ML20111A647

Revised, Inservice Insp Program,Prairie Island Nuclear Generating Plant, Technical Evaluation Rept

ML20111A647
Person / Time
Site:	Prairie Island
Issue date:	09/07/1984
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20111A598	List: ML20111A647
References
CON-NRC-03-82-096, CON-NRC-3-82-96 SAI-84-1651-01, SAI-84-1651-1, SAI-84-1651-R01, SAI-84-1651-R1, TAC-53456, TAC-53457, NUDOCS 8501080057
Download: ML20111A647 (44)

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T SAI-84/1651 4-PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 and 2 0

INSERVICE INSPECTION PROGRAM TECHNICAL EVALUATION REPORT

,. Submitted to:

U. S. Nuclear Regulatory Comissio'n Contract No. NRC-03-82-096 Science Applications, Inc.

Idaho Falls, ID 83402 September 7. 1934 8501000057 841228 -

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CONTENTS INTRODUCTION . . . .-. . . . . . . . . . . . . . . . . . . . . . . . . 1 I. CLASS 1 COMPONENTS . . . . . . . . . . . . . . . . ... . . . . . . 3 A. Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . . 3

1. Relief Request No. 54, Reactor Vessel Support Lugs , Ca tegory B-H, I tem B8.10 . . . . . . . . . . . . . . 3 B. Pressurizer ......................... 5

1. Relief Request No. 66. Nozzle Inner Radii, Category B-D, Item B3.120 ................ 5 C. Heat Exchangers and Steam Generators . . . . . . . . . . . . . 7

1. Relief Request No. 45, Pressure Retaining Welds in Other Than' Reactor Vessels, Category B-B,

' Items B2.51 and B2.60 .................. 7

2. Relief Request No. 66, Steam Generator, Regenerative Heat Exchanger, and Excess Letdown Heat Exchanger 4 -

Nozzles, Category.B-D, Items B3.140 and B3.160 . . . . . . 9 D. . Piping Pressure Bounda ry . . . . . . . . . . . . . . . . . . . 11

1. Relief Request No. 50 (Unit 1 Only), Safety

-Injection- Low-Head Piping Welds, Category B-J, Item B9.11 . . . . . . . . . . . . . . . . . . . . . . . . 11 E . ' ' Pump Pres su re Bounda ry . . . . . . . . . : . . . . . . . . . . . 13

1. Relief Request'No. 63, Reactor Coolant Pump Casing Welds, Category B-L-1; and Pump Casings, . Category B-L-2, Items B12.10 and B12.20 . . . . . . . . . . . . . . 13 F. . Valve Pressure Boundary (no relief requests)

,- G. General ........................... 16

1. Relief Request No. 52, Support Components, Categories F-A, F-B, F-C-. . . . . . . . . . . . . . . . . 16 II. CLASS 2 COMPONENTS . . . . . . . . . . . . . . . . . . . . . . . . 18 ,

A. Pressure Vessels . . . . . . . . . . . . . . . . . . . . . . . 18

1. Relief Request 30. 45, RHR Heat Exchangers, Ca tegory C- A , I tem C1.10 . . . . . . . . . . . . . . . . .

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2. Relief Request No. 66, Main Steam and Feedwater Nozzles and Accumulator Nozzles, Category C-B.,

. Item C2.22 . . . . , . . . . . . . . . . . . . . . . . ... . 18 B. -Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

1. Relief Request No. 50 (Unit 1 Only), Piping and Supports, Categories C-F and C-C, Items C5.11, C 5 . 21, a nd C 3 . 2 0 . . . . . . . . . . . . . . . . . . . . . . 19

2. Relief Request No. 50 (Unit 2 Only), Piping and Supports, Categories C-F, C-C, F-A, F-B, and F-C, Items C5.11, C5.12, C5.21, C5.22, C5.31, and C3.20 . . . . . 22

-C. Pumps (no relief requests)

D. Valves (no relief requests)

E. General ............................ 25 1

1. Relief Request No. 52, Support Components,

, Categories F-A, F-B, and F-C . . . . . . . . . . . . . . . . 25 III. CLASS 3 COMPONENTS (no relief requests)

IV. PRESSURE TESTS . . . . . . . . . . . . . . . . . . . . . . . . . . .

26 A. General (no relief requests)

B. Class'1 System Pressure. Tests- ................. 26

1. Relief Request No. 60, Class -1 Piping Between 3I329 and VC-8-3, Category B-P, Items B15.50 and B15.51 .. 26 C. Class 2 System Pressure Tests ................. 28

1. Relief Request No. 29, Class 2 Piping, Category C-H, Items C7.10 and C7.20 ............ 28

2. Relief Request No. 68, Steam Generator Secondary Side, Ca tegory C-H, Item C7.20 . . . . . . . . . . . . . . . 30 D'. Class 3 System Pressure Tests ................. 32

1. Relief Request No. 28, Cooling Water Supply and Return Headers, Category D-A, Item D1.10 . . . . . . . . . . 32

2. Relief Request No.:30, Diesel Generator Air and

-Cooling Water Piping, Category 0-A, Item D1.10 . . .-. . . .

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-< . 3. Relief. Request N'o. 31, Diesel Cooling Water and Fuel Piping, Category D-A, Item D1.10 . . . . . . . . . . . 36 V. -GENERAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37

1. Relief Request No. 48, UT Procedures for Bolts and Studs . . . .

. . . . . . . . . . . . . . . . . . . 37

2. Relief Request No. 56, UT Calibration Blocks ....... 39 REFERENCES . . . . . . . . . ................... 40 4

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TECHNICAL EVALUATION REPORT INSERVICE INSPECTION PROGRAM PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 and 2 -

INTRODUCTION This report evaluates requests for relief from Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code

• sub-mitted to the Nuclear Regulatory Commission (NRC) by the licensee, Northern States Power Company (NSP), for the Prairie Island Nuclear Generating Plant, Units 1 and 2. The relief requests cover the second 120-month inspection inter-val starting December 16, 1983, for Unit 1 and December 21, 1984, for Unit 2.

The requests are based upon the 1980 Edition of Section XI with addenda through the Winter of 1981, as specified in the applicable revision of 10 CFR 50.55a.

The rest of this introduction summarizes (a) the scope of this report, (b) the previous review of relief requests (I) , and (c) the history of Prairie Island Nuclear Generating Plant, Units 1 and 2 since the earlier review'.

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_ The current revision to 10 CFR 50.55a requires that Inservice Inspection (ISI) programs be . updated each 120 months to meet the requirements of newer

' editions of Section XI. Specifically, each program is to meet the requirements (to the extent practical) of the edition and -addenda of the Code incorporated .

in the regulation by reference in paragraph (b) 12 months prior to the start of the current 120-month interval.

The regulation recognizes that the requirements of the later editions and addenda of the Code might not be practical to implement at facilities because of limitations of design, geometry, and materials of construction of compon-ents and systems. It, therefore, permits exceptions to impractical examination or testing requirements to be evaluated. Relief from these requirements can be granted, provided the health and safety of the public are not endangered, giving due consideration to the burden placed on the licensee if the requirements

• Hereinafter referred to asSection XI or Code.

were imposed. This report only evaluates requests for relief dealing with inservice examinations of components and with system pressure tests. In-

. service test programs for pumps.and valves (IST programs) .are being evaluated ,

separately. ,

Finally,Section XI of the Code provides for certain components and systems to be exempted from its requirements. In some instances, these exemp-tions are not acceptable to the Nuclear Regulatory Commission (NRC) or are only acceptable with restrictions. As appropriate, these instances are also discussed in this report.

In its previous Safety Evaluation Report dated November 14, 1980 II) ,

NRC evaluated relief requests for Prairie Island Nuclear Generating Plant, LUnits 1 and 2, covering the first 120-month interval. The previous evaluation was based on submittals from the licensee dated October k5,1976, for Unit 1( )

.and October 12, 1977, for Unit 2(3) An additional evaluation of a relief

. request related to pump casing welds was transmitted to the licensee on October 12,1983(4) . On December 22, 1983, NPS submitted a new ISI program for the second 120-month interval which superseded all previous transmittals ( }.

- The relief requests contained in the December 22, 1983, submittal were based u.pon the 198Q Edition of Section XI of the Code, with addenda through Winter 1981. The Code edition and inspection intervals were in accordance with the

, revision of 10 CFR 50.55a applicable at the time.

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Additional information was required to evaluate the revised NSP ISI plan, and a ' request for additional information was submitted to the licensee (6) . The licensee responded to the request by submitting additional infonnation, with-drawing Relief Request No.= 67 and providing revisions to some relief requests ( ).

The relief requests contained in Reference 5, along with revisions contained in h Reference 7, are evaluated in this report. All the relief requests are identical for both units except for Relief Request No. 50. Accordingly, all of the follow-ing evaluations apply to both units except for Relief Request No. 50, which is evaluated separately for each unit.

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I. CLASS 1 COMPONENTS A. Reactor Vessel

1. Relief Request No. 54, Reactor Vessel Support Lugs, Category B-H, Item B&.10 ,

Code Requi.ement Integrally welded attachments to the reactor vessel must be examined by volumetric or surface methods, as applicable,in accordance with IWB-2500-13, 14, and 15. Examination is limited to attachment welds joining the attachment to the pressure retaining membrane of the components and. where the attachment base material design thick-ness is 5/8 inch or greater. Weld buildup on nozzles that serve as supports is excluded. The examination includes essentially 100% of the length of the weld to vessel and the integral attachment weld to a cast or forged integral attachment to the vessel, as applicable.

One-hundred percent of the welding of each lug' on the vessel is included in the examination. Deferral of the examination to the end of the interval is not permitted.

Code Relief Request Relief is requested to defer inspection until near the end of the 10-year inspection period.

Proposed Alternative Examination

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- When the core barrel is removed from the reactor vessel, at or near the'end-of the inspection interval, the supports will be in-spected 100%.

Licensee's Basis for Requestin~g Relief As the result of the reactor vessel design, the two integrally welded supports are not accessible from the OD of the vessel. Ul tra-sonic examination through the vessel wall from the ID surface appears to be the only means of examination. This examination would require the core barrel to be removed to gain access to the vessel's I'D surface.

Evaluation The two integrally welded supports are not accessible for exami-nation from outside the reactor vessel and are only accessible for examination from inside the reactor vessel when the core barrel is removed. Considering the cost, radiation exposure, and potential plant downtime, it is not practical to remove .the core , barrel to

implement inspection of the support lugs. The core barrel will be removed at or near the end of the inspection interval and the licensee has committed to examine the support lugs at that time.

• Since there.is'no published history of reactor vessel support lug failure for this' or any other vessel design, deferral of the vessel support lug examination to the end of the interval should have no significant impact on plant safety. .

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical. It is further concluded that the alternative examination discussed will provide the necessary added assurance of structural reli-ability. Therefore, the following is recommended:

Relief should be granted for deferral of inspection of the reactor vessel support lugs to the end of the interval when the core barrel is removed.

References

. References 5 and 1.

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B. Pressurizer

'1. Relief Request No. 66, Nozzle Inner Radii, Category B-D, Item B3.120 Code Requirement-The nozzle-inside-radius section of Category B-D nozzles in the pressurizer must be examined volumetrically in accordance with IWB-2500-7(a)-(d) during each inspection interval. Category B-D nozzles L

include nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles, but exclude manways and handholes either welded to or integrally cast in the vessel. If the examina-4 tions are conducted from inside the component and the nozzle weld is examined by the straight-beam ultrasonic method from the nozzle bore, the remaining examinations required to be conducted from the shell may be performed at or near the end of each inspection interval.

Code Relief Request Relief is requested from the volumetric examination require-

, ;ments of the nozzle inner radii.

Proposed Alternative Examination LThe pressurizer spray nozzle may be susceptible to a thermal fatigue mechanism due to the potential for high cyclic temperature gradients; therefore, an attempt will be made to ultra-sonically examine these inner radius areas. If service defects. are detected by these examinations, the relief, surge, and safety nozzles shall be assessed for similar examinations. Meanwhile, if a more comprehensive technique is developed and qualified, it will be implemented.

e Licensee's Basis fo'r Requesting Relief The Code required volume will'not be examined, based on the following criteria:

!' (a) The pressurizer relief, surge, and safety nozzles do not t

L experience high cyclic temperature gradients during normal operation, therefore, the. conditions for pro-ducing a thermal fatigue mechanism are not applicable.

t (b) Presently, there is no comprehensive inspection. tech-nique available, nor guidance for such in the ASME Code, which would provide a conclusive assessment of the Code required volumes of the inner radii, particularly since no preservice results are available for comparison.

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(c) Upon consideration of the above factors, a best effort examination approach to these nozzle inner radius sec-tions is not consistent with st:ndard ALARA practices.

An estimated 8 to 10 man-rem exposure rate over the

, - interval, per unit, at the present radiation levels, woisld be ex'perienced in attempting to' perform su~ch inconclusive examinations. .

Evaluation As the licensee has stated, currently available equipment and procedures for examination of nozzle inner radius sections are limited and generally applied on a best-effort basis. The inner radius sections most prone to cracking are those subjected to severe thermal cycling, and it is appropriate to emphasize inspection of these areas. Accordingly, the licensee has proposed a reasonable program for examination of nozzle inner radius sec-tions on the pressurizer by implementing examination of the pressurizer spray nozzle which is subjected to thermal cycling.

The remaining nozzles will be examined if indications are found in the spray nozzles. In addition, the licensee has agreed to

' . broaden the scope of the inspections to include the other pressurizer nozzlesif suitable examination techniques become available.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the

. - areas discussed above, the Code requirements are impractical. It is further concluded that the alternative examination discussed will provide the necessary added assurance of structural reliability.

Therefore, the following is recomended:

Relief should.he granted from complete volumetric examination of pressurizer nozzle inner radii in accordance with IWB-2500-7, provided that:

(a) Best effort volumetric examinations of the pressurizer spray nozzle inner radii are conducted.

(b) The remaining pressurizer nozzles are examined if indica-tions are detected in the spray nozzle.

References References 5, 1, 6, and 7.

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C. Heat'Exchangers and Steam Generators

1. Relief Request No. 45, Pressure Retaining Welds in Other Than Reactor Vessels, Category B-B, Items B2.51 and B2.60 Code Requirement Vessels 2 inches thick and over shall be examined in accordance with Article 4 of Section V as amended in IWA-2232.

Code Relief Request Relief is requested to use the ultrasonic inspection procedure for pipe welds instead of the heavy wall vessel examination procedure for thin wall vessels. Specifically, relief is requested to use the procedures in Appendix III of Section XI for examination of vessels fabricated from piping components rathar than Article 4 of Section V.

Proposed Alternative Examination The examination procedures will comply with Appendiic III of the

-* 1980 Edition through the Winter 1981 Addenda of ASME Section XI as they apply to ultrasonic examination of pipe welds.

Licensee's Basis for Requesting Relief

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The design service requireiaents for the regenerative heat exchanger and excess letdown heat exchangers resulted in the rela-tively small and thin wall vessels which pennitted them to be fabricated from piping components. Therefore, ultrasonic inspec-tion procedures for pipe welds would be more applicable than procedures for examination of heavy wall vessels. ,

Evaluation The licensee has proposed an alternate examination which is to ultrasonically examine the regenerative heat exchangers and excess letdown heat exchanger in accordance with Appendix III of the 1980 Edition through the Winter 1981 Addenda of ASME Section XI.

The NSP procedure for ultrasonic examination of pipe welds utilizes a minimum of 1-1/2 node metal path examination. The re-quired scanning area is defined as "the greater of 3t or 3 inches" from the toe of the weld on each side, to the extent practical, precluding any geometric limitations.

The Code required volume (CRV) for thin-walled components, as determined by Section XI, "the weld +1/2t either side,", will be more than covered by the NSP piping procedure.

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Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical. It

' is further concluded that the alternative examination discussed will provide the nece'ssary added assurance of structural r'eli-ability. Therefore, the following is recommended:

Relief should be granted for use of the pipe weld . inspection procedure based on Appendix III df Section XI,1980 Edition, with addenda through Winter 1981, for examination of the vessel welds in the regenerative heat exchangers and the excess letdown heat exchar.ger.

References References 5, 1, 6, and 7.

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2. Relief Request No. 66, Steam Generator Primary Inlet and Outlet and Regenerative Heat Exchanger Nozzles, Category B-D, Items 83.140 and B3.160 Code Requirement The nozzle-inside-radius section of Category B-D nozzles in the steam generator and regenerative heat exchanger must be examined volumetrically in accordance with IWB-2500-7 during each inspection interval. Category B-D r.ozzles include nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles, but exclude manways and handholes either welded to or integrally cast in the vessel. If the examinations are conducted from inside the com-ponent and the nozzle weld is examined by the straight beam ultrasonic method from the nozzle bore, the remaining examinations required to be conducted from the shell may be performed at or near the end of each inspection interval.

Code Relief Request Relief is requested from the volumetric examination. requirements

, of the nozzle inner radii.

Proposed Alternative Examination The steam generator feedwater nozzles and the pressurizer spray nozzle may be susceptible to a thermal fatigue mechanism due to the

' potential for high cyclic temperature gradients; therefore, an attcmpt will be made to ultrasonically examine these inner radius areas. If service defects are detected by these examinations, steam generator primary inlet and outlet nozzles and the regenerative heat exchanger nozzles shall be assessed for similar examinations. Meanwhile, if a more comprehensive technique is developed and qualified, it will be .

unplemented.

Licensee's Basis for Requesting Relief Relief from examining the Code required volume is requested, based upon the following criteria:

(a) The steam generator primary inlet and outlet and regenera-tive heat exchanger nozzles do not experience high cyclic temperature gradients during normal operation; therefore, the conditions for producing a thermal fatigue mechanism are not applicable.

(b) Presently, there is no comprehensive inspection technique available, nor guidance for such in the ASME Code, which would provide a conclusive assessment of the Code required volumes of the inner radii, particularly since no preservice results are available for comparison.

(c) Upon consideration of the above factors, a best effort examination approach to these nozzle inner radius sec-

. tions is not consistent with standard ALARA practices.

An estimated 8 to 10 man-rem exposure ' rate over the interval, per unit, at the present radiation levels would be experienced in attempting to perform such inconclusive examinations. .

Evaluation As the licensee has stated, currently available equipment and procedures for examination of nozzle inner radius sections are limited and generally applied on a best effort basis. The inner radius sections most prone to cracking are those subjected to severe themal cycling, and it is appropriate to emphasize in-spection of these areas. Accordingly, the licensee has proposed a reasonable program for examination of nozzle inner radius sections on the steam generator and pressurizer. The remaining steam gen-erator and regenerative heat exchanger nozzles will be examined if indications are found in inspected nozzles. In addition, the licensee has agreed to broaden the scope of the inspections to include the other nozzles if suitable examination techniques

.. become available.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the areas discussed above, the Code requirements are impractical. It is.further concluded that the alternative examination discussed will provide the necessary added assurance of structural reliability.

Therefore,the following is recommended.

• Relief should be granted from complete volumetric examination of steam generator primary inlet and outlet and regenerative heat ex -

. changer nozzle inner radii in accordance with IWB-2500-7, provided that:

(a) best effort volumetric examinations of the steam generator feedwater and pressurizer spray nozzle inner radii are conducted (b) the remaining steam generator primary inlet and outlet and regenerative heat exchanger nozzles are examined if indica-tions are detected in the steam generator nozzle.

I References References 5, 1, 6, and 7.

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D. Piping Pressure Boundary

1. Relief Request No. 50 (Unit 1 Only), Safety Injection Low-Head Piping Welds, Category B-J, Item B9.11

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Code Requirement -

For circumferential welds with nominal pipe size 4 inches and greater, surface plus volumetric examinations in accordance with IWB-2500-8 shall be performed during each inspection interval, and shall include the following:

(a) All terminal ends in each pipe or branch run connected to vessels.

(b) All tenninal ends and joints in each pipe or branch run connected to other components where the stress levels ex-ceed the following limits under loads associated with specific seismic events and operational conditions.

• (1) primary plus secondary stress intensity range of 2.4 Sm for ferritic steel and austenitic steel, and (2) cumulative usage factor U of 0.4.

(c) All dissimilar metal welds between combinations of:

(1) carbon or low alloy steels to high alloy steels, (2) carbon or low alloy steels to high nickel alloys, and (3) high alloy steels to high nickel alloys.

(d) Additional piping welds so that the total equals 25% of -

the circumferential joints in the reactor coolant piping system. This total does not include welds excluded by IWB-1220. These additional welds may be located in one loop (one loop is currently defined for both PWR and BWR plants in the 1980 Edition).

Code Relief Request Relief is requested from the examination requirements for cir-cumferential pipe welds in the safety injection low-head piping.

Proposed Alternative Examination None.

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Licensee's Basis for Requesting Relief The piping is imbedded in concrete.

Evaluation Access to volumetrically and/or surface examine these welds is restricted by not having access to the outside ~ surface due to con-

~ crete. Alternatively, the area surrounding the inaccessible welds should be visually eiamined for leakage after a'4-hour hold at the pressure test requirements. .

Conclusions and Reconmendations Based on the above evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical.

It is further concluded that the alternative examination specified below will provide- the necessary added assurance of structural reliability. Therefore, the following is recommended:

Relief should be granted from complete volumetric examination of safety injection low-head piping welds in accordance with IWB-2500-8, provided that:

All welds identified above as being inaccessible shall be visually inspected for leakage by observing the general area after

.. a 4-hour hold at the pressure test requirements as stated in IWB-5000.

References References 5, 1, 6, and 7.

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E. Pump Pressure Boundary

1. Relief Request No. 63, Reactor Coolant Pump Casing Welds, Category

, B-L-1; and Pump Casings, Category B-L-2, Itens B12.10 and B12.20 Code Requirement Essentially 100% of the weld length of all the pump casing welds in one pump in each group of pumps performing similar functions in the system must be volumetrically examined in accordance with IWB-2500-16 during each interval. A supplementary surface examination of the pump Gasing welds may be performed as required in IWB-3518.1(d). Visual examination (VT-3) of the internal surfaces in one pump in each group of pumps performing a similar function in the system is also to be implemented in each interval . The visual examination may be performed on the same pump selected for volumetric examination of the welds.

Code Relief Request Relief is requested from volumetric examination of the casing

• welds and visual examination of the internal surfaces for the reactor coolant pump.

Proposed Alternative Examination As an alternate to the B-L-1 and B-L-2 examinations, NSP will do the following:

(a)-Visually inspect the exterior of the pump casing during the hydrostatic pressure tests required by IWB-5000.

(b) Perform a surface examination of the external surface ,

of the welds to the extent practicable.

(c) If maintenance or operational problems are encountered which require the disassembly of the pump, the pump's interior surface will be visually inspected. The need for performance of a volumetric examination will also be evaluated at that time.

Licensee's Basis for Requesting Relief The licensee provided the following reasons as justification for the requested relief:

1. The radiation exposure for inservice inspection would raore than double due to the pump inspection alone. The ISI radiation exposure for 1980,1981., and 1982 were 42.2, 43.9, and 40.8 man-rem. Radiation exposure at other plants for the pump inspection ranged from 35 to 100 man-rem. A plant recently completed an inspection on the

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same model pump as that of Prairie Island, during which the exposure was 46 man-rem,10 of which were received to obtain a second radiograph. The fi.rst radiograph was not acceptable. -

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2. Additional exposure is received by personnel from' the move-ment of the upper internals. The upper internaJs need to be placed in the reactor to minimize exposure and airborne contamination when the cavity and reactor coolant system is drained for the pump casing inspection. The upper inter-i nals need to be. removed again for core reload.

3. The visual and/or volumetric examination will require com-plete disassembly of the pump. The. pump manufacturer (Westinghouse) does not require or recomend pump disassembly to perform normal maintenance and inspections. There has been limited experience for personnel doing this task.

Therefore, significant damage or degradation of the pump 8.

may result.

4. The estimated cost for the disassembly, inspection and assembly is approximately $500,000. This cost does not

• include additional loss in revenue if the outage is ex-tended due to the inspection.

5. A visual inspection was performed on one RCP in 1982 when disassembled for repairs. The internal surface was visu-ally inspected using an underwater TV camera. The pump-casing was not drained. The inspection did not reveal any

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problems. The visual inspection was completed to the

, requirements of ASME Code Section XI. '

6. .The reactor coolant pumps at Prairie Island have additional monitoring equipment not originally supplied with the pump.

The instruments monitor the shaft vibration, frame vibration,-

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thrust position, phase monitoring, and locked rotor. It was this instrumentation which alerted the plant personnel to the 21 RCP problem in 1981.

7.. The' reactor coolant pump casing consists of two cast rings E made from Type 316 stainless steel. This type of material b

is widely used in the nuclear industry and has performed

well.

'~ 8. EPRI is conducting a study of inspection frequencies for the inservice inspection program. A portion of that study is

' directed at the reactor coolant pumps. The preliminary find-ings indicate the interval for reactor coolant pumps casing weld inspection could be increased without significant risk.

EPRI will be issuing shortly a report on Research Project 2057. The report will discuss the reactor coolant pump

' casing weld ' inspection program. This report is titled "EPRI Report on Long-Tenn Inspection Requirements for Nuclear Power 3 Plants Components."

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Evaluation The reactor coolant pumps at Prairie Island Units 1 and 2 are i fabricated from two cast, Type 316 stainless steel rings joined to-

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gether by one circumferential weld. The weld -and internal surfaces of the pump are required to be volumetrically and visuallf examined, respectively. Volumetric examination of the weld by radiography and visual examination of the internal surfaces requires complete i disassembly of~the pump. The disassenbly, examination preparation, and reassembly of _ the pump would cause maintenance and examination personnel to be exposed to high levels of radiation for extended periods of time. Volumetric examination of the casing weld by

'.j ultrasonic would produce unacceptable results because of the high j ultrasound attenuation characteristics of cast material.

!= Conclusions and Recomm'ndations e Based on the above evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical.

It is further concluded that the alternative examination discussed i- will' provide the necessary added assurance of structural reli-i ability. Therefore, the following is recommended:

. Relief should be granted from volumetric examination of the pump casing welds and visual examination of the pump casing inter-nal surfaces, provided that

T (a) The pump casing exterior is visually inspected during.

! the hydrostatic test of the primary coolant system in

.. . . . accordance with IWB-5000.

(bfThepumpcasingweldsandheataffectedzonearesubjected

, to a surface examination over.100% of the weld length.

(c) The pump interior surfaces are examined if the pump is -

disassembled for maintenance.

f References

References 5, 1, and 4.

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F. Valve Pressure Boundary

. -NoLrelief requests.

G. General

1. Relief Request No. 52, Support Components, Categories F-A,' F-B, F-C Code Requirement

. Plate and shell type supports (F-A), linear type supports (F-B),

and component standard supports (F-C) shall be visually examined in accordance with IWF-1300-1 each inspection interval.

Code Relief Request

-Insulation will not be removed for complete examination of all supports.

Proposed Alternative Examination The insulation will be removed from a support compo.nent for

, further inspection whenever the connections and welds cannot be examined, or an abnormality is detected that may have been a result of a loss of support capability or inadequate restraint.

Licensee's Basis for Requesting Relief Any loss of support capability or inadequate restraints can us'ually be detected through the inspection of the uninsulated portion of the sVpport and the surrounding insulation. The governing codes .

and regulations used in the design and construction of those systems thet are now classified as Class 2 and 3 did not require provisions for inspection access for'these systems. .

Thus, it would be an undue burden without. compensating increase in safety to require insulation removal for support inspection.

Evaluation

The examination of supports to be conducted if relief is granted, will include all welds and mechanical connections for the required supports. Insulation will not be removed to examine support compon-ents which do not contain welds or mechanical connections. The insulation will be removed from a supported component for further >

inspections whenever an abnormality is detected that may have been a result of a loss of support capability or inadequate restraint.

This-approach should assure an adequate examination of support components.

5

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the supports discussed above, the Code requirements'.are impractical.

It is further concluded that the alternative examination discussed will provide the necessary added assurance of structural reli-ability. Therefore, the following is recommended: ,

Relief should be granted from complete visual examination of the Class 1 supports in accordance with IWF-1300-1, provided that:

The insulation must be removed sufficient to allow inspection of all mechanical connections, such as eyelets, bolts, adjustments, and locking devices. Any welds which might be on the support also require insulation removal to allow direct visual inspection of the weld.

References References 5 and 1.

e e b

G e

17-

~

II. CLASS 2 COMPONENTS A. Pressure Vessels

1. Relief Request No. 45, RHR Heat Exchangers, Category C-A, Item C1.10 This request for relief is essentially the same as previously discussed in Section I.C.1. Accordingly, based on the previous evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical. It is further concluded that the alternative examination discussed previously will provide the necessary added assurance of structural reliability. There-fore, the following is recommended:

Relief should be granted for use of the pipeweld inspection procedure based on Appendix III o'f Section XI,1980 Edition, with addeoda through Winter 1981, for examination of the vessel welds in the regenerative heat exchangers and the excess letdown heat exchangers.

References References 5, 1, 6, and 7.

2. Relief Request No. 66, Main Steam and Accumulator Nozzles, Category C-B, Item C2.22 This request for relief is essentially the same as previously discussed in Section I.B.1 and I.C.2. Accordingly, based on the previous evaluation, it is concluded that for the areas discussed ,

above, the Code requirements are impractical. It is further con-cluded that the alternative examination discussed previously will provide the necessary added assurance of structural reliability.

Therefore, the following is recommended:

Relief should be granted from complete volumetric examination of main steam and accumulator nozzle inner radii in accordance with IWB-2500-7, provided that:

The nozzles for which relief is granted are examined if indications are found during examination of the pressurizer spray nozzles or the stean generator feedwater nozzles.

References References 5, 1, 6, and 7. ~

. B. Piping

1. Relief Request No. 50 (Unit 1 Only), Piping and Supports, Categories C-F and C-C, Items C5.11, C5.21, and C3.20 t

Code Requirement .

For circumferential welds with nominal wall thickness greater than 1/2 inch (C5.21), surface plus volumetric examinations in ac-cordance with IWC-2500-7 shall be performed during each inspection interval for 100% of each weld. For circumferential welds with a nominal wall thickness less than or equal to 1/2 inch, only sur-face examination is required (C5.11). The examinations under both items shall include:

(a) all welds at locations where the stresses under the loadings resulting from Nonnal and l'pset plant conditions as calculated by the sum of Eqs. (9) and (10) in NC-3652 exceed 0.8 (1.2 Sn +

S );

A (b) all welds at terminal ends (see (e) below) of piping or branch runs;

.. (c) all dissimilar metal welds;

, (d) additional welds, at structural discontinuities (see (f) below) such that the total number of welds selected for examination includes the following percentages of circumferential piping welds:

(1) none of the welds exempted by IWC-1220,

- ~ (2) none of the welds in residual heat removal and emergency

, core cooling systems, (3)~10% of the main steam system welds 8-inch nominal pipe size and smaller,

' (4) 25% of the welds in all other systems.

(e) terminal ends .are the extremities of piping runs that connect to structures, components (such as vessels, pumps, valves), or pipe anchors, each of which act as rigid restraints or provide at least two degrees of restraint to piping thermal expansion; (f) structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, etc. conforming to ANSI B16.9),

and pipe branch connections and fittings; (g) exanination requirements are under development.

The welds initially selected for examination shall be re-examined over the service lifetin:e of the piping component. For welds in carbon or low alloy steels, only those welds showing reportable preservice transverse indications need to be examined for transverse reflectors. .

~

For integrally welded attachments on components required to be examined under Category C-F or C-G and whose base material design thickness is 3/4 inch or greater, 100% of the weld length must be examined by surface methods in accordance with IWC-

.. 2500-5. -

Code Relief Request Relief from examination of the following components is requested.

System Item Identification

'KAIN STEAM SYSTEM PIPING WELDS (ENCAPSULATED AT GUARD PIPE) 31-MS-2 Welds MS-160, 71, 72, 73, 74, 75, 76, 77, 78, 79 Welds MS-74 to 75, 76 to 77, 78 to 79 30-MS-2 Welds MS-68, 70, 159, 108 Welds MS-159 to 160 6-MS-2 Welds MS-108A, 134 31-MS-1 Welds MS-14 to 15 30-MS-1 Welds MS-51, 52W

. Welds MS-182 to 183 6 MS-1 Welds MS-SIC, 62 MAIN STEAM SYSTEM SUPPORTS (ENCAPSULATED AT GUARD PIPE) 31-MS-1 Supports I 30-MS-1 Supports J'

- - 31-MS-2 Supports I,J,K 30-MS-2 Supports E,F.G,L FEEDWATER SYSTEM PIPING WELDS (ENCAPSULATED BY GUARD PIPE) 16-FW-16 Welds FW-202, 203, 204, 225, 205, 206, 207, ~

208, 209, 210, 211, 212, 219, 213, 214 FEEDWATER SYSTEM SUPPORTS (ENCAPSULATED AT GUARD PIPE) 16-FW-16 Supports L. LL, N, 0, Q R, S CONTAINMENT SUMP B DISCHARGE PIPING WELDS (IMBEDDED IN CONCRETE) 14-51-33A Welds SI-11, 217, 12, 13 14-SI-34A Welds SI-14 14-SI-33B Welds SI-1, 217, 12, 13 14-51-24B Welds SI-4 CONTAINMENT SUMP B DISCHARGE PIPING WELDS (IMBEDDED IN CONCRETE) 14-SI-33A Welds SI-11, 217, I2, 13 14-SI-34A Welds SI-14 14-SI-33B Welds SI-4 .

14-S1-34B Welds SI-4

' ~'

System Item Identification CONTAINMENT SUMP B DISCHARGE SUPPORTS (IMBEDDED IN CONCRETE) 14-SI-33A Supports A,B,C 14-SI-33B Supports A,B,C Proposed Alternative Examination None.

Licensee's Basis for Requesting Relief The components specified are not accessible for examination.

Evaluation Access to volumetrically and/or surface examine these welds is restricted by not having access to the outside surface due to the interference from steel plate or concrete. Alternatively, the area surrounding the inaccessible welds should be visually examined for leakage after a 4-hour hold at the pressure test requirements. In addition, the encapsulated supports should be visually examined (VT-3 and VT-4) in accordance with IWF-2500-1. .

Conclusions and Reconnendations Based on the above evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical. It is further concluded that the alternative examination specified below will pro-vide the necessary added assurance of structural reliability. There-fore, the following is recomended:

Relief should be granted from complete volumetric and surface examination of the specified welds in accordance with IWC-2500, pro-vided that: '

(a) All welds identifed above as being inaccessible shall be visually inspected for leakage by observing the general area after a 4-hour hold at the pressure test requirements stated in IWB-5000 and IWC-5000. This examination, and other volu-metric inspections required by Section XI of similar systems which can be performed, will provide assurance that no deg-radation has occurred and the piping pressure boundary will remain structurally acceptable during the inspection interval.

(b) The encapsulated supports should be visually examined (VT-3 and VT-4) in accordance with IWF-2500-1.

References

. References 5, 1, 6, and 7.

2. Relief Request No. 50 (Unit 2 Only), Piping and Supports, Categories C-F, C-C, F-A, F-B, and F-C, Items C5.11, C5.12, C5.21, C5.22, C5.31, and C3.20 Code Requirement

~

For circumferential welds with nominal wall thickness greater than 1/2 inch (C5.11), surface plus volumetric examinations in ac-cordance with IWC-2500-7 shall be performed during each inspection interval for 100% of each weld. For circumferential welds with a nominal wall thickness less than or equal to 1/2 inch, only surface examination is required (C5.21). The examinations under both items shall include:

(a) all welds at locations where the stresses under the loadings resulting from Normal and Upset plant conditions as calculated by the sum of Eqs. (9) and (10) in NC-3652 exceed 0.8 (1.2 Sn +

SA )i (b) all welds at terminal ends (see (e) below) of piping or branch runs; (c) all dissimilar metal welds;

.. (d) additional welds, at structural discontinuities (see (f) below) such that the total. number of welds selected for examination includes the following percentages of circumferential piping welds:

(1) none of the welds exempted by IWC-1220, (2) none of the welds in residual heat removal and emergency

, . . core cooling systems.

(3),10%ofthemainsteamsystemwelds8-inchnominalpipe size and smaller, (4)25%oftheweldsinallothersystems.

(e) terminal ends are the extremities of piping runs that connect to structures., components (such as vessels, pumps, valves), or pipe anchors, each of which act as rigid restraints or provide at least two degrees of restraint to piping thermal expansion; (f) structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, etc. conforming to ANSI B16.9),

and pipe branch connections and fittings; (g) examination requirements are under development.

For longitedinal welds in piping less than or equal to 1/2 inch n'ominal wall thickness (C5.12), a surface examination covering 2.5t at the intersecting weld shall be conducted in accordance with IWC-2500-7. For longitudinal welds in piping greater than 1/2 inch in nominal wall thickness (C5.22), both a surface and volumetric examination covering 2.St at the intersecting circumferential weld shall be conducted in accordance with IWC-2500-7. ,

For circumferential branch connection welds in piping greater than 4 inches nominal (C5.31), surface examinations covering 100%

• of each weld shall be conducted in accordance with IWC-2500-9 to

-13, inclusive.

The welds initially selected for examination shall b'e re-examined over the ' service lifetime of the piping components. .For welds in carbon or low alloy steels, only those welds showing reportable preservice transverse indications need to be examined for transverse reflectors.

For integrally welded attachments on components required to be examined under Categories C-F, C-G, C-C, F-A, F-B, and F-C and whose base material design thickness is 3/4-inch or greater,(C3.20), 100%

of the weld length must be examined by surface methods in accordance with IWC-2500-5.

Plate and shell type supports (F-A), linear type supports (F-B),

and component standard supports (F-C) shall be visually examined in accordance with IWF-1300-1 each inspection interval.

Code Relief Request

.. Relief from examination of the following components is requested.

System Item Identification

~

MAIN

~

STEAM SYSTEM PIPING WELDS (ENCAPSULATED BY GUARD PIPE) 31-2MS-1 Welds MS-19, MS-20 Welds MS-19 to MS-20 30-2MS-1 Weld MS-22 Welds MS-185B, MS-185D 6-2MS-1 Weld MS-33 -

-Welds MS-166, 92, 93, 94, 95, 96, 97, 98, 99, 117, 170 Welds MS-165 to MS-166, MS-95 to MS-96 MS-97 to MS-98, MS-99 to MS-117 Weld MS-988 30-2MS-2 Welds MS-88, 89, 90, 91, 165, 100 Welds MS-89 to MS-90 Welds MS-183C, MS-183A 6-2MS-2 ' Weld MS-11 MAIN STEAM SYSTEM PIPING SUPPORTS (ENCAPSULATED BY GUARD PIPING) 30-2MS-1 Support 0 30-2MS-2 Supports D,E,F,G,H 31-2MS-1 Supports L,M,N 31-2MS-2 Supports -I,K.L,M,Q FEEDWATER SYSTEM PIPING WELDS (ENCAPSULATED BY GUARD PIPE) 16-2FW-16 Welds FW-119, 120, 121, 122, 123, 124, 125, 126, 127, 185, 128, 129, 130W, 131, 132 Item Identification

. .. System._

FEEDWATER SYSTEM PIPING SUPPORTS (ENCAPSULATED BY GUARD PIPE) 16-2FW-16 Supports A,B,C,D,E,F,G,H CONTAINMENT SUMP A & B DISCHARGE PIPING WELDS (IMBEDDED IN CONCRETE)

. 14-2SI-33B Welds 1, 2, 3, 207 14-2SJ-34B Weld 4 I4-2St-33A Welds 13, 14, 15 14-2SI-34A Weld 16 -

CONTAINMENT SUMP A & B DISCHARGE (IMBEDDED IN CONCRETE) 14-2SI-33B Supports A,B,C 14-2SI-33A Supports A,B,C Proposed Alternative Framination None.

Licensee's Basis for Requesting Relief

. The components specified are not accessible for examination.

Evalua tion Access to volumetrically and/or surface examine these welds is restricted by not having access to the outside surface due to the interference from steel plate or concrete. Alternatively, the area surrounding the inaccessible welds should be visually examined for leakage after a 4-hour hold at the pressure test requirements. In addition, the encapsulated supports should be visually examined

(%T-3 and VT-4) in accordance with WF-2500-1.

Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical. It is further.

concluded that the alternative examination specified below will provide the necessary added assurance of structural reliability. Therefore, the following is recomended:

v ','

Relief should be granted from complete volumetric and surface examination of the specified welds in accordance with IWC-2500, pro-vided that:

(a) All welds identified above as being inaccessible shall be visually inspected for leakage by observing the general area

. after a 4-hour hold at the pressure test requirements stated in IWB-5000 and IWC-5000. This examination, and other volu-i metric inspections required by Section XI of similar systems which can be perfomed, vill provide assurance that no deg-radation has occurred and the piping pressure boundary will remain stru'cturally acceptable during the inspection interval.

(b) The encapsulated' supports are visually examined (VT-3 and VT-4) in accordance with IWF-2500-1.

References__

References 5, 1, 6, and 7. '

l '

C. Pumps No relief requests.

D. Valves No relief Requests. -

E. General

1. Relief Request No. 52, Support Components, Categories F-A, F-B, and F-C This request for relief is tiie same as the request discussed in Section I.G.I. Accordingly, based on the previous evaluation, it is concluded that for the supports discussed, the Code require-ments are impractical. It is further concluded that the alternative examination discussed previously will provide the necessary added assurance of structural reliability. Therefore, the following is

, recommended:

Relief should be granted from complete visual examination of the Class 2 supports in accordance with IWF-1300-1, provided that:

The insulation is removed sufficient to allow inspection of all mechanical connections, such as eyelets, bolts, adjustments, a,nd locking devices. Any welds which might be on the support also re, quire insulation removal to allow direct visual inspection of the' weld.

References .

References 5 and 1.

i III. CLASS 3 COMPONENTS No relief requests.

IV. PRESSURE TESTS A. General No relief requests. -

8. Class 1 System Pressure Tests

1. Relief Request No. 60, Class 1 Piping Between 31329 and VC-8-3, Category B-P, Items B15.50 and B15.51 Code Requirement A system leakage test in accordance with IWB-5221 shall be conducted prior to startup following each refueling outage, and a system hydrostatic test shall be conducted at or near the end of each interval in accordance with IWB-5222.

.. Code Relief Request Relief from pressure testing this section of piping is requested.

Proposed Alternative Examination The section of piping will be given a surface examination eKch inspection interval.

Licensee's Basis for Requesting Relief =

This section of piping is not isolatable from the RCS. .

Performing a leakage test at functional pressure causes pressur-izer spray which causes a reduction in RCS pressure. Spraying water into the pressurizer from the auxiliary spray line is an abnormal operation. The spray line is designated for 10 such inadvertent operations.

Evaluation Because of the design of the Auxiliary Spray System, piping between the motor-operated valve #31329 and check valve #VC-8-3 cannot be pressurized to the proper test pressure without bypassing the check valve or opening the motor-operated valve. It is im-practical to pressurize this portion of the piping system at the frequency required by the Code because of the risk associated with the inadvertent operation of the pressurizer sprays. This section  !

of piping is also examined by surface methods in accordance with the rules of IWB-2000.

. Conclusions and Recommendations Based on the above evaluation, it is concluded that for the piping discussed above, the Code requirements are impractical.

It is further concluded that the alternative examination dis-cussed will provide the necessary added assurance of structural reliability. Therefore, the following is recommended:

Relief should be granted from pressure testing the spray system piping in accordance with IWB-5221 and -5222 provided that:

Surface examination of 100% of the piping welds is conducted in accordance with IWB-2000.

References References 5 and 1.

I e e a

6 9

9

. C. Class 2 System Pressure Tests

1. Relief Request No. 29, Class 2 Piping, Category C-H, Items C7.10 and C7.20 1

Code Requirement

• A system leakage test in accordance with IWC-5221 shall be conducted each period, and a system hydrostatic test in accordance with IWC-5222 shall be conducted at or near the end of each interval.

Code Relief Request Relief is requested from testing the following piping at the pressures required by IWC-5000.

Components:

Safety Injection Piping unisolatable from Class 1 Piping (NF-39813)

Reactor Coolant System Piping 3/4" and smaller that is

.- unisolatable from Class 1 Piping (NF-39807)

Residual Heat Removal System Piping unisolatable from Class 1 Piping (hF-39813)

RCP Seal Injection Piping 3/4" and smaller that is uniso-latable from Class 1 piping (NF-39809)

, , , RCP Seal Return Piping unisolatable from Class 1 (NF-39809)

Charging Line Piping unisolatable from Class 1 (NF-39809)

Sample System Piping unisolatable from Class 1 (NF-39807)

Proposed Alternative Examination The piping will.be tested to the Class 1 requirements, i.e.:

1. The unisolated portions of the Class 1 piping will be visually examined for evidence of leakage at the system nominal operating pressure in accordance with the require-ment of IWB-5221. This inspection will be performed prior to startup following each reactor refueling outage.

2. .The unisolated portions of the Class 2 piping will be hydrostatically tested when the Class 1 piping is tested.

Licensee's Basis for Requesting Relief The piping is not isolatable from the C-lass 1 piping.

Evaluation The Class 2 piping specified cannot be isolated from Class 1 piping for pressure testing. The licensee proposes to presst're

, test the Class 2 piping at the same time the Class 1 piping is pressure tested. Depending on the design temperature of the Class 2 piping, this would result in slightly reduced test pres-sures; however, the pressure test should still be adequate to confirm the structural integrity of the system. j Conclusions and Recommendations Based on the above evaluation, it is concluded that for the pressure tests discussed above, the Code requirements are impracti-cal. It is further concluded that the alternative examination discussed will provide the necessary added assurance of structural reliability. Therefore, the following is recommended:

Relief should be granted from pressure testing the specified Class 2 piping in accordance with IWC-5000, provided that:

(a) The specified piping is pressure tested in accordance

. with the requirements of IWB-5000.

(b) The licensee performs a visual examination for evidence of leakage on those portions of the above systems at the system nominal operating pressure in accordance with the requirements of IWB-3221. This examination shall be per-formed prior to startup following each reactor refueling ou tage.

Re'ferences References 5'and 1.

t

-29

2. Relief Request No. 68, Steam Generator Secondary Side, Category C-H, Item C7.20 Code Recuirement -

A system hydrostatic test shall be conducted at oc near the end of each interval in accordance with IWC-5222.

Code Relief Request Relief is requested from using the 10-year hydrostatic test pressure as specified by IWC-5000.

Proposed Alternative Examination The steam generator will be tested in accordance with IWB-5000 requirements.

Licensee's Basis for Requesting Relief For the following reasons, the steam generator secondary sides, main steam line to the main steam isolation valves, the feedwater line inlets to the steam generators, the auxiliary feedwater inlet to the steam generators, and the steam generator blowdown lines from the steam generators (to the first isolation) are to be hydro-tested in accordance with Article IWB-5000 of the ASME Boiler and Pressure Vessel Code: .

1.- The maximum allowable secgndary to primary pressure differ-ential is 670 psig at 650 F. To avoid violation of the design differential, the Reactor Coolant System pressure would have to be elevated above 677 psig and to an over-pressure. condition.

2. The main steam safety valves would not require gagging, thereby precluding any chance of overpressurizing the steam generator.

3. Since the steam generator is integrally tied to the Reactor Coolant System, it is logical as well as practical to test

" them at hot shutdown per IWB-5000 of the Code. This allows a hot hydro in lieu of cold hydro.

Evaluation Hydrostatic testing of the steam generator secondary side and related piping would be done at 1.25 times the system pressure if implemented in accordance with IWC-5222. Pressurization of the secondary to this level would result in a differential pressure

between the primary and secondary side that was the reverse of i , norwal operation and in excess of the maximum allowable differ-ential by at least 7 psig. Excessive reverse pressure differen-tial in the steam generator tubes is not a desirable system test. Hydrostatic testing of the secondary system in conjunction with the primary sy' stem hydrostatic tests as proposed by~ the licensee is an acceptable alternate test provided that the visual examinations required by IWC-5222 are conducted.,

- Conclusions and Recommendations Based on the above evaluation, it is concluded that for the pressure -tests discussed above, the Code requirements are imprac-tical. It is further concluded that the alternative examination discussed will provide the necessary added assurance of structural reliability. Therefore, the following is recommended:

Relief should be granted from hydrostatic testing of the steam generator secondary piping in accordance with IWC-5222, provided that:

(a) The specified piping is hydrostatically tested in accordance with IWB-5000.

~

(b) The visual examinations required by IWC-5222 are e conducted.

- References References 5 and 1.

O 9

e

-. 4 w - s,>, w,, -e-, , re. -- w w g---~ ~ -m-we~w -->-e-r- es ,e w ,w-> w - ---n - <

m--r----- --rv,r--m,-,,,p-- --

---,,--vr--gm wy-- ww s-- - - *~-

, D. Class 3 System Pressure Tests

1. Relief Request No. 28, Cooling Water Supply and Return Headers, Category D-A, Item D1.10 Code Requirement .

A system hydrostatic test in accordance with IWD-5223 shall be conducted at or near the end of each interval.

Code Relief Request Relief is requested from hydrotesting the piping each inspec-tion interval as required by IWD-2410.

Proposed Alternative Examination The Cooling Water System will be visually examined by every one-third of each inspection interval for conditions adverse to system operation. Additionally, the system is in constant operation and any leaks would be immediately known. Portions that are iso-

.- latable from the main headers will be pressure tested in accordance with the applicable requirements.

Licensee's Basis for Requesting Relief The Cooling Water System design is such that Unit I and Unit 2

. safeguards equipment is supplied 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 requi/e-ments.

Evaluation

, The cooling water system is in continuous operation, serving both units in order to meet licensing requirements. The main headers cannot be isolated from the system for pressure testing.

!- Alternatively, the licensee has proposed that normal operational surveillance of the system would detect any leaks that developed.

The isolatable portions of the Cooling Water System will be hydro-statically tested as required by the Code.

Conclusions and Recomraendations Based on the above. evaluation, it is concluded that for the piping system discussed above, .the Code requirements are imprac-tical. It is further concluded that the alternative examination discussed will orovide the necessary added assurance of structural reliability. Therefore, the following is recommended:

Relief should be granted from hydrostatic testing of the cooling water supply headers in accordance with IWD-5000, pro-

. vided that:

. (a) Operational surveillance of the system is mainta.ined such that leaks will be detected.

(b) The headers are visually examined each one-thi~rd interval for conditions adverse to system operation.

References References 5 and 1.

g e

e e

e e

n I

2. Relief Request No. 30, Diesel Generator Air and Cooling Water Piping, Category D-A, Item D1.10

. Code Requirement .

A system hydrostatic test in accordance with IWD-5223 shall be conducted at or near the end of each interval. -

Code Relief Request Relief is requested from hydrotesting portions of the Class 3 piping each inspection interval as required by IWD-2410. The Starting Air, Air Intake, and Cooling Water Piping associated with 11 and 12 diesel generator' (NF-39822) are the specific systems affected.

Proposed Alternative Examination The piping will be visually examined by every one-third of each inspection interval for conditions adverse to system opera-tion. Additionally, the systems are in constant operation and any leaks would be immediately known. Portions that are isolatable

,i from the diesel generators will be pressure tested in accordance with the applicable requirements.

Licensee's Basis for Requesting Relief The piping is not isolatable from the diesel generators.

Evaluat' ion The diesel generator starting air, air intake, and cooling water piping cannot be isolated for hydrostatic testing. The licensee has proppsed to visually examine the piping every one-third of each inspection interval for conditions adverse to system operation. Additionally, the systems are in constant operation and any leaks would be known. Portions that are isolatable from the diesel generators will be pressure tested in accordance with the applicable requirements.

Conclusions and Recmanendations Based on the above evaluation, it is concluded that for the piping systems discussed above, the Code requirements are imprac-tical. It is further concluded that the alternative examination discussed will provide the necessary added assurance of structural-reliability. Therefore, the following is recommended:

Relief should be granted from hydrostatic testing of the specified piping in accordance with IWD-2410, provided that:

(a) Operationa1 surveillance is naintained such that leaks will be detected.

(b) The piping is visually examined each one-third interval for conditions adverse to system operation.

References References 5 and 1.

m o c#

D e

D

~-

3. Relief Request No. 31, Diesel Csoling Water and Fuel Piping, Category D-A, Item D1.10

. Code Requirement .

A system hydrostatic test in accordance with IWD-5223 shall be conducted at or near the end of each interval.

Code Relief Request Relief is requested from inspecting the fuel oil piping (visual examination or pressure test) as required by IWD-2410 (IWD-5223).

Proposed Alternative Examination None.

Licensee's Basis for Requesting Relief

. The tanks and most of the piping are underground and not accessible for testing and inspection. Any leakage from the fuel oil storage tanks will be detected during daily checks of the storage tanks levels. Also, each tank is annually tested for i

moisture content to further verify its integrity. Monthly checks of the diesel generator and diesel cooling water pump day tank levels and day tank alarms will indicate any problems in the

. fuel oil transfer piping system.

Evaluation i

The fuel oil storage tanks and most piping are underground ,

and therefore inaccessible for examination. Operational surveil-I lance by the licensee on a daily, monthly, and yearly basis should provide adequate monitoring of the fuel systems.

Conclusions and Reconmendations

. Based on the above evaluation, it is concluded that for the fuel tanks and piping discussed above, the Code requirements are impractical. .It is further concluded that the operational moni-

- toring discussed will provide the necessary added assurance of structural reliability. Therefore, the following is recommended:

Relief should be granted frem hydrostatic testing of diesel fuel tanks and piping in accordance with IWD-5223.

References References 5 and 1.

)

V. GENERAL

1. Relief Request No. 48, UT Procedures for Bolts and Studs

~

Code Requirement Ultrasonic examinations shall be performed in accordance with Article 5 of Section V when the provisions of AppendixIII~of Sec-tion XI do not apply.

Code Relief Request Relief is requested to use the back reflection method for exami-nation of bolts and studs for the reactor coolant pump flange bolting.

Proposed Alternative Examination The items will be examined using the back reflection method correlated with an as-built sketch of the particular bolt or stud being examined. ASME Section XI will be used for evaluation criteria.

-* Licensee's Basis for Requesting Relief The Section V technique utilizing the calibration test bar was not used for the baseline examinations and is not as sensitive to detect discontinuities as the presently applied back reflection method.

~

~ A qualification program was initiated by NSP and documented in the PI-ISI submittal for the first 10-year inspection interval.

During the qualification test, it was demonstrated that the NSP-UT-4 procedure, which utilizes a back reflection technique for flaw evalu-ation, was a more sensitive examination than the technique specified in ASME Section V, Article 5, paragraph T-525.2.

The results indicated that at the same nominal metal path, the NSP procedure was approximately 6db more sensitive than the ASME technique. In addition to the percent-of-DAC reporting level, the NSP procedure dictates that any reflector, regardless of amplitude, which is accompanied by a 50% loss of back wall reflection must be reported / evaluated. As poorer end reflecting surfaces are encoun-tered, the NSP procedure tends to become a much more conservative approach to bolt and stud examination.

Evaluation The licensee has developed, implemented, and documented a back reflection UT method for examination of reactor coolant pump flange bolts and studs. The method was successfully used during the first ISI interval. The method appears to be suitable and its continued use provides for comparison with previous inspections.

s Conclusions and Recommendations Based o.' the above evaluation, it is concluded that for the examinations discussed above, the Code requirements are impracti-cal. It is further concluded that the alternative examination discussed will providi the necessary added assurance of structural reliability. Therefore, the following is recomended: ,

Relief should be granted provided that use of the NSP-UT-4 back reflection method for examination of reactor coolant pump studs and bolts.is verified by the resident inspector.

Referer.ces References 5, 1, 6, and 7.

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2. Relief Request No. 56, UT Calibration Blocks Code Requirement When using Appendix III of ASME Section XI, Winter 1982 Addenda, the basic calibration blocks shall be made from material of the same nominal diameter as those to be examined.

Code Relief Request The licensee requests relief to use flat calibration blocks for pipes greater than 20-inches in diameter.

Proposed Alternative Examination For surface curvature, the rules of Article 5 of Section V, 1980 Edition through Winter 1981 Addenda, will apply for examination of pipe welds and welds in components fabricated from piping. In addition, the other requirements of Appendix III basic calibration blocks will be met.

Licensee's Basis for Requesting Relief A flat basic calibration block gives the same results as a block essentially the same curvature for components greater than 20-inches in diameter. Any difference in accuracy and sensitivity for ultrasonic examination when using a flat basic calibration block versus a curved-basic calibration block for components greater

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than 20 inches in diameter is within the accuracy of the test. NSP belleves ,that compliance with Appendix III requirements for basic calibration block curvature would be an undue burden with no increase in public safety.

Evaluation The rules provided in Section V for surface curvature are acceptable for examination of piping greater than 20 inches in diameter.

Conclusions and Reconnendations Based on the above evaluation, it is concluded that for the examinations discussed above, the Code requirements are impractical.

It is further concluded that the alternative examination discussed will provide the necessary added assurance of structural reliability.

Therefore, the following is recommended:

Relief should be granted to use Section V, Article 5 of the 1980

[ Edition, through Winter 1981 Addenda for curvature of calibration blocks. -

! References l References 5 and 1.

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l. REFERENCES

1. R. A. Clark (NRC) to L. O. Mayer (NSP), Safety Evaluation Report by

. the Office of Nuclear Reactor Regulation Related to Amendment No. 43 to Facility Operating License No. DPR-42 and Amendment No. 37.to Facility Operating License-No. DPR-60, Northern States Power Company, Prairie Island Nuclear Generating Plant, Unit No's. I and 2. Docket No's. 50-582 and 50-306, November 14, 1980.

2. Transmittal letter (NSP) to (NRC), Inservice Inspection Technical Specifi-cations, Unit No. 1, October 15, 1976.

3. . Transmittal letter (NSP) to (NRC), Inservice Inspection Technical Specifi-cations, Unit No. 2, October 12, 1977.

4. J. R. Miller (NRC) to D. M. Musolf (NSP), Relief Request from the Inservice Inspection of Reactor Coolant Pump Casing Welds - Prairie Island Nuclear Generating Plant Units 1 and 2, October 12, 1983.

5. D. M.~Musolf (NSP) to Director (NRR), Submittal of the 2nd 10-Year Inservice Inspection and Testing (ISI/IST) Program, Prairie Nuclear Generating Plant, December 22, 1983.

. 6. Request for Additional Information, March 23, 1984.

7. D. M. Musolf (NSP) to Director (NRR), Response to NRC Request for Supple-mental Information on the Second Ten-Year ISI Program Request for Relief, June 11, 1984.

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NA-182, the expansion of the spent fuel storage capacity to accommodate both NA-1&2 and 500 Surry spent fuel assemblies will not create any significant additional radiological effects. The additional total body dose that might be received by an individual at the site boundary and the estimated dose to the total body of the population within a 50-mile radius of the plant is less than 0.1 mrem per year and 0.1 person-rem per year, respectively. These doses are extremely small compared to the fluctuations in the annual dose this t population receives from background radiation. This population dose represents an increase of less than 1 percent of the dose previously evaluated in the FES 'for NA-1&2. The occupational radiation dose to the work force engaged in the modification of the spent fuel storage racks (including present rack disposal) and the loading / unloading of 500 Surry spent fuel assemblies is estimated by the licensee to be 31 person-rem. This is a small fraction of the total person-rems from occupational dose at NA-1&2. The small increase in

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radiation dose should nct affect the licensee's ability to maintain individual occupational dose within the limits of 10 CFR Part 20, and as low as reasonably achievable. Finally, pursuant to 10 CFR 51.52, the radiological impact to the environment related to the transshipment of 500 Surry spent fuel assemblies from Surry to NA-1&2 is well within the scope of Table S-4, and is

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therefore acceptable.

8.0 Basis and Conclusion for Not Preparing an Environmental Impact Statement The staff has reviewed this proposed facility modification relative to the requirements set forth in 10 CFP. Part 51 and the Council on Environmental Quality's Guidelines, 40 CFR 1500.6. Based on this assessment, we propose to find that the actions specified will not either separately or combined significantly impact on the quality of the human environment. These actions are:

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