stated): >

I "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 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."

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

"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

nozzle inside radius sections of Class I heat exchangers.

j However, as stated by the licensee, the size and geometry of the i 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 perform. To meet the Code requirement, the

• Drawing was included in the licensee's response to the NRC's RAI, but, is not included in this report.

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and replaced. This would represent a considerable burden for the

, licensee.

The volumetric examination is considered impractical to perform en the regenerative heat exchanger nozzle inside radius sections.

The visual examination of the subject area offered by the licensee in conjunction with the Code required volumetric examination of the associated vessel-to-head weld should detect I any significant inservice degradation that may occur.

Conclusion:

Based on the above evaluation, the volumetric examination is considered impractical to perform on the regenerative heat exchanger nozzle inside radius sections.

s Newever, the examinations being performed will provide reasonable assurance of operational readiness. Therefore, it is recommended j

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

t 3.1.4 Pinism Pressure Boundary (No requests for relief) 3.1.5 Pumn Pressure Boundary 3.1.5.1 Reauest for Relief No. 3. Examination Cateaories B-L-1 and B-L-2.

Items B12.10 and B12.20. Pumo Casina Welds and Pumo Casina i

jg1E: In the July 6,1995, response to the NRC's conference call, the licensee withdrew Request for Relief No. 3.

j 3.1.6 Valve Pressure Boundarv (No requests for relief) '

3.1.7 General (No requests for relief) 3.2 Class 2 r===onents (No request for relief) 3.3 Class 3 Cannonents (No requests for relief) 3.4 Pressure Tests 10

3.4.1 Class 1 System Pressure Tests (No requests for relief)

, 3.4.2 Class 2 System Pressure Tests 3.4.2.1 Reauest for Relief No.1 Table IWC-2500. Examination Cateaory C-H. Class 2 Pressure Retainina Pipina Non-Isolable from Class 1 Pinina HQII: -In the March 28, 1995, response to the NRC's request for additional information, the licensee withdrew Request for i

Relief No.1, based on their decision to use Code Case N-498.

3.4.2.2 Reauest for Relief No. 4 Table IWC-2500. Examination Cateaory C-H. Class 2 Steam Generator Hydrostatic Testina 4

HQIf: In the March 28, 1995, response to the NRC's request for additional information, the licensee withdrew Request for Relief No. 4, based on their decision to use Code Case N-498.

3.4.3 Class 3 System Pressure Tests 3.4.3.1 Reauest for Relief No. 2. Examination Catecory D-C. Item D3.10.

Hydrostatic Testina of Class 3 Pressure Retainino Components,.in the Coolina Water System Code Reauirement: Table IWD-2500-1, Examination Category D-C, Item D3.10, requires a system hydrostatic test as specified by IWD-5223. IWD-5223 states that the system hydrostatic test 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.

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

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

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' safeguards equipment is supported from both sides of the cooling 2-water system header. Consequently, the entire supply and return 1

header must be in operation at all times to meet operating

license requirements."

Licensee's Proposed Alternative Examination (as stated):

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 j for Class 3 pressure-retaining components. The licensee stated j that the subject lines are the only source of cooling water for the Cooling Water Supply and Return headers and would have to be

taken out of service to conduct the Code-required hydrostatic test. Since the plant operating license requires cooling water flow during both operation and refueling modes, taking these lines out of service would be a violation of the operating  ;

license. Therefore, the Code requirements are impractical for these lines. In lieu of this requirement, the licensee proposed i

to perform a system inservice test with an associated VT-2 visual examination, providing reasonable assurance of the system's operational readiness.

Conclusion:

Based on the above evaluation, it is concluded that the hydrostatic test of the subject portions of the cooling water

system is impractical to perform at Prairie Island, Unit 1. The j licensee's proposed alternative will provide reasonable assurance i

of the system's operational readiness. Therefore, it is

• recommended that relief be granted pursuant to

10CFR50.55a(g)(6)(1).

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3.5 General

, 3.5.1 Ultrasonic Examination Techniaues 3.5.1.1 Reauest for Relief No. 6. ASME Section XI. Anoendix I. l

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

Code Reauirement

Paragraph I-2100 requires ultrasonic -

l examination of vessel welds greater than 2-in. thickness to be 5

conducted in accordance with Article 4 of Section V.Section V,

Article 4, Paragraph T-441.1.2.1 requires the calibration blocks I

to be fabricated from one of the following:

(a) Nozzle drop out from the component;  !

(b) A component prolongation; i (c) Material of the same material specification, product  ;

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

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! Paragraph I-2200 requires ultrasonic examination of vessel welds less than or equal to 2-in. thickness to be conducted in

accordance with Appendix III. Appendix III, Paragraph III-3411 requires

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

(c) Where examination is to be performed from only one i side of the joint, the calibration block material i shall be of the same specification as the material on j that side,of the joint.

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

available, material of similar chemical analysis, tensile properties, and metallurgical structure may be used. j l Licensee's Code Relief Reauest: Relief is requested from material requirements of Section V, Article 4,

- Paragraph T-441.1.2.1(a,c,d), and Section XI, Appendix III, Paragraph 111-3411 for existing calibration blocks. ,

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t Licensee's Basis for Reauestino Relief (as stated):

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

The P-number grouping provides adequate assurance that the blocks 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 calibratjon blocks in order to satisfy the documentation requirements of the current Code. Existing records, indicate the appropriate P-number grouping, thereby providing adequate assurance that the blocks will establish the proper ultrasonic calibration and 4 sensitivity."

Licensee's Proposed Alternative Examination (as stated):

i " Existing calibration blocks will be used as is.

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

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

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

" Existing calibration blocks greater than 1" thick have been 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 indication is detected, a comparison will be made between the 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 material of similar chemical, tensile, and metallurgical properties. However, the calibration blocks at Prairie Island were constructed to the 1971 Edition of Section III which only required that they be of the same P-number grouping.

14

i t

l l, When using the existing calibration blocks that lack the ,

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.

The use of existing calibration blocks, fabricated as required by  :

i the original construction Code, will provide an acceptable method of establishing the proper ultrasonic calibration and

, sensitivity, provided the acoustical properties are similar to

{ those of the examination volume. Furthermore, requiring the l- licensee to replace all calibration blocks would impose a

considerable burden.

f.

i The existing blocks have been proven satisfactory for performing calibrations. Therefore, any increase in plant safety that might I occur with new blocks would not compensate for the burden placed

on the licensee to fabricate new calibration blocks to satisfy ,

f the current Code requirements.

I I

Conclusion:

Requiring the licensee to fabricate new l calibration blocks would result in a burden without a i

compensating increase in the level of quality and safety. The existing calibration blocks will provide an acceptable examination sensitivity provided the acoustical properties are verified to be similar to the examination area being examined.

Therefore, it is recommended that the proposed alternative be authorized with the above condition, pursuant to 10 CFR 50.55a(a)(3)(ii).

3.5.2 Exenoted Components 3.5.3 Other 15

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

I l

4. CONCLUSION ,

1

.. 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 cases of Requests for Relief Nos. 02 and I

05, Rev 2, (Part 2), the licensee has demonstrated that specific Section XI requirements are impractical; it is therefore recommended that relief be J

) granted as requested. The granting of relief will not endanger life, ,

i property, or the common defense and security and is otherwise in the public

interest, giving due consideration to the burden upon the licensee, that could l result if the requirements were imposed on the facility.

4 4

Pursuant to 10 CFR 50.55a(a)(3)(1), it is concluded that for Request for

{ Relief No. 05, 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 f alternative be authorized. l

) Pursuant to 10 CFR 50.55a(a)(3)(ii), it is concluded that for Request for l Relief No. 06 the licensee has demonstrated that specific Section X1 4

requirements would result in hardship or unusual difficulty without a i compensating increase in the level of quality and safety. In this case, it is recommended that the proposed alternative be authorized, only if the licensee l i satisfy the conditions stated in the above request for relief evaluation. 1 l

Requests for Relief Nos. 01, 03, Rev 1, and 04 were withdrawn by the licensee, and deleted from the ISI Program Plan by letter dated March 28, 1995, in i response to the NRC's request for additional information, and by letter dated ,

July 6,1995, in response to the NRC's conference call.

I j This technical evaluation has not identified any practical method by which the j licensee can meet all the specific inservice inspection requirements of 1

l Section XI of the ASME Code for the existing Prairie Island Nuclear Generating Plant, Unit 1. Compliance with all of the Section XI examination requirements j would necessitate redesign of a significant number of plant systems,

, procurement of replacement components, installation of the new components, and j performance of baseline examination for these components. Even after the l

l 16 '

1 j redesign efforts, complete compliance with the Section XI examination

' j 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. i i

j The licensee should continue to monitor the development of new or improved <

l examination techniques. As improvements in these areas are achieved, the 1 l licensee should incorporate these techniques in the ISI program plan I examination requirements. -

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

, 10-Year Interval Inservice Inspection Progran Plan, Revis1on 1, the licensee's

! response to the NRC's request for additional information, and the

. recommendaticas for granting relief from the ISI examinations that cannot*be

performed to the ext.ent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified. I i

t i

4 17

'w

L , . .

5. REFERENCES

1. Code of Federal Regulations Title 10, Part 50.

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

i 1989 Edition 1974 Edition Summer 1975 Addenda

3. Prairie Island Nuclesr Generating Plant, Third 10-Year Interval Inservice.

Inspection Program Plan, Revision 0, dated August 5, 1995.

l

4. NUREG-0800, Standard Review Plan for the Reviaw of Safety Analysis '

Reports for Nuclear Power Plants, Section 5.2.4, " Reactor Coo 1 ant Boundary Inservice Inspection and Testing," and Section 6.6, " Inservice

Inspection of C1 ass 2 and 3 Components," July 1981. l

, 5. Letter, dated January 20, 1995, Charles R. Thomas (NRC) to Roger 0.  ;

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

6. Letter, dated March 28, 1995, Roger 0. Anderson (Northern States Power) l to Document Control Desk (NRC), containing the response to NRC request for additional information.

1

7. Prairie Island Nuclear Generating Plant, Third 10-Year Interval Inservice  ;

i Inspection Program Plan, Revision 1, dated July 6, 1995, i

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

l

9. NRC Reguintory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds a

During Preservice and Inservice Examinations, Revisio'n 1, February 1983.

10. NRC Regulatory Guide 1.14. Reactor Coolant Pump flywheel Integrity, f Revision 1, August 1975.

i

11. Supplemental Reply to a Notice of Deviation NRC inspection Report Nos.

282/92008 and 306/92008 Final Safety Analysis Report Commitment for Inservice Examination of Nigh Energy Line Piping, dated February 10, 1993.

l

i J

a 4

I e

i 18

. **s g .:.0 an u s muetena ascutaroav coeuesno~ . g;ggi, ,

Eo S BIBLIOGRAPHIC DATA SHEET

~~ " '

i .

as ..,,,- ..,, . a .-,. .

rir6e 4~o suaritte INEL-95/0547 Technical Evaluation Report on the Third 10-Year

. Interval Inservice Inspection Program, Plan: 3 cars as, car ,uis3 i:

i Northern States Power Company -o r- .m Prairie Island Nuclear Generating Plant, Unit 2 November 1994

, Docket Number 50-282 om on caa~r suvesa FIN-L2556 (T0 77)

$ AurHoRt$) 6 rvPE of REPomi Technical

~

M. T. Anderson, K. W. Hall, A. M. Porter utnico Covento,, .,

ee.n.onMm.,o.on o. o"-- a . u.s ~-- a+--* c- .a. a a - a* - o'--=< - -

~ . ->omearios - =4us amo aoon ess u, =ac.,, .

INEL/LITCO P.O. Box 1625 Idaho Falls, ID 83415-2209 e stoyoagoasuzariou - =4=s ano 4ooness <,, c. ,, w- . .<, ,--,-. ~ac o. o --a . us a -a. , c-Civil and Geosciences Branch Office of Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission .

Washington, D.C. 20555 l

10. SuPPLEMeNTAmy 8 sores 4

11. AB5TR ACT (Jac e er ma>

1 This report presents the results of the evaluation of the Prairie Island Nuclear I Generating Plant, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan,  !

submitted August 5,1994 including the requests for relief from the American Society l of Mechanical Engineers Boiler and Pressure Vessel Code Section XI requirements that the licensee has determined to be impractical. The Prairie Island Nuclear Generating Plant, Unit 1, 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 1

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

l

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15. NUMBER of PAGES 16 PRsCL haC 8Omu 234 42498