PY-CEI-NRR-2310, Submits Inservice Exam Relief Requests for Second 120 Month Interval.Second Interval Will Start on 981118.Attachment 1 Provides Summary & History of Proposed Relief Requests for ISI Program

Submits Inservice Exam Relief Requests for Second 120 Month Interval.Second Interval Will Start on 981118.Attachment 1 Provides Summary & History of Proposed Relief Requests for ISI Program

ML20237F187
Person / Time
Site:	Perry
Issue date:	08/26/1998
From:	Myers L CLEVELAND ELECTRIC ILLUMINATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
PY-CEI-NRR-2310, NUDOCS 9809020113
Download: ML20237F187 (90)

Text

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Perry Nuclear Power Plant P ox 9 m

j Perry Ot.b 44081 l

NWW 440-280-5915

' Vice President Fax: 440u280-8029 August 26,1998 PY-CEl/NRR-2310L United States Nuclear Regulatory Commission Document Control Desk .

Washington,DC 20555 Perry Nuclear Power Plant

- Docket No. 50-440 Submittal ofInservice Examination Relief Requests for Second 120 Month Interval l

Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(gX4Xii), the Inservice Examination Program for the Perry Nuclear Power Plant is being updated to comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference into 10 CFR 50.55a.(b) twelve months prior to the start of the second 120 month interval. The second interval will start on November 18,1998.

Thus, the edition of the ASME Section XI inspection Code for the second interval is the 1989 edition. Based on this edition there are certain relief requests being submitted per 10 CFR 50.55a.(gX5Xiii). Attachment I provides the summary and history of the proposed relief -

requests for the inservice inspection program. Attachment 2 provides the detailed disenssion of the individual relief requests.

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This letter and its attachments do not contain egulatory commitments (relief request revisions

/

require NRC approval). If you have questions or require additional information, please contact Mr. Henry L. Hegrat, Manager - Regulatory Affairs, at (440) 280-5606. I Very truly yours, .

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Attachments ,v cc: NRC Project Manager NRC Resident inspector  ;

NRC Region 111  !

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. 9809020113 980826  !

l PDR. ADOCK 05000440 1 l 0 PDR g l

l- I c _ _ - - - _ - - _ _ _ - - - - - - _ - _ _ - _ _ _ _

J Attachment I  !

PY-CEl/NRR 2310 i Pageiof4 1 RELIEF REQUEST lilSTORY AND CilANGE

SUMMARY

FOR SECOND INTERVAL l l

I Relief Request Previous Submittal NRC Safety Change Summary for Submittal of Relief

& New Rev . Letters Evaluations Request for the Second Interval  !

Nu. i 1

IR-001, R-2 PY-CEl/NRR-0919L, TAC #61443, Relief request narrative was updated rao revised j i i/18/88 [Rev. 0] 04/25/90 [Rev. 0] for clarification. Also, componer- .nat can be l PY-CEl/NRR-1521 L, TAC #M84418, essentially 100% examined (i.e., at least 90%

07/13/92 [Rev.1] 02/24/94 [Rev.1] coverage) were removed from the table and PY-CEl/NRR-1552 L, completion percentages were adjusted.

I1/25/92 [Rev. Il i IR-002, R-1 PY-CEl/NRR-0919L, TAC #61443, Relief request narrative was updated and revised l l I/18/88 [Rev. 0] 04/25/90 [Rev. 0] for clarification. Also, a ligure was added to depict the coverage limitations.

IR-003, N/A PY-CEl/NRR-0919L, TAC #61443, IR-003 is not being re-submitted for the second ,

i 1/18/88 [Rev. 0] 04/25/90 [Rev. 0] interval. In accordance with Code Case N-460 i (approved by the NRC in RG 1.147), reliefis no I longer necessary as coverage is 90% or greater.

IR-004, R-2 PY-CEl/NRR-0919L, TAC #6l443 Relief request narrative was updated and revised 1 11/18/88 [Rev. 0] 04/25/90 [Rev. 0] for clarification. Also, components that can be PY-CEl/NRR-1334L, TAC #M75334, essentially 100% examined or are no longer  ;

03/19/91 [Rev.1] 02/14/92 [Rev.1] scheduled ihr examination were removed from the table.

IR-005, R-2 PY-CEl/NRR-0919L, TAC #61443, Relief request narrative was updated and revised i l/18/88 [Rev. 0] 04/25/90 [Rev. 0] for clarification.

PY-CEI/NRR-2068L, TAC #M95898, 06/28/96 [Rev.11 05/27/97 [Rev. Il l IR-006, N/A PY-CEl/N RR-0919L, TAC #61443, IR-006 is not being re-submitted for the second j 11/18/88 [Rev. 0] 04/25/90 [Rev. 0] interval. In accordance with Code Case N-460 i (approved by the NRC in RG l.147), reliefis no i longer necessary as coverage is 90% or greater.

IR-007, R-1 PY-CEl/NRR-0919L, TAC #61443 Relief request narrative was updated and revised iI/I8/88 [Rev. 0] 04/25/90 iRev. 0] for clarification. )

IR-008, N/A PY-CEl/NRR-0919L, TAC #61443, IR-008 is not being re-submitted for the second i 1/18/88 [Rev,0] 04/25/90 [Rev. 0] interval. In accordance with Code Case N-460 (approved by the NRC in RG 1.147), reliefis no longer necessary as coverage is 90% or greater.

IR-009, R-l PY-CEl/NRR-0919L, TAC #61443, Relief request narrative was updated and revised  !

!1/18/88 and 04/25/90 [Rev. 0] for clarification. Also, a ligure was added to PY-CEl/NRR-1208L. depict the coverage limitations.

08/10/90 [Rev. 0) l IR 010, N/A PY-CEl/NRR-0919L, TAC #61443, IR-010 is not being re-submitted for the second 11/18/88 [Rev. 0] 04/25/90 [Rev. 0] interval. In accordance with Code Case N-460 (approved by the NRC in RG 1.147), reliefis no l longer necessary as coverage is 90% or greater.

( IR-011, N/A PY-CEl/NRR-0919L, TAC #61443, IR-01 I is not being re-submitted for tne second 11/18/88 [Rev. OJ 04/25/90 [Rev. 0] ~ interval. In accordance with Code Case N-460 (approved by the NRC in RG 1.147), reliefis no longer necessary as coverage is 90% or greater.

L____________________________-________

Attachment i PY-CEl/NRR-2310 Page 2 of 4 stelieritequest Previous Submittal NitC Safety Change Summary for Submittal of Relief

& New stev Letten s Es atuations Request for the Second Interval No.

IR-012, R-2 PY-CEl/N RR-0919L, TAC #61443, Relief request narrative was updated and revised i1/18/88 [Rev. OJ 04/25/90 [Rev. 0] for clariticatica. Also, components that can be PY-CEl/NRR- 1334L, TAC #M75334, essentially 100% examined or are no longer 03/19/91 [Rev.1] 02/14/92 [Rev. IJ scheduled for examination were removed from the table.

IR-013, R-1 PY-CEl/NRR-0919L, TAC #61443. Relief request narrative was updated and revised i I/t 8/88 [Rev. 01 04/25/90 l Rev. 0] for clarification.

IR-014, N/A PY-CE!/NRR-0919L, 'l AC #61443, IR-014 is not being re-submitted for the second I l/18/88 [Rev. 0] 04/25/90 [Rev. 0] interval. In accordance with Code Case N-460 (approved by the NRC in RG l.147), reliefis no longer necessary as coverage is 90% or greater.

IR-015, R- 1 PY-CEl/NRR-0919L, TAC #61443, Relief request narrative was updated and revised i I/18/88 lRev. 0] 04/25/90 [Rev. 01 for clarification.

IR-016 N/A Previously Withdrawn N/A N/A IR-017, N/A Previously Withdraw n N/A N/A IR-018, R-2 PY-CEl/NR R-1078L, No SE tor Rev. O Relief request narrative was updated and revised i 1/17/89 [Rev. 0] TAC uM75334, for clarification. Also, one component support s

PY-CEl/NRR- 1334 L, 02/14/92 [Rev, l j that can be essentially 100% examined was

~

03/19/91 IRev.1] removed from the table.

IR-019, R-l PY-CEl/N RR-1078 L, TAC #M75334, Relief request narrative was updated and revised i 1/17/89 [Rev. 0] 02/14/92 [Rev. 0] for clarification. Also, one component support that is no longer scheduled for examination was removed from the table.

IR-020 N/A PY-CEl/NR R- 1078L, TAC #M75334, IR-020 is not being re-submitted for the second i 1/17/89 [Rev. 0] 02/14/92 [Rev. 0] interval. In accordance with Code Case N-460 (approved by the NRC in RG 1.147), reliefis no longer necessary as cos erage is 90% or greater.

IR-021, R-4 PY-CEl/NRR-1078L, No SE for Rev. O Relief request narrative was updated and revised I i/I 7/89 [Rev. Oj TAC #M75334, ior clarification.

PY-CEl/N RR-1334L, 02/14/92 [Rev. IJ 03/19/91 [Rev. ll TAC #M84418, PY-CEl/N RR- 1521 L, 02/24/94 [Rev. 2j 07/13/92 [Rev. 2j TAC #M95898, PY-CEl/NRR-2068L, 05/27/97 [Rev. 3]

06/28/96 lRev. 31 1R-022, N/A Previously Withdrawn N/A N/A IR-023, R-l P Y-CEl/N RR-1078L, TAC # 61303 Relief request narrative was updated to reflect i 1/17/89 [Rev. 0] 09/07/90 relocation of snubber requirements from Technical Specifications to the Operational Requirements Manual and revised for clarification.

IR-024, R-1 PY-CEl/N RR- 1334 L, TAC #M75334, Relief request narrative was updated and revised 03/19/91 l Rev. 0 } 02/14/92 [Rev. 0l for clarification.

IR-025, R- 1 PY-CEI/NRR-1334 L. TAC HM75334, Relief request narrative was updated and revised 03/19/91 l Rev. 0l 02/14/92 l Res . 0] for clarification.

IR-026, R-1 PY-CEl/NRR-1334 L, TAC #M75334 Relief request narrative was updated and revised 03/19/91 l Rev. 0] 02/14/92 l Rev. 0 } for clarification.

IR-027, R-1 PY-CEl/N RR-1521 L, TAC nM84418, Relief request narrative was updated and revised 07/13/92 (Rev. 01 02/24/94 lRev. 0] for clarification.

Attachment i l PY-CEl/NRR-2310 Page 3 of 4 RcHef Request Previous Submittal NRC Safety Change Summary for Submittal of Relief

& New Rev Letters l$ valuations Request for the Second Interval j N o. I 1R-028, N/A PY-CEl/N RR-2068L, TAC #M95898, IR-028 is not being re-submitted for the second

, 06/28/96 [Rev. 0] 01/15t97 [Rev. 0] interval. In accordance with the 1989 Edition of j PY-CEl/NRR 2136L, TAC #M95898, ASME XI, this relief request is no longer '

02/12/97 [Rev. 01 05/27/97 lRev. 0l necessary.

IR-029, R-l PY-CEI!NRR 2068L, TAC #M95898, Relief request narrative was updated and revised 06/28/96 [Rev. 0] 01/15/97 [Rev. 0] for clarification. It was also revised to incorporate l PY-CEl/NRR-2136L, TAC #M95898, relief for Main Steam System welds that have I 02/12/97 [Rev. 01 05/27/97 [Rev. 0] similar jet shield obstructions. j

'IR-030, N/A PY-CEl/NRR-2210L, TAC #M99474, This relief request was not interval specific, rather 08/28/97 { Rey,0l 09/18/97 [Rev. 0] it was for deferral of the RPV circumferential shell PY-CEl/NRR-221 l'L, weld examinations until RFO8. Therefore, IR-030 09/04/97 [Rev. 0] is not being re-submitted for the second interval.

l PY-CEl/NRR-2219L, i 09/I6/97 [Rev. 0l IR-031 N/A PY-CEl/NRR-2244 L, 1 AC #MA0330 IR-031 is not being re-submitted for the second >

l2/18/97 [Rev. 0] 4/20/98 [Rev. 0] interval. Relief request was for the use of ASME XI Code Case N-524. N-524 is listed as an accepted Code Case in DG-1050, Revision 12 to l Regulatory Guide 1.147. Revision 12 of RG 1.147 is expected to be published prior to the start of j l- Perry's second 10-year interval.

IR-032 PY-CEl/N RR-2291 L, No SE yet These are Perry's IWE/lWL Relief Requests that

Through IR. 08/20/98 w ere recently submitted.

041 R-0 i IR-042 R-0 N/A N/A New relief request.  ;

1R-043, R-0 N/A N/A New reliet request. i PT-001, R-l PY-CEl/N RK-1078L, TAC #75334, Relief request narrative was updated and revised 11/17/89 and 06/13/90 [Rev. 0] for clarification. Also made some minor editorial PY-CEI/N RR-1208L, TAC #M75334, revisions, one deletion, and 6 additions to the 08/10/90 [Rev. 0] 02/14/92 [Rev. 0] component list. 1 PT-002 N/A PY-CEl/N RR-1078L, TAC #75334, PT-002 is not being re-submitted for the second  !

I I/17/89 and 06/13/90 [Rev. 0] interval. With the use of Code Cases N-498 l and PY-CEI/NRR 1208L, TAC #M75334, N-416-1 this reliefis no longer necessary.

08/10/90 [Rev. 01 02/14/92 [Rev. 0]

I PT-003, N/A Previously Withdrawn N/A N/A P T-004, N/A PY-CEl/NRR-2125 L, TAC #M97694, PT-004 is not being re-submitted for the second 01/07/97 lRev. 0] 09/l 1/97 [Rev. 0) interval. Relief request was for the use of ASME l PY-CEl/NRR-2136L, XI Code Case N-498-l. N-498-1 is listed as an 02/12/97 [Rev. 0] accepted Code Case in DG-1050, Revision 12 to

l. Regulatory Guide 1.147. Revision 12 of RG l.147 l is expected to be published prior to the start of Perry's second 10-year interval.

PT-005, N/A PY-CEl/NRR-2125L, TAC #M97694, PT-005 is not being re-submitted for the second ,

01/07/97 [Rev. 0] 05/15/97 [Rev. 0] interval. Relief request was for the use of ASME PY-CEl/NRR-2177L, TAC #M97694, XI Code Case N-522. N-522 is listed as an 07/02/97 [Rev. 0] 09/l I/97 [Rev. 0] accepted Code Case in DG-1050, Revision 12 to Regulatory Guide 1.147. Revision 12 of RG 1.147 is expected to be published prior to the start of Perry's second 10-year interval.

i u

Attachment 1 PY-CEl/NRR-2310 Page 4 of 4 Relief Request Previous Submittal NRC Safety Change Summary for Submittal of Relief

& New Rev Letters Evaluations Request for the Second Interval N o.

PT-006, R-1 PY-CEl/NRR-2125L, TAC 4M97694, The referenced requirements were updated to 01/07/97 [Rev. OJ 05/15/97 [Rev 0] reflect those of the 1989 Edition of Section XI and {

PY-CEl/NRR-2136L, TAC #M97694, the relief was revised to reflect its request for the 02/12/97 [Rev. 0) 09/l 1/97 [Rev. OJ second 10-year interval. -

PY-CEl/NRR-2177L, '

07/02/97 J Rev. 01 PT-007, R-1 PY-CEl/NRR-2176L, TAC #M99021. The referenced requirements were updated to 06/12/97 [Rev. 0] 08/12/97 [Rev. 0] retlect those of the 1989 Edition of Section XI and the relief was revised to tellect its request for the second 10-year interval.

PT-008, R-0 N/A N/A New relief request.

Use of CC PY-CEl/NRR-1851 L, TAC #M91005, Request for use of N-416-1 is not being re-N-416-1 (not 01/31/94 02/10/95 [Rev. N/A] submitted for the second interval. Code Case N-submitted in 416-1 is listed as an accepted Code Case in RR format, DG-1050, Revision 12 to Regulatory Guide 1.147.

just a letter Revision 12 of RG l.147 is expected to be ,

request) published prior to the start of Per y's second I 10-) ear interval, i

1

)

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Attachment 2 PY-CEI/NRR-2310L Page l of 85 Sheet'l of 5' Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-001, Rev. 2 i- '

I. Identification of Components System: Reactor Pressure Vessel, Class I A. Category: B-A, Pressure Retaining Welds Component' Description (See attached table for component ide: tification numbers)

1. . Head welds (Item No. Bl.22) l

2. Head to flange (Item-No. Bl.40)

B. Category: B-D, Full Penetration Welds of Nozzles in Vessels Component

Description:

-(See attached table for componeat identification numbers)

1. ~ Nozzle to shell welds (Item No. B3.90)

2. Nozzle inside radius sections (Item No. B3.100)

II '. 'ASME B&PV Section XI RequirementsSection XI, Division 1, Table IWB-2500-1, Categories B-A, B-D and B-F, require volumetric examination of welds and/or required volumes as shown in the applicable examination requirements figures.

III.: Relief Requested j Pursuant to 10 CFR 50.55a(a) (3) (ii), relief is requested from examining ,

essentially'100%.of the weld and/or required volume for those components listed in the attached table (See the table for percent' complete for each specific component) for Perry's second 10-year inspection interval.

IV. Basis for Relief I The structural integrity of the reactor pressure vessel welds was demonstrated during construction by meeting the requirements of the ASME Code Section III, and additionally by meeting the requirements-of ASME Section XI during preservice inspections. All welds were examined in 4 accordance with.the' appropriate Code requirements, weld' techniques and -]

welders were' qualified in accordance with Code requirements, and ;aaterials  !

were purchased and; traced in accordance with.the' appropriate Code and NRC requirements and guidelines. The Perry Unit-1 reactor vessel had no i reportable indications from preservice or first interval inspection results.  !

I

'The pressure boundary passed the required preservice hydrostatic tests and

'first interval' pressure tests, and has operated for a total of about 2,662

. equivalent full power days between November 1987 and March 1998, without detectable pressure boundary leakage.

I 9

Attechment 2 PY-CEI/NRR-2310L l Page 2 of 85 l Sheet 2 of 5 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-001, Rev. 2 Catastrophic reactor vessel failure is precluded by avoiding nil ductile temperatures at significant stress levels according to the design, surveillance and operating provisions described in the Perry USAR Sections

'5.3.1 and 5.3.2, and Technical Specifications 3.4.11. Additionally, none of the comoonents for which relief is being requested are located in the l beltline region of the reactor vessel. I Partial examinations, meeting all the requirements of the ASME Section XI except complete coverage, will continue to be performed. Although the examination coverage is limited, the most critical areas of the weld and/or required volume (i.e., the root of the weld and adjacent base material for welds and the radius blend area of inner radius exams) receive full coverage. Since the construction, operating conditions, environmental conditions of the non-examined portions are identical to the examined portions, it is reasonable to apply the satisfactory results from

.the examined to the non-examined portions.

The specific cause of the examination limitations for each component is shown in the attached table. In general, the limitations are a result of manual and/or automated scanning restrictions due to the component geometry itself or adjacent interferences. All of the subject components are within Perry's biological shield where access is difficult and the dose rate in the areas of the components typically exceeds 1 REM. In those cases where the component geometry precludes complete examination, some additional coverage could be achieved with the use of additional examination techniques, but only through undue hardship and a significant increase in dose.

Revision 1 of this relief request, which requested similar relief for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. M84418, dated 2/24/94).

In summary, because of the initial vessel condition free of reportable indications, successful code hydrotest and operating experience without leakage indications, the capability to detect pressure boundary leakage, protection against brittle reactor vessel failure, the capability to perform partial examinations which include full examination of the most critical areas, and the previous approval of similar relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

V. Alternate Examination Partial examinations, which meet all the requirements of ASME Section XI, 1989 Edition, except coverage, will be performed.

l Attachment 2 )

PY-CEl/NRR-2310L l Page 3 of 85 Sheet 3 of 5 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-001, Rev. 2 SYSTEM: Reactor Pressure Vessel

% COMPLETE ITEM NO. WELD I.D. DESCRIPTION CODE CATEGORY 1 //

B3.90 1-B13-N1A-KA N1 NOZZLE TO SHELL WELD BD 83 63

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N1B-KA N1 NOZZLE TO SHELL WELD BD 83 63

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N2A-KA N2 NOZZLE TO SHELL WELD BD 82 67

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N2B-KA N2 NOZZLE TO SHELL WELD BD 82 67

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N2C-KA N2 NOZZLE TO SHELL WELD BD 81 67

• SCAN PATH LIMITED BY NOZZLE GEOMETRY AND OBSTRUCTED BY ADJACENT N9 JET PUMP INSTRUMENTATION NOZZLE.

B3.90 1-B13-N2D-KA N2 NOZZLE TO SHELL WELD BD 81 67

• SCAN PATH LIMITED BY NOZZLE GEOMETRY AND OBSTRUCTED BY ADJACENT N9 JET PUMP INSTRUMENTATION NOZZLE.

B3.90 1-B13-N2E-KA N2 NOZZLE TO SHELL WELD BD 82 67

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N2F-KA N2 NOZZLE TO SHELL WELD BD 82 67

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N2G-KA N2 NOZZLE TO SHELL WELD BD 82 67

• SCAN I PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N2H-KA N2 NOZZLE TO SHELI. WELD BD 81 67

• SCAN PATH LIMITED BY NOZZLE GEOMETRY AND OBSTRUCTED BY ADJACENT N9 JET PUMP INSTRUMENTATION NOZZLE.

B3.90 1-B13-N2J-KA N2 NOZZLE TO SHELL WELD BD 81 67

• SCAN PATH LIMITED BY NOZZLE G30 METRY AND OBSTRUCTED BY ADJACENT N9 JET PUMP INSTRUMENTATION NOZZLE.

B3.90 1-B13-N2K-KA N1 NOZZLE TO SHELL WELD BD 82 67

• SCAN PATH OBSTRUCTED BY NOZZLZ GEOMETRY.

]

B3.90 1-B13-K4A-KA N4 NOZZLE TO SHELL WELD BD 83 69

• SCAN {

PATH OBSTRUCTED BY NOZZLE GEOMETRY. )

B3.90 1-B13-N4B-KA N4 NOZZLE TO SHELL WELD BD 83 69

• SCAN l PATH OBSTRUCTED BY NOZZLE GEOMETRY. ,

l 1 PERPENDICULAR SCAN

// PARALLEL SCAN PERPENDICULAR WELD EXAMINATION LIMITED TO ONE DIRECTION TOWARD NOZZLE CENTERLINE.

L Attachment 2 PY-CE!/NRR-2310L Page 4 of 85 Sheet 4 of 5 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-001, Rev. 2 SYSTEM: Reactor Pressure Vessel

% COMPLETE ITEM NO. WELD I.D. DESCRIPTION CATEGORY CODE 1 //

B3.90 1-B13-N4C-KA N4 NOZZLE TO SHELL WELD BD 83 69

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3,90 1-B13-N4D-KA N4 NOZZLE TO SHELL WELD BD 83 69

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N4E-KA N4 NOZZLE TO SHELL WELD BD 83 69

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N4F-KA N4 NOZZLE TO SHELL WELD BD 83 69

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-NSA-KA N5 NOZZLE TO SHELL WELD BD 83 68

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N5B-KA N5 NOZZLE TO SHELL WELD BD 83 68

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N6A-KA N6 NOZZLE TO SHELL WELD BD 84 70

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N6B-KA N6 NOZZLZ TO SHELL WELD BD 84 70

• SCAN PATH LIMITED BY NOZZLE GEOMETRY.

B3.90 1-B13-N6C-KA N6 NOZZLE TO SHELL WELD BD 84 70

• SCAN PATH LIMIIED BY NOZZLE GEOMETRY.

Bl.40 1-B13-AG TOP HEAD TO TOP HEAD BA 50 50 SCAN PATH. LIMITED TO TOP HEAD SIDE ONLY.

FLANGE WELD B3.90 1-B13-N9A-KA N9 NOZZLE TO SHELL WELD BD 88 62

• SCAN PATH OBSTRUCTED BY ADJACENT N2 RECIRCULATION INLET NOZZLES AT 90' AND 120' AZ.

B3.90 1-B13-N9B-KA N9 NOZZLE TO SHELL WELD BD 88 62

• SCAN PATH OBSTRUCTED BY ADJACENT N2 RECIRCULATION INLET NOZZLES AT 270' AND 300* AZ.

B3.90 1-B13-N15-KA N15 NOZZLE TO BOTTOM HEAD BD 0 0 OBSTRUCTION PRESENTED BY CRD TUBE BUNDLE B3.100 1-B13-N15-IR N15 NOZZLE INNER RADIUS BD 0 0 OBSTRUCTION PRESENTED BY CRD TUBE BUNDLE 1 PERPENDICULAR SCAN

// PARALLEL SCAN PERPENDICULAR WELD EXAMINATION LIMITED TO ONE DIRECTION TOWARD NOZZLE CENTERLINE.

l

Attachment 2 PY-CEUNRR-2310L Page 5 of 85 Sheet 5 of 5 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-001, Rev. 2 SYSTEM: Reactor Pressure Vessel

% COMPLETE ITEM NO. WELD I.D. DESCRIPTION CODE CATEGORY 1 //

Bl.22 1-B13-DG BOTTOM HEAD CENTER PLATE BA 29 29 OBSTRUCTION PRESENTED BY CRD TUBE BUNDLE AND TO SIDE PLATES, 270* SIDE SKIRT KNUCKLE.

Bl.22 1-B13-DH BOTTOM HEAD CENTER PLATE BA 29 29 OBSTRUCTION PRESENTED BY CRD TUBE BUNDLE AND TO SIDE PLATES, 90' SIDE SKIRT KNUCKLE.

1 PERPENDICULAR SCAN

// PARALLEL SCAN PERPENDICULAR WELD EXAMINATION LIMITED TO ONE DIRECTION TOWARD NOZZLE CENTERLINE.

THE ABOVE LISTED ITEMS CAN BE FOUND ON ISI ISO'S SS-305-006-102 THROUGH 111.

Sheet 1 of 4

Attachm:nt 2 PY-CEl/NRR-2310L Pagc 6 0t 85 Sheet 1 of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-002, Rev. 1 I. Identification of Components Class 1, Category B-G-1, Item No. B6.180, pump bolts and studs, in place (See attached table for ID numbers).

II. ASME B&PV Section XI RequirementsSection XI, Table IWB-2500-1, Category B-G-1 requires a volumetric examination of the volume defined by Figure IWB-2500-12.

III. Relief Requested Pursuant to 10 CFR 50.55a(a) (3) (ii), relief is requested from examining essentially 100% of the required volume for the components listed in the attached table (see the table for percent complete) for Perry's second 10-year interval.

IV. Basis for Relief The end configuration of the reactor recirculation pump stads, with a chamfered lip, plug hole and elongation measurement bore hole, prohibits all but a very limited 0-degree examination from the end of the stud.

Therefore, the volumetric examination of the reactor recirculation pump studs is performed by the use of forward and aft looking angle beam bore probes. The elongation measurement borehole (approximately 0.5-in. dia.)

extends through 80% of bolt length. Examination coverage from the borehole is slightly limited by the stud geometry as depicted in Figure IR-002-1. The volume affected is approximately 22% of the total required volume.

The structural integrity of the recirculation pump bolting was demonstrated during construction by meeting the requirements of the ASME Code Section III, and additionally by meeting the requirements of ASME Section XI during preservice inspections. Materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines. There were no reportable indications observed from preservice or first interval inspections. The pressure boundary passed the required preservice hydrostatic test and inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leak indication attributable to the subject components.

The major area of interest, the thread root area, received 100% volumetric examination. Material in the examined volume is identical to the non-examined portion of the studs. Since the construction, operating conditions and environmental conditions of the non-examined portions are identical to the examined volume, it is reasonable to apply satisfactory results obtained from the inservice inspections to the non-examined volume.

l l i

i

Attachment 2 PY-CEI/NRR-2310L Page 7 of 85 Sheet 2 of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-002, Rev. 1 An increase in the coverage, but still not complete coverage, could be obtained by grinding or machining off the chamfered lips at the ends of the studs to facilitate 0-degree examination from the end surface.

However, this would cause undue hardship because it would require either disassembly of the pump or an estimated 1 REM of dose to grind them in place.

Revision 0 of this relief request, which requested the same relief for the subject studs for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. 61443, dated 4/25/90)

In summary, because of acceptable initial bolt condition, successful code hydrotest and operating experience without related leakage indications, the capability to examine about 78% of bolt volume on a continuing basis, and the previous approval of the same relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination Partial examination, which meets all the requirements of ASME Section XI, 1989 Edition except coverage, will be performed.

J j

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l l

t

Attachment 2 PY-CEUNRR-2310L Page 8 of 85 Sheet 3 of 4 l Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-002, Rev. 1 g Chamfered lip at stud end JA NK Unexamined \

Volume J-K-N-O e---- Threads O N Bore hole O R lN Unexamined Volume Q-R-L-M M \ L i ll Figure IR-002-1 Code Required Volume J-K-L-M

i l

Attachment 2 PY-CEI/NRR-2310L Page 9 of 85 1 l

Sheet 4 of 4 4

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-002, Rev. 1 COMPONENT I.D. 1 DESCRIPTION i, 1-B33-C001B-1B i REACTOR RECIRC. PUMP B STUD 1 '

1-B33-C0018-2B REACTOR RECIRC. PUMP B STUD 2 1-B33-C001B-3B- REACTOR RECIRC. PUMP B STUD 3 1-B33-C001B-4B REACTOR RECIRC. PUMP B STUD 4 1-B33-C001B-5B REACTOR RECIRC. PUMP B STUD 5 1-B33-C001B-6B REACTOR RECIRC. PUMP B STUD 6 1-B33-C001B-7B REACTOR RECIRC. PUMP B STUD 7 1-B33-C001B-8B REACTOR hECIRC. PUMP B STUD 8 1-B33-C001B-9B REACTOR RECIRC. PUMP B STUD 9 1-B33-C0018-10B REACTOR RECIRC. PUMP B STUD 10 1-B33-C001B-11B REACTOR RECIRC. PUMP B STUD 11 1-B33-C0018-12B REACTOR RECIRC. PUMP B STUD 12 1-B33-C001B-13B REACTOR RECIRC. PUMP B STUD 13 1-E33-C0018-14B REACTOR RECIRC. PUMP B STUD 14 1-B33-C001B-15B REACTOR RECIRC. PUMP B STUD 15 1-B33-C001B-16B REACTOR RECIRC. PUMP B STUD 16 The above listed items can be found on ISI ISO SS-305-602-105 1

1 i

l f

i I

F .

Attachmmt 2 PY-CEl/NRR-2310L Page 10 0f 85

~ Sheet'l of 3 Perry Nuclear Power Plant Unit 1.

RELIEF REQUEST #IR-004, Rev. 2.

l I. Identification of Components l

Class 1, Category.B-J (Item numbers in attached table), piping welds 4 finches.NPS and greater.

II. ASME B&PV Section XI Requirements I.

ASME XI, 1989 Edition, Table IWB-2500-1, Category B-J requires surface and volumetric examination of_the areas and volumes defined by Figure IWB-2500-8.

III. Relief' Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining essentially 100% of the weld and required volume for the component listed in the attached table (see the table for percent complete) for Perry's second 10-year interval.

IV. Basis for Relief The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the.ASME Code Section III,'and additionally by meeting the requirements of ASME Section XI during preservice inspections. The subject weld was examined l in accordance with the appropriate. Code requirements, weld techniques and welders were qualified in accordance with Code requirements,'and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines. There were no reportable indications during preservice or first interval inspection.

.The pressure boundary passed the required preservice hydrostatic test and first interval pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject weld or similar welds.

In addition to partial inspection of the subject weld, complete examinations meeting the requirements of the ASME Code Gection XI are performed on welds of similar-configurations that utilize the same weld techniques, procedures and materials. The examined welds are subject to the same operating and environmental conditions as the partially examined welds.

Since the construction, operating conditions and environmental conditions of the non-examined portion of the welds are identical to the examined portions, it is reasonable to apply satisfactory results from examined to Ethe non-examined portions.

u____--_-__----

Attachm:nt 2 PY-CEl/NRR-2310L ,

Page 11 of 85 i Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-004, Rev. 2 ,

l Complete surface examinations are performed. l Design, procurement and operational provisions against nil ductile failure of the subject welds remain as described in the Perry USAR.

The limitation described for the subject weld is such that only redesign and replacement of the piping at the weld location would provide for '

I complete examination.

Revision 1 of this relief request, which requested the same relief for the subject wela for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. M75334, dated 2/14/92).

In summary, becauce of acceptable initial condition, successful code hydrotest and operating experience without related leakage indications, protection against bitttle failure, the capability to ultrasonically examine most of the suMect weld volume and perform complete surface examinations on a contineing basis, and the previous approval of the same relief it is concluded that the specified requirements would result in hardship or unusual diffictity without a compensating increase in the level of quality or safety.

V. Alternate Examination Partial examination, which meets 2_' the requirements of ASME Section XI, 1989 Edition except coverage will be pctformed.

I i

1 I

Attachm:nt 2 PY-CEl/NRR-2310L Page 12 of 85 Sheet 3 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-004, Rev. 2 NATURE OF ITEM NO. COMPCNENT I.D. SYS./DWG. NO. DESCRIPTION OBSTRUCTION EST. % COMPLETE B9.11 1E12-0880 RHR/642-143 12" PENETRATION INTRADOSE OF ELBOW DOES 80%,

PROCESS PIPE TO NOT ALLOW FOR THE FULL SHORT RADIUS ELBOW V SCAN PATH NECESSARY FOR THIS ONE-SIDED EXAM RHR = RESIDUAL HEAT REMOVAL SYSTEM i

i

Attachment 2 PY-CEl/NRR-2310L l Page l3 0f 85 I Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-005, Rev. 2 I. Identification of Components 1

Class 1, Category B-J, Item B9.11, piping welds 4 inches NPS and greater (see attached table for I.D. numbers) .

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-J requires surface and volumetric examinaslon of the areas and volumes defined by Figure IWB-2500-8.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from ultrasonically examining essentil'ly 100% of the required volume for the components listed in the attached table (see the table for percent complete) for Perry's second 10-year interval.

IV. Basis for Relief Ultrasonic examinations conducted on welds in the recirculation loops l which were inlaid and overlaid with corrosion resistant cladding require specialized techniques. Typical techniques identified in Appendix III of Section XI proved to be ineffective.

To overcome the metallurgical properties impeding conventional shear wave ultrasonic transmission, refracted longitudinal wave examinations are employed. Personnel qualified for IGSCC detection in accordance with NUREG-0313, Rev. 2, perform the examinations. The acoustic properties of refracted longitudinal wave propagation limit the technique to 1/2 vee path. When examining from one side only, the Code required volume necessitates a full vee path through the weld and required volume.

Therefore, when access to a butt weld was limited to one side only due to component geometry (e.g., pipe to valve) the perpendicular examination is considered to be only 50% complete.

During construction, the subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines. In addition, there were no reportable indications during preservice or first interval inspections. l The pressure boundary passed the required preservice hydrostatic and first interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 l without leakage indication attributable to the subject welds.

Attachment 2 PY-CEl/NRR-2310L Page 14 of 85 Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-005, Rev. 2 1

l Since the construction, operating conditions and environmental conditions of the non-examined portion of the welds are identical to the examined i portions, it is reasonable to apply satisfactory results from examined to i the non-examined portions.

Complete surface examinations are performed. l Design, procurement and operational provisions against nil ductile failure of the subject welds remain as described in the Perry USAR.

The limitation described for the subject welds is such that only a complete redesign and replacement of the piping at the weld locations would provide for complete examination.

Revision 1 of this relief request, which requested the same relief for the '

subject welds for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. M95898, dated 5/27/97)

In summary, because of acceptable initial condition, successful code hydrotest and operating experience without related leakage indications, protection against brittle failure, the capability to ultrasonically examine half of the subject weld volume and perform complete surface examinations on a continuing basis, and the previous approval of the same relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination Ultrasonic examinations, which meet all the requirements of ASME Section XI, 1989 Edition and NUREG-0313, Rev. 2 to the extent practical will be performed.

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1 1

Attachment 2 PY-CEI/NRR-2310L Page 16 of 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-007, Rev. 1 1

i I. Identification of Components Class 1, Category B-K-1, Item No. B10.10 integrally welded support attachments for piping (See attached table for ID numbers).

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-K-1 requires surface examination of the areas defined by Figures IWB-2500-12 through 13.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining 100% of the required surface areas for penetration to process pipe attachment welds. Due to the inaccessibility of the weld face within the ID of the penetration, only 50% of the required surface area can be examined. The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines.

Examinations meeting the requirements of the ASME Code Section XI were performed on the accessible face of the attachment welds with acceptable results during preservice and first interval inspections.

In accordance with Perry's USAR, penetration attachment welds within the l high-energy break exclusion regions of piping systems are ultrasonically examined from the OD surface of the penetration. Although not performed specifically to supplement the limited surface examinations, these examinations do provide additional assurance of structural integrity.

The pressure boundary passed the required preservice hydrostatic and inservice pressure tects, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject weld or similar welds.

Since the construction, operating conditions and environmental conditions of the non-examined portion of the welds are identical to the examined '

portions, it is reasonable to apply satisfactory results from examined to the non-examined portions.

Atuchment 2 )

PY-CEl/NRR-2310L '

Page 17 of 85 Sheet 2 of 3 I

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-007, Rev. 1 Design, procurement and operational provisions against nil ductile failure >

.of the subject welds remain as described in the Perry USAR.

Redesign of the piping systems to facilitate access is not practical and '

performance of supplemental ultrasonic examinations, beyond those already required for high energy break exclusion regions, would present undue 3 hardship. Calibration blocks would have to be designed and fabricated and the supplemental examinations would require as much as 2 REM to perform.

The NRC-(reference TAC No. 61443, dated 4/25/90) approved revision 0 of this relief request, which requested similar relief for the subject welds for Perry's first 10-year inspection interval.

In summary,.because of the acceptable initial condition, successful code hydrotest and operating experience, the capability to examine half of the required surface areas on a continuing basis, protection against brittle failure, and the previous approval of similar relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination Partial examinations, which meet all the requirements of ASME Section XI, 1989 Edition except coverage, will be performed.

I 4

1 i

l

Attachnunt 2 PY-CEI/NRR-2310L Page 18 of 85 Sheet 3 of 3 i

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-007, Rev, 1 l

COMPONENT I.D. SYSTEM /ISI ISO SS-305-1-E12-P411-WA RHR/642-135 l

1-E12-P421-WA RHR/642-101 1-E12-PRB2035-WA RHR/642-143 1-E12-PRB2036-WA RHR/642-139 1-E12-PRB2044-WA RHR/642-126 1-E21-P112-WA LPCS/705-109 1-E21-PRB3046-WA LPCS/705-110 1-E22-P410-WA HPCS/701-109 1-E22-PRB3052-WA HPCS/701-110 1-E51-P123-WA RCIC/631-106 1-C41-PRB4031-WA SLC/691-102 1-E51-P422-WA RCIC/632-102 l

• • 1-N27-P121-WA FW/082-101

• • 1-N27-P414-WA FW/082-101 1-G33-P131-WA RWCU/671-104 l 1-N22-P423-WA MS/121-103 l

• • 1-B21-P122-WA MS/605-109

• • 1-B21-P124-WA MS/605-107

• • 1-B21-P415-WA MS/605-110

• • 1-B21-P416-WA MS/605-103 RHR = Residual Heat Removal FW = Feedwater LPCS = Low Pressure Core Spray RWCU = Reactor Core Isolation Cooling HPCS = High Pressure Core Spray MS = Main Steam RCIC = Reactor Core Isolation Cooling SLC = Standby Liquid Control

• • Receive augmented ultrasonic examination in accordance with USAR requirements for high-energy break exclusion regions.

Attachment 2 PY-CEI/NRR-2310L Page 19 of 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-009, Rev. 1 I. Identification of Components Class 1, Category B-0, Item B14.10, flange welds in control rod drive housing (See attached table for I.D. numbers). l l

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-O requires surface or volumetric examination of the areas or volumes defined by Figure IWB-2500-18.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining 100% of the required surface areas because of partial inaccessibility due to control line interferences. The insert and withdraw control lines limit the examination coverage to 85% (see Figure IR-009-1). The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The structural integrity of the subject welds was demonstrated during construction by meeting the requirements of the ASME Code Section III, and additionally by meeting the requirements of ASME Section XI during preservice inspections. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were j purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines. The subject welds had no reportable indications during preservice or first interval inspections. l The pressure boundary passed the required preservice hydrostatic and first interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject welds.

Portions of welds examined are subject to the same operating and environmental conditions as the unexamined portions. It is, therefore, l reasonable to apply the results from examined weld portions to the l unexamined portions. j l

Disassembly of the CRD mechanisms does not facilitate additional coverage for the required surface examination areas and redesign of the CRD mechanisms to provide access is not practical.

AttIchmen2 2 PY-CEI/NRR-2310L Page 20 0f 85 Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-009, Rev. 1 Design, procurement and operational provisions against nil ductile failure of the subject welds remain as described in the Perry USAR.

The NRC (reference TAC No. 61443, dated 4/25/90) approved revision 0 of this relief request, which requested the same relief for the subject wolds for Perry's first 10-year inspection interval.

In summary, because of acceptable initial weld condition, successful code hydrotest and operating experience without leakage indications, the capability to examine most of the weld surface on a continuing basis, protection against brittle failure, and the previous approval of the same relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination Partial examinations, which meet all the requirements of ASME Section XI, 1989 Edition except coverage, will be performed.

O Withdraw Control Line

CRD Housing Insert __, '

Control Line CRD Housing to Flange Weld

,- (Appx. 15% of circumference is

> < obstructed by the two control f lines)

I i

i CRD Flange U U i i Figure IR-009-1

(

I L

Attachment 2 PY-CEl/NRR 2310L Page 21 of 85 Sheet 3 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-009, Rev. 1 WELD I.D. WELD I.D.

1-B13-02/23-FW 1-B13-30/59-FW 1-B13-02/27-FW 1-B13-34/03-FW 1-B13-02/31*FW 1-B13 '34/59-FW 1-B13-02/35-FW 1-B13-38/03-rW 1-B13-02/39-FW 1-B13-38/59-FW 1-B13-06/15-FW 1-B13-46/07-FW 1-B13-06/47-FW 1-B13-46/55-FW 1-B13-10/11-FW 1-B13-50/11-FW 1-B13-10/51-FW 1-B13-50/51-FW 1-B13-14/07-FW 1-B13-54/15-FW 1-B13-14/55-FW 1-B13-54/47-FW 1-B13-22/03-FW 1-B13-58/23-FW 1-B13-22/59-FW 1-B13-58/27-FW 1-B13-26/03-FW 1-B13-58/31-FW 1-B13-26/59-FW 1-B13-58/35-FW 1-B13-30/03-FW 1-B13-58/39-FW The above listed items can be found on ISI ISO drawing SS-305-006-110. l u_____________ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _

Attachment 2 PY-CEl/NRR-2310L Page 22 of 85 Sheet 1 of 3 Parry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-012, Rev. 2 I. Identification of Components Class 2, Category C-C, (Item and component numbers in attached table) integrally welded support attachments.

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWC-2500-1, Category C-C requires surface examination of the areas defined by Figure IWC-2500-5.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining 100% of the required surface areas for those pump support integral i attachments welds that are only partially accessible due to pedestal interference (see Figure IR-012-1). Coverage is limited to approximately 83%. The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The structural integrity of the subject pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and J guidelines. There were no reportable indications during ASME Section XI l preservice or first interval inservice inspections. ]

I The pressure boundary passed the required hydrostatic test and first interval inservice system pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998, without leakage indication attributable to the subject welds.

In addition to partial inspection of the subject welds, complete examinations meeting the requirements of the ASME Code Section XI are performed on welds of similar configurations that utilize essentially similar weld techniques, procedures and materials. The examined welds are subject to the same operating and environmental conditions as the partially examined welds.

Since the construction, operating conditions and environmental conditions of the non-examined portions of the walds are identical to the examined portions, it is reasonable to apply satisfactory results to the non-examined portions.

Attachment 2 PY-CEL/NRR 2310L Page 23 of 85

' Sheet 2'of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-012, Rev. 2 Design, procurement and operational provisions against nil ductile failure t,

of the subject welds remain as described in the Perry USAR.

Full access to the subject pump casing integral attachment welds could only be provided by redesign of the pump's pedestal mounting Revision 1 of this relief request, which requested similar relief for the subject welds for Perry's first 10-year inspection interval, was approved  ;

by the NRC (reference TAC No. M75334, dated 2/14/92). '

In summary, because of the acceptable initial condition, successful test and operating experience, the capability to examine at least 83% of the weld surfaces on a continuing basis, protection against brittle failure, and the previous approval of similar relief it is concluded that the-specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination  ;

i Partial examinations, which meet all the requirements of ASME Section XI, 1989 Edition except coverage, will be performed.

l 1

1 1

Integrally Attached "U" Shaped Lug Welded on All Sider Front View of Lug (Typical of 4) 7, p/ 4 Pump Casing p

Inaccessible +---- I Beam Support of Lug Area Behind I Beam 4, Pedestal Figure IR-012-1 1

1 3

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SO 3 3 3 3 EC 8 8 8 8 LS LS LS LS AS N

AS AS AS TE TE TE TE O SC SC SC SC FI EC EC EC EC OT DA DA DA DA C E E E E EU PS PS PS PS RR K K K K UT PC PC PC PC MO TS MO MO MO AB UL UL UL UL NO PB PB PB PB 1 G G G G N N N N t 2 I I I I i S S S S n .

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Attaclunent 2 PY-CEl/NRR-2310L Page 25 of 85 Sheet 1 of 4 l

I Perry Nuclear Power Plant Unit 1 1 RELIEF REQUEST #IR-013, Rev. 1 I. Identification of Components Class 2, Category C-G, Item C6.10, pump casing welds (See attached table for ID numbers).

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category C-G requires surface examination of the areas defined by Figure IWC-2500-8.

III. Relief Requested i

Pursuant to 10 CFR 50.55a(a) (3) (ii), relief is requested from performing routinely scheduled examinations of those pump casing welds that are inaccessible because they are located within those portions of vertical line shaft pump casings that are installed below the floor elevations (i.e., bolted into a pit). The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The structural integrity of the subject pressure boundaries was demonstrated during construction by meeting the requirements of the ASME Code Section III, and additionally by meeting the requirements of ASME Section XI during preservice inspections. The subject welds were examined in accordance with the appropriate Code requirements, weld techniquca and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines.

The pressure boundary passed the required preservice hydrostatic and first interval l inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998, without leakage indication attributable to the subject welds.

The subject welds are below the floor elevation. Only disassembly of the pump and removal of all the pump internals would provide examination access to the these welds. Disassembly would involve removal of the 7,800-lb. pump motor, and the 22-ft long, 16,000-lb. pump head and vertical line shaft assembly.

l l

C_______________._____..___ _ . _ _ _ - - - - - _ - - _ . _ _ _ _ - _ . - _ - - - - - - - - - - - - - - - - - - - - - -

Attachment 2 PY-CEl/NRR-2310L Page 26 of 85 Sheet 2'of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-013, Rev. l' The pump casing welds above the-floor elevation can be fully examined. Since the construction and operating conditions of these accessible pump casing welds are identical to those of the inaccessible welds, it is reasonable to apply satisfactory results from examined. welds to the non-examined welds.

The NRC (reference TAC No. 61443, dated 4/25/90) approved revision 0 of this relief request,' which requested similar relief .for the subject welds for Perry's ' first 10-year inspection interval.

In summary, be'cause of the acceptable initial condition, successful code hydrotest and operating experience, the capability to examine the similar accessible welds on a continuing basis, and the previous approval of similar relief it is concluded

'that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

L ..

V. Alternate Examination l If the subject welds.become accessible through the disassembly of the pumps for maintenance, repair ar modification, examinations which meet all the requirements of ASME Section XI, 1989 Edition will be performed.

l l

l l.

1 l

Attachment 2 l PY-CEl/NRR 2310L l Page 27 of 85 l

l Sheet 3 of 4 l

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-013, Rev. 1 WELD NO. ISI ISO SS-305- DESCRIPTION 1-E22-C001-001 701-114 Pump Head to Barrel Shell Weld 1-E22-C001-002 701-114 Pump Shell to Shell Weld 1-E22-C001-003. 701-114 Pump Shell to Shell Weld 1-E22-C001-004 701-114 Pump Shell to Flange Weld 1-E22-C001-013 701-114 Pump Barrel Longseam 1-E22-C001-014 701-114 Pump Barrel Longseam 1-E22-C001-015 701-114 Pump Barrel Longseam 1-E21-C001-001 705-113 Pump Head to Barrel Shell Weld 1-E21-C001-002 705-113 Pump Shell to Shell Weld 1-E21-C001-003 705-113 Pump Shell.to Flange Weld 1-E21-C001-012 705-113 Pump Shel to Shell Weld 1-E21-C001-013 705-113 Pump Barrel Longseam 1-E21-C001-014 705-113 Pump Barrel Longseam 1-E21-C001-015 705-113 Pump Barrel Longseam 1-E12-C002A-001 641-120 Pump Head to Barrel Shell Weld 1-E12-C002A-002 641-120 Pump Shell to Shell Weld 1-E12-C002A-003 641-120 Pump Shell to Flange Weld 1-E12-C002A-012 641-120 Pump Shell to Shell Weld 1-E12-C002A-013 641-120 Pump Barrel Longseam 1-E12-C002A-014 641-120 Pump Barrel Longseam I

1-E12-C002A-015 641-120 Pump Barrel Longseam 1 E12 = Residual Heat Removal E22 - High Pressure Core Spray E21 = Low Pressure Core Spray 4

_______.__.____________.__.-__________a

Attachment 2 PY-CEI/NRR-2310L Page 28 of 85 Sheet 4 of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-013, Rev. 1 WELD NO. ISI ISO SS-305- DESCRIPTION 1-E12-C002C-001 643-122 Pump Head to Barrel Shell Weld 1-E12-C002C-002 643-122 Pump Shell to Shell Weld 1-E12-C002C-003 643-122 Pump Shell to Flange Weld 1-E12-C002C-012 643-122 Pump Shell to Shell Weld 1-E12-C002C-013 643-122 Pump Barrel Longseam 1-E12-C002C-014 643-122 Pump Barrel Longseam 1-E12-C002C-015 643-122 Pump Barrel Longseam E12 = Residual Heat Removal t

j i

J

Attachment 2 PY-CEI/NRRa310L Page 29 of 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-015, Rev. 1

.I . Identification of Components Class 2, Category C-C, Item No. C3.20 integrally welded support attachments for piping (See attached table for ID numbers).

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category C-C requires surface examination of the areas defined by Figure IWB-2500-5.

-III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining 100% of the required surface areas for penetration to process pipe attachment welds. Due to the inaccessibility of the weld face within the ID of the penetration, only 50%

of the required surface area can be examined. The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines.

Examinations meeting the requirements of the ASME Code Section XI were performed on the accessible face of the attachment welds with acceptable results during preservice and first interval inspections. l The pressure boundary passed the required preservice hydrostatic and inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days _between November 1987 and March 1998 without leakage indication attributable to the subject weld or similar welds.

Since the construction, operating conditions and environmental conditions of the non-examined portion of the welds are identical to the examined portions, it is reasonable to apply satisfactory results from examined to the non-examined portions.

l l

l I

Attachment 2 PY-CEl/NRR-2310L Page 30 0f 85 Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-015, Rev. 1 Design, procurement and operational provisions against nil ductile failure of the subject welds remain as described in the Perry USAR.

Redesign of the piping systems to facilitate access is not practical and performance of supplemental ultrasonic examinations would present undue hardship.

Calibration blocks would have to be designed and fabricated and the supplemental examinations would require as much as 1 REM to perform.

The NRC (reference TAC No. 61443, dated 4/25/90) approved revision 0 of this relief request, which requested the same relief for the subject weld for Perry's first 10-year inspection interval.

In summary, because of the acceptable initial condition, successful code hydrotest and operating experience, the capability to examine half of the required surface areas on a continuing basis, protection against brittle failure, and the previous approval of similar relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination Partial examinations, which meet all the requirements of ASME Section XI, 1989 Edition, except coverage, will be performed.

i l

l Attachment 2 PY-CEl/NRR-2310L Page 31 of 85 Sheet 3 of 3 l

l l

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-015, Rev. 1 ISI ISO COMPONENT I.D. SYSTEM /SS-305-1G33-P132-WA RWCU/672-102 1E12-P105-WA RHR/642-121 l 1E12-P407-WA RHR/642-126 1

1E12-P113-WA LPCI/642-133 1E12-P412-WA LPCI/642-137 RWCU = Reactor Water Cleanup RHR = Residual Heat Removal LPCI = Low Pressure Coolant Injection l

1 1

1

- _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _-___.__._-_-____----_------_-_--------------------------------------------O

l l l l l Attachment 2 PY-CEl/NRR-2310L Page 32 of 85 Sheet 1 of 3 l I l

Perry Nuclear Power Plant Unit 1 l

RELIEF REQUEST #IR-018, Rev. 2 '

I. Identification of Components Class 1, Category B-K-1, Item No. B10.10 integrally welded support attachments for piping (See attached table for ID numbers).

II. .ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-K-1 requires surface examination of the areas defined by Figures IWB-2500-13 through 15.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining 100% of the required surface areas of support lug to process pipe attachment welds due to inaccessibility of the weld face at the pipe clamp or box guide to support lug interface. At least 65% of the required surface is accessible and was examined during the first inspection interval and will be examined during subsequent intervals. This relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Cede and NRC requirements and guidelines.

The pressure boundary passed the required preservice hydrostatic and first inspection interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject welds.

Complete examinations meeting the requirements of the ASME Code Section XI are performed on welds of similar configurations that utilized the same weld techniques, procedures and materials. The examined welds are subject to the same operating and environmental conditions as the partially examined welds.

l Since the construction, operating conditions and environmental conditions of the non-examined portion of the welds are identical to the examined portions, it is reasonable to apply satisfactory results from examined to i the non-examined portions. l 1

_ - _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _____ ._- _ _ ___ _ - _ __________ _ _ _ _ a

b Attachment 2 PY-CEI/NRR-2310L Page 33 of 85 Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-018, Rev. 2 Design, procurement and. operational provision against nil ductile failure

'of the subject-welds remain.as described in the Perry USAR.

Component' support removal or in some cases (i.e., the rigid guides)

= complete redesign of the component supports would be necessary to facilitate full access to the integral attachment welds. All the supports are located in radiologically controlled areas with dose' rates that range from 10 mr/hr to 300 mr/hr-in the case of the reactor recirculation piping supports.

The NRC (reference TAC No. M75334, dated 2/14/92) approved revision 1 of this relief request, which requested similar relief for the subject welds for Perry's first 10-year inspection interval.

In summary, because.of the' acceptable initial condition, successful code hydrotest and operating experience, the capability to examine at least 65%

of the required surface areas on a continuing basis, protection against

' brittle failure, and the previous approval of similar relief it.is concluded that the specified requirements would result.in hardship or unusual difficulty without a compensating increase in the level of quality or-safety.

V. Alternate Examination Partial examinations, which meet all the requirements of ASME Section XI, i 1989 Edition, except coverage, will be performed.

t i

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l l

Attachment 2 PY-CEl/NRR-2310L Page 35 of 85 l Sheet 1 of 3 l Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-019, Rev. 1 l

I. Identification of Components Class 2, Category C-C (Item and component numbers in attached table),

l integrally welded support attachments.

II. ASME ;&PV Section XI Requirements l

ASME XI, 1989 Edition, Table IWB-2500-1, Category C-C requires surface l examination of the areas defined by Figure IWB-2500-5.

l III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii) , relief is requested from examining 100% of the required surface areas of support lug or anchor to process l pipe attachment welds due to inaccessibility of the weld face at the pipe

, clamp or structural interferences. At least 80% of the required surface is accessible and was examined during the first inspection interval and will be examined during subsequent intervals. This relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief l The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III, and additionally by meeting the requirements of ASME l Section XI during preservice inspections. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC l

requirements and guidelines. There were no reportable indications during preservice inspections.

The pressure boundary passed the required preservice hydrostatic and first inspection interval inservice pressure tests, and has operated for a total  ;

of about 2,662 equivalent full power days between November 1987 and March I 1998 without leakage indication attributable to the subject welds.

1 l

i Complete examinations meeting the requirements of the ASME Code Section XI were performed on welds of similar configurations that utilized essentially similar weld techniques, procedures and materials. The examined welds.are subject to the same operating and environmental conditions as the partially examined welds.

l.

l Anachment 2 PY-CEUNRR-2310L l Page 36 of 85 Sheet 2 of 3 i

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-019, Rev. 1 Since the construction, operating conditions and environmental conditions of the non-examined portion of the welds are identical to the examined portions, it is reasonable to apply satisfactory results from examined to the non-examined portions. ,

1 Design, procurement and operational provision against nil ductile failure of the subject welds remain as described in the Perry USAR.

Component support removal or in some cases (i.e., the anchors) complete redesign of the component supports would be necessary to facilitate full access to the integral attachment welds. All the supports are located in radiologically controlled areas with dose rates that range from 10 mr/hr ,

to 50 mr/hr. I The NRC (reference TAC No. M75334, dated 2/14/92) approved revision 1 of this relief request, which requested similar relief for the subject welds  !

for Perry's first 10 year inspection interval. '

In summary, because of the acceptable initial condition, successful code hydrotest and operating experience, the capability to examine at least 80%

of the required surface areas on a continuing basis, protection against brittle failure, and the previous approval of similar relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

1 \

V. Alternate Examination l

i Partial examinations, which meet all the requirements of ASME Section XI, 1989 Edition, except coverage, will be performed. i I

i u

Attachment 2 PY-CEl/NRR-2310L Page 37 of 85 Sheet 3 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-019, Rev. 1 NATURE OF EST %

ITEM NO. COMPONENT I.D. 'SYS./ISI ISO DESCRIPTION OBSTRUCTION CCMPLETE C3.20 1C11-H0032-WA CONTROL ROD DRIVE WELDED LUGS ADJACENT 86%

SS-305-871-103 FOR PIPE SUPPORT STRUCTURE C3.20 1C11-H0048-WA CONTROL ROD DRIVE WELDED LUGS ADJACENT 86%

SS-305-871-101 FOR PIPE SUPPORT STRUCTURE C3.20 1C11-H0665-WA CONTROL ROD DRIVE WELDED LUGS PIPE CLAMP 87%

SS-305-871-104 FOR PIPE SUPPORT C3.20 1C11-H0675-WA CONTROL ROD DRIVE WELDED LUGS PIPE CLAMP 87%

SS-305-871-102 FOR PIPE SUPPORT C3.20 1E12-H0670-WA RESIDUAL HEAT WELDED LUGS PIPE CLAMP 87%

REMOVAL FOR PIPE SUPPORT SS-305-642-137 C3.20 1E22-H0027-WA HIGH PRESSURE PIPE ANCHOR PIPE CLAMP 81%

CORE SPRAY SS-305-701-102 i

l i I

l- Attachment 2 l PY-CEUNRR 2310L l_ Page 38 of 85 Sheet 1 of 6 l Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-021, Rev. 4

'I. Identification of Components e

L l Class 3,1 Category.D-B, Item D2.20, Integral. Attachment - Component Supports and

! Restraints (See attached table for component identification).

II. ASME B&PV eection XI Requirements l

ASME XI, 1989 Edition, Table IWD-2500-1, Category D-B requires visual examination of the integral attachments as defined by Figure IWD-2500-1.

III. Relief Requested

- Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining 100% of the integral attachment welds that are inaccessible or only partially accessible l due to their location. They ,are located within floor or wall penetrations filled H

-with fire retardant sealant, . within suppression pool penetrations that are sealed with rubber boots, or within the Emergency Service Water punphouse forebay (See attached table for component identifications and percent complete). The relief is requested -for Perry's second 10-year inspection -interval.

IV. -Basis for Relief i-i' l The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III.

. All welds

[ were inspected in accordance with the appropriate Code requirements. Weld l techniques and welders were qualified in accordance with Code requirements and L materials were purchased and traced in accordance with the appropriate Code and NRC l-requirements.and guidelines.

J The pressure boundary passed the required preservice hydrostatic and first inspection interval inservice pressure tests, and has operated for a total of about

- 2,662 equivalent full power days between November 1987 and March 1998 without l leakage indication attributable to the subject welds.

i l: Complete examinations meeting the requirements-of the ASME Code Section XI are. performed on those portions of .the integral attachments that are accessible and {'

on integral attachments with similar configurations to those that are completely

-inaccessible. 265 Category D-B integral attachments were examined over the course of the first inspection interval and no indications were found.

Since the construction and operating conditions of the inaccessible or partially J

' inaccessible welded attachments are similar to that of welded attachments that are examined, it is reasonable to extend the satisfactory results of the accessible integral attachments to the inaccessible ones.

i.

u

Attachment 2 PY-CEl/NRR-2310L Page 39 0f 85 Sheet 2 of 6.

Perry Nuclear Power Plant Unit 1 RELIEF. REQUEST #IR-021, Rev. 4 Removal of the firt _stardant sealant within the wall and floor penetrations or the sealed rubber boet from the suppression pool penetrations would be necessary to provide full access to the integral attachment welds.

In both cases, special procedures and cure times are necessary for reinstallation of the sealants. The suppression pool penetrations are located in the drywell with dose rates that range from 10 mr/hr to 50 mr/hr. The integral attachments within the Emergency Service Water pumphouse forebay (which cannot be drained) present a safety hazard to get to and'could only be accessed by constructing a floating scaffold platform.

The NRC (reference. TAC No. M95898, dated 5/27/97) approved revision 3 of this relief request, which' requested similar relief for the subject welds. for Perry's first 10-year inspection interval.

In smwaary, because of the acceptable initial condition, successful-code hydrotest and operating experience, the capability to examine the partially accessible.

integral attachments on a continuing basis, and the previous approval of similar relief it is concluded that the specified requirements would result in hardship or-unusual difficulty without a compensating increase in.the level of quality or safety.

.V. Alternate Examination Examinations that meet all the requirements of ASME Section XI, 1989 Edition will be performed on the accessible portions of the integral' attachments.

l

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l l

1 I

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Attachment 2 l PY-CEl/NRR-2310L Page 44 of 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-023, Rev. 1 I. Identification of Components All safety-related hydraulic and mechanical snubbers.

II. ASME B&PV Section XI Requirements

, ASME XI, 1989 Edition, Subarticle IWF-5000, Inservice Inspection l Requirements for Snubbers, states that preservice examinations and tests i of snubbers (IWF-5200), inservice examinations and tests of snubbers I

(IWF-5300), and examination and tests of snubber repairs and replacements

-(IWF-54 00 ) shall be in accordance with the first Addenda to ASME/ ANSI OM-1987, Part 4. Furthermore, 10 CFR 50.55a (b) (2) (viii) specifies that the ASME/ ANSI OM Part 4 Edition and Addenda to be used shall be the OMa-1988 l Addenda to the OM-1987 Edition.

l III. Relief Request Pursuant to 10 CFR 50.55a (a) (3) (1), relief is requested from performing preservice examinations and tests of snubbers, inservice examinations and tests of snubbers, and examinations and tests of snubber repairs and replacements in accordance with ASME/ ANSI OM Part 4 of the OMa-1988 Addenda to the OM-1987 Edition.

IV. Basis for Relief Perry performs preservice and inservice examinations and tests of l

snubbers, and examinations and tests of snubber repairs and replacements in accordance with the technical requirements within section 6.4.1 of the Operational Requirements Manual (ORM). The requirements were relocated from (and are unchanged) Technical Specification 4.7.4 upon Perry's incorporation of Improved Technical Specifications. The basic technical l requirements within ORM 6.4.1 for examination and testing of snubbers are essentially the sane as those within ASME/ ANSI OM Part 4 of the OMa-1988 Addenda to the OM-1987 Edition. Additionally, the ORM provides requirements for the inspection of snubbers following transient events and a snubber service life replacement program.

Revision 0 of this relief request, which requested similar relief for snubbers for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. M61303, dated 9/7/90).

V. Alternate Testing Preservice and inservice examinations and tests of snubbers, and examinations and tests of snubber repairs and replacements in accordance with the technical requirements within section 6 4.1 of the Operational Requirements Manaal (ORM). The functional testing requirements therein are as follows:

1

l Attcchment 2 PY-CEl/NRR-2310L Page 45 of 85 i

I Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-023, Rev. 1 At least once per refueling interval, a representative sample of snubbers shall be tested using one of the following 3 plans for each type of snubber. The sample plan shall be selected prior to the test period and l cannot be changed during the test period.~ The snubber functional test j period may start ninety days prior to a scheduled refueling outage and

l. shall be completed prior to the end of the scheduled refueling outage.

The Nuclear Regulatory Commission shall be notified in writing pursuant

!- to 10 CFR 50.4 of the sample plan selected prior to the test period or

!- the sample plan used in the prior test period shall be implemented.

1) At least 10% of the total of each type of snubber shall be

functionally tested either in-place or in a bench test. For each snubber of a type that does not meet the functional _ test acceptance criteria an additional 5%~of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of.that type have been functionally. tested; or l 2) 'A representative sample of each type of snubber shall be

, functionally tested in accordance with Figure 6.4.1-1. "C" is the L total number of snubbers of a type found not meeting the acceptance

! requirements. .The cumulative number of snubbers of a type tested is I denoted by "N". At the end of each day's testing, the_new values of "N" and "C" (previous day's total plus current day's increments) shall be plotted on Figure 6.4.1-1. If at any time the point plotted falls on or below the " Accept" line, testing of snubbers of that type may be terminated. When the point plotted lies in the

" Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the " Reject" L region, or all the snubbers of that type have been tested. Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later 7 . time, providing all snubbers tested with the failed equipment during i

the day of the equipment failure are retested; or

(' 3) An initial representative sample of 55 snubbers of each type shall

. be functionally tested. For each snubber type which.does not meet

the functional test acceptance criteria, another sample of at least one-half the size of the Anitial sample shall be tested until the total number tested is equal to the initial sample size multiplied by the_ factor, 1 + C/2, where "C" is the number of snubbers found j which do not. meet the functional test acceptance criteria. The results from this sample plan shall be plotted using an," Accept" line which follows the equation N = 55(1 + C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point l plotted falls on or below the " Accept" line, testing of that type of I L snubber may be terminated. If the point plotted falls above the

Attachment 2 PY-CE!/NRR-2310L Page 46 of 85 Sheet 3 of 3 Perry Nuclear Power Plant Unit 1 l RELIEF REQUEST #IR-023, Rev. 1 l

3) " Accept" line, testing must continue until the point falls on or below the " Accept" line or all the snubbers of that type have been tested. The representative sample selected for the function test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure as far as practical that they are representative of the various configurations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same locations as snubbers which failed the previous functional test shall be retested at the time of the next functional but shall not be included in the sample plan, and failure of this functional test shall not be the sole cause for increasing the sample size under the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional testing results shall be reviewed at the time to determine if additional samples should be limited to the type of snubber which has failed the functional testing.

10 9

8 7

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SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST Figure 6.4.1.-1

Attachment 2 PY-CEI/NRR-2310L Page 47 of 85 Sheet 1 of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-024, Rev. 1 I. Identification of Components Class 1, Category B-F, Item B5-10, Pressure Retaining Dissimilar Metal Welds (see attached table for ID numbers).

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-F requires surface and volumetric examination of the areas and volumes defined by Figure IWB-2500-8.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from ultrasonically examining essentially 100% of the required volume for the components listed in the attached table (see the table for percent complete) for Perry's second 10-year interval.

IV. Basis for Relief Safe-end to safe-end extension welds of the Core Spray and Residual Heat Removal nozzles, which are inconel to carbon steel bimetallic welds, can not be effectively ultrasonically examined using conventional shear wave techniques.

To overcome the metallurgical properties impeding the conventional shear wave ultrasonic transmission, refracted longitudinal wave examinations are employed. The acoustic properties of refracted longitudinal wave propagation limit the technique to 1/2 vee path. The Code required volume necessitates either 1/2 vee path scanning from both sides of the weld or full vee path scanning from one side through the weld and required volume. Therefore, when joint geometry precludes adequate scan paths on both sides of a weld for 1/2 vee scanning, the perpendicular examination of the weld and required volume will be limited. For the subject safe-end to safe-end extension welds, a safe-end taper limits scanning from one side of the weld to approximately 60% resulting in an overall perpendicular examination completion percentage of approximately 80% (see Fig. IR-024-1 below).

The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines. There were no reportable indications during ASMC Section XI preservice or first interval inservice inspections.

Attachment 2 PY-CEI/NRR-2310L Page 48 of 85 Sheet 2 of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-024, Rev. 1 Although the volumetric examinations are limited, the most critical areas l of the weld and required volume are adequately covered. The root of the weld receives full two dimensional coverage and both the heat affected zones receive coverage which is essentially perpendicular to the end preparation. Additional assurance of the weld integrity is provided by the complete surface examinations.

Additional scanning of the subject welds with customized transducers and techniques could obtain some additional coverage. However, to do so would present undue hardship as the dose rates in the areas of the welds range from 400 mr/hr up to as much as 1 R/hr.

Since the construction, operating conditions and environmental conditions of the non-examined portion of the welds are identical to the examined portions, it is reasonable to apply satisfactory results from the examined to the non-examined portions.

The pressure boundary passed the required preservice hydrostatic and first interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject welds.

Design, procurement and operational provisions against nil ductile failure of the subject welds remain as decribed in the Perry USAR.

,. Revision 0 of this relief request, which requested the same relief for the subject welds for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. M75334, dated 2/14/92) l In summary, because of acceptable initial condition, successful code hydrotest and operating experience without related leakage indications, protection against brittle failure, the capability to ultrasonically examine at least 80% of the subject weld volume and perform complete surface examinations on a continuing basis, and the previous approval of j i similar relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

I' V. Alternate Examination Ultrasonic examinations, which meet all the requirements of ASME Section XI, 1989 Edition to the extent practical will be performed, i

E___________.________._____ _ _ _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ . _ _ _ . _ _ _ . _

__..________._.______j

Attachment 2 PY-CEUNRR 2310L Page 49 of 85 Sheet 3 of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-024, Rev. 1 Transducer Scan Transducer Scan Towards RPV , Away From RPV n ERPM99

[

Safe-E d Safe-End Pipe Extensio u u Examination Volume = ABCD Full Coverage Limited Coverage FIGURE IR-024-1

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%L P P P P P P  %  %  %  %  %

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/ / / / /

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Anachment 2 PY-CEl/NRR-2310L Page 51 of 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-025, Rev. 1 Z. Identification of Components Class 1, Category B-K-1, Item No. B10.10 integrally welded support attachments for piping (See attached table for ID numbers) .

II ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-K-1 requires a 100% surface examination of the areas defined by Figures IWB-2500-12 through 13.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from the required 100% surface examination of the support lug to process pipe attachment welds because access limitations from the surrounding guide structure prohibit surface preparation and examination of the attachment welds without disassembly of the guide. The relief is requested for Perry's second 10-year inspection interval.  ;

IV. Basis for Relief The welded attachments identified in the attached table are pipe lugs within large and complicated guide supports for the 26" main steam piping. Disassembly (and the subsequent reassembly) of the guides to provide access for the equired surface exams requires over 320 manhours for each guide in a general radiation area of approximately 10 mr/hr. Without disassembly, access is sufficient for VT-1 examination (utilizing mirrors and/or a fiberscope) of the welds. Utilization of j the VT-1 exams in lieu of surface exams maintains an adequate level of quality and '

safety without the hardships which would be incurred in disassembly.

The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject welds were examined in accordance with the appropriate Code requirements, weld techniques and welders were 'ualified in accordance with Code requirements, and materials were purchased and aced in accordance with the appropriate Code and NRC requirements and guidelines.

The pressure boundary passed the required preservice hydrostatic and first interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject welds.

l Design, procurement and operational provisions against nil ductile failure of the subject welds remain as described in the Perry USAR.

l L_

i i

Attachment 2 )

PY-CEl/NRR-2310L Page 52 0f 85 l

Sheet 2 of 3 I

Perry Nuclear Power Plant Unit 1 I RELIEF REQUEST #IR-025, Rev. 1 l Revision 0 of this relief request, which requested the same relief for the subject  !

welds for Perry's first 10-year inspection interval, was approved by the NRC  !

(reference TAC No. M75334, dated 2/14/92)

In summary, because of acceptable initial condition, successful code hydrotest and operating experience without related leakage indications, protection against j brittle failure, the capability to completely visually examine the welds, and the I previous approval of the same relief it is concluded that the specified ,

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

increase in the level of quality or safety.

j V. Alternate Examination l

VT-1 examinations will be performed, to the extent and frequency required by Table IWB-2500-1, in lieu of surface examinations.

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• l' AttIchment 2 PY-CEI/NRR-2310L Pagc 54 of 85 l

l Sheet 1 of.3 Perry Nuclear Power Plant Unit 1

[ RELIEF REQUEST #IR-026, Rev. 1 I. Identification of Components

-, . Class 2,' Category C-C, Item No. C3.20, integrally welded support attachments for piping (see attached table for ID numbers) . -

'II. ASME B&PV Section XI Requirements

-ASME'XI, 1989 Edition, Table IWC-2500-1, Category C-C requires a 100% surface examination of the. areas defined by' Figure IWC-2500-5.

L III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from the. required l 100% surface examination of the support lug to process pipe attachment. welds because access limitations from the surrounding guide structure prohibit surface preparation and examination of the attachment welds

~

without disassembly of the guide. The relief is requested for Perry's second 10-year inspection interval.

l l

IV. Basis for Relief The welded attachments identified in the attached table are pipe lugs within large and complicated guide supports for 26" main steam and 20" feedwater piping.

Disassembly (and the subsequent reassembly) of the guides to provide access for the' L required surface exams requires over 320 manhours for each guide in a general L radiation area of approximately 5 mr/hr. Without disassembly, access is sufficient for VT-1 examination (utilizing mirrors and a fiberscope) of the' welds.

Utilization of the VT-1 exams in lieu of surface exams maintains an adequate level of quality and safety without the hardships which would be incurred in disassembly.

, The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject welds'were examined in accordance with the appropriate Code requirements, weld techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines.

l The pressure' boundary passed the required preservice hydrostatic and first interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject welds.

LDesign, procurement and operational provisions against nil ductile failure of the subject welds remain as described in the Perry USAR.

. _ _ _ - = _ _ _ _ _ _ - _ - _ - - - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ - _ _ _ _ _ _ - _ - _ - _ _ - - _ _ _ _ - - ___-____ - _-_- - --_-_-_ __-____-_

l Att chment 2 PY-CEI/NRR-2310L Page 55 of 85 Sheet 2 of 3

( Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-026, Rev. 1 I

l Revision 0 of this relief request, which requested the same relief for the subject welds for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. M75334, dated 2/14/92)

In summary, because of acceptable initial condition, successful code hydrotest and operating experience without related leakage indications, protection against brittle failure, the capability to completely visually examine the welds, and the previous approval of the same relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination VT-1 examinations will be performed, to the extent and frequency required by Table IWC-2500-1, in lieu of surface examinations.

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l Attachment 2 PY-CEI/NRR-2310L Page 57 of 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-027, Rev. 1 I. Identification of Components l Class 3, Category D-B, Item D2.20, Integral Attachments: Component supports and l restraints. (See attached table for component identification).

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWD-2500-1, Category D-B requires visual examination of the integral attachments as defined by Figure IWD-2500-1.

III Relief Requested Pursuant to 10 CFR 50.55a(a) (3) (ii), relief is requested from performing direct l visual examination of the subject integral attachments d'le to access limitations.

l The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief I

l The structural integrity of the pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. All welds were inspected in accordance with the appropriate Code requirements. Weld techniques and welders were qualified in accordance with Code requirements and materials were purchased and traced in accordance with he appropriate Code and NRC requirements and guidelines.

The integrally attached (welded) anchors on the fuel oil day tanks are buried in fire retardant Pyrocrete in order to meet the PNPP fire protection program requirements per 10CFR50 Appendix R, and Branch Technical position APCSB 9.5-1, Appendix A (see PNPP USAR Appendix 9A, Section 90A.5 (D) (2) (1)) . Pyrocrete is a

[ hard, rigid material. When applied, it is considered as a permanent feature of the  ;

system to endure through the life span of the facility. To remove this material from the day tanks would require cutting and chipping.

1 Complete examinations meeting the requirements of the ASME Code Section XI, Category F-A, are performed on the accessible portion of two of the day tank l component supports. At the time of the support exams, the Pyrocrete covering their integral attachments is examined for any condition which might indicate that their integral attachments are structurally degraded (i.e., severely cracked or missing l Pyrocrete, support detached from component, etc.). First interval examinations

{

produced acceptable results with no visible signs of structural degradation. )

l 1

l 1

l l

1 i

l L

l Attachment 2 i l

PY-CEI/NRR-2310L Page 58 of 85 1

Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 -

RELIEF-REQUEST #IR-027, Rev. 1 The pressure boundary passed the required preservice hydrostatic and first interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication attributable to the subject welds.

Revision 0 of this relief request, which requested the same relief for the subject welds for Perry's first 10-year inspection interval, was approved by the NRC (reference TAC No. M84418, dated 2/24/94)

In summary, because of acceptable initial condition, successful code hydrotest and operating experience without related leakage indications, the capability to visually examine the pyrocrete for indications of degradation of the underlying attachment welds, and the previous approval of the same relief it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination At the time of the scheduled Category F-A visual examinations of the day tank anchors, the Pyrocrete covering their integral attachments will be examined for conditions which could indicate structural degradation of the buried integral attachment welds.

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PY-CEl/NRR-2310L Page 60 0f 85 Sheet 1 of 5 l

Perry Nuclear Power Plant Unit 1 l RELIEF REQUEST #IR-029, Rev. 1 l

l j I. Identification of Components l

Class 1, Category B-J (Item numbers in attached table), piping welds 4 inches NPS and greater.

f TI. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-J requires 100% l surface and volumetric examination of 25% of the circumferential butt welds and l their intersecting longitudinal welds. In accordance with Category B-J Note 1 of the Table, the examinations are to include all terminal ends, joints where the

, seismic and operational load stress levels exceed primary plus secondary stress intensity range of 2.4Sm (i .e. , "high stress" location) or a cumulative usage factor U of 0.4 (i.e., "high fatigue" location), all dissimilar metal welds, and additional welds (if necessary) such that the total number of circumferential butt l welds equals 25%.

IZI. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from selecting the welds in accordance with Note 1 when structural interference. make such selection impractical. The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The welds identified in the attached table are "high stress" welds, but examination i is impractical as they are in radiation areas and are encased in jet impingement shields. The jet shields are elbow or tee-shaped structural steel enclosures around Reactor Recirculation (RR) System and Main Steam (MS) System piping welds. The smallest of the RR jet shields weighs over 1600 lbs. 3 and is assembled with 48 bolts. Each of the MS jet shields weighs over 2240 lbs. j and is assembled with 180 bolts. The bolting for all of the jet shields is high {

strength, one time use, bolting that must be torqued to 10-16K. Disassembly, for inspection, and reassembly of these jet shields would be a labor-intensive effort with over 100 man-hours each. General area dose rates for the MS jet shield locations range from 20-50 mr/hr and contact dose rates for the RR piping beneath the RR jet shields range from 200-400 mr/hr. Therefore, removal of any of the jet l shields would require significant dose expenditure.

The structural integrity of the piping pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III, and additionally by meeting the requirements of ASME Section XI during preservice inspections. The subject RR and MS welds were examined (prior to installation of the jet shields) in accordance

Attachment 2 PY-CEI/NRR-2310L '

Pagc 61 of 85 -

' Sheet 2 of 5-Perry Nuclear Power Plant Unit 1

. RELIEF REQUEST #IR-029, Rev. l' with the appropriate Code requirements, weld _ techniques and welders were qualified in accordance with Code requirements, and materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines. There were no reportable indications during_.

preservice inspection. Additionally, the MS jet shields were removed i'n the first' inspection. interval (at considerable dose and monetary cost), the welds received inservice examinations, and they were found to be free of reportable indications.

-The pressure boundary passed the required preservice. hydrostatic and first interval inservice pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998 without leakage indication

-attributable to the subject welds.

Complete examinations meeting the requirements of the ASME Code Section XI

have been performed on similar "high stress" welds within the RR and MS Systems where~ jet shields are not present or are easily _ removed, with satisfactory results. These welds are subject to the same operating and

. environmental conditions as the obstructed welds.

Other RR and MS welds of the same size and configuration, but that are not "high' stress" welds, will be , examined in place of- the obstructed welds. In accordance with ASME Research White Paper, " Risk-Based Alternative Selection Process for Inservice Inspection of LWR Nuclear Power Plant Components," (Library of Congress 3 Catalog Number 94-71660) a recent industry survey, which included 50 nuclear units representing 733 cumulative' years of operation, found that there is no apparent relationship between the type of welds selected for inspection (i.e., high design stress / fatigue welds versus low stress / fatigue welds) and the detection of flaws.

Design, procurement and operational provisions against nil ductile failure of the subject welds remain as described in the Perry USAR.

Revision 0 of this relief request, which_ requested the same relief for the subject welds for Perry's first 10-year inspection interval, was approved by the NRC

(reference TAC No. M75334, dated 2/14/92)

^

In summary, because of the dose burden, acceptable initial condition, successful l Code hydrotest_and operating experience without related leakage indications, the i' satisfactory examination of identical welds, the substitution of welds of similar

. size and configuration, protection against brittle failure, and the previous approval of the same relief it is concluded that the specified requirements would result in hardship or unusual j

-difficulty without a compensating increase in the level of quality or safety.  ;

l l

l  !

L_i__-.____-._-.________-. . - _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ - . _ _ _ . _ _ _ _ - . . - - - _ - . . _ . - - _ . _ _ _ _ _ _ . - - _ _ _ _ - . . . _ _ _ _ _ _ _ _ _ _ _ - . - - - - - _ . _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Attachment 2 PY-CEI/NRR-2310L Page 62 of 85 Sheet 3 of 5 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-029, Rev. 1 V. Alternate Examination Welds of the same size and similar configuration, but that are not "high l stress" welds, will be examined in place of the obstructed welds to maintain the 25% selection requirement.

l i

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Attachment 2 PY-CEl/NRR-2310L Page 65 of 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-042, Rev. 0

1. Identification of Components Class 1, Category B-H, Item No. B8.10 Integrally Welded Attachments for Reactor Vessels, Perry component ID No. 1B13-CG, Reactor Vessel Bottom Head to Skirt Weld.

II. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-H requires surface examination of the areas on both sides of the integral attachment weld as defined by Figures IWB-2500-13.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining the inside surfaces of the RPV bottom head to skirt attachment weld due to limited access and dose concerns. The relief is requested for Perry's second 10-year 1 inspection interval.

{

IV. Basis for Relief l Examination of the inside surface of the RPV bottom head to skirt attachment weld 3 requires access inside the vessel skirt. Access is through an 18" x 24" manway and {

is considered a confined space entry. Once inside, rigid insulation panels must be disassembled and removed to provide access to the inside surface of the attachment weld. Then, within these tight confines, the weld surfaces must be mechanically cleaned of surface rust, dirt and scale. The dose rates inside the skirt (i.e.,

the bottom head area) range from 40 mr/hr to 100 mr/hr while the dose rates outside the skirt average about 5 mr/hr. Figure IR-042-1 depicts the skirt attachment weld j

configuration.

The structural integrity of the pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject ,

weld was examined in accordance with the appropriate Code requirements. Weld I techniques and welders were qualified in accordance with Code requirements.

Materials were purchased and traced in accordance with the appropriate Code and NRC ]

requirements and guidelines.

Code Case N-323, which has been annulled, prev'iously allowed examination of the vessel to skirt weld from only the outside surface if the stress intensities at the inside surface of the weld did not exceed 80% of the Levels A, B, C, and D Service l Limits (NB-3000) and the cumulative useage factor U [NB-3222. 4 (e) (5) ] did not exceed 0.1. Perry's skirt weld met these requirements with a stress intensity of 23.4 KSI versus an allowable of 80.1 KSI and a useage factor of 0.456.

t l

t C______._---____._____________________-__-_____-_-------------. - - - - - - - - - - - - - - - --

Anachment 2 PY-CEl/NRR-2310L Page 66 of 85 Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-042, Rev. O I

Design, procurement and operational provisions to protect against nil ductile failure of the subject welds remain as described in the Perry USAR.

For reactor vessel integral attachments that are configured like that of Figure IWB-2500-14, Note 4 of Table IWB-2500-1, Category B-H allows volumetric examination from the outside surface in lieu of surface examination. In a similar manner, Perry will supplement the outside surface examination of the vessel to skirt weld with ultrasonic examination. The supplemental ultrasonic examination will not cover the entire weld volume, but will provide coverage of the weld and heat affected zones at the inside surface. Py performing all the vessel to skirt weld examinations from the outside surface, it is estimated that as much as 1 REM can be saved over the course of the inspection interval.

In summary, because of the acceptable initial condition, low stress intensity ar.d useage factor at the inside surface of the weld, protection against brittle failure, the capability to examine 100% of the required outside surface areas, and the performance of supplemental ultrasonic examinations it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination Surface examinations which meet all the requirements of ASME Section XI, 1989 i Edition will be performed from the outside surface of the vessel to skirt weld. In l

lieu of performing the required surface examination of the'inside surface of the vessel to skirt weld, ultrasonic examinations that will provide coverage of the

' inside surface will be performed.

1 a s l

l l

l l'

l____ _.

Attachment 2 PY-CEl/NRR-2310L Page 67 of 85 Sheet 3 of 3 Perry Nuclear Power Plant Unit 1 l RELIEF REQUEST #IR-042, Rev. 0 l

1 I

Vessel to Skirt Weld (1813-CG)

Reactor Vessel Bottom Head Supplemental UT _ g From this surface g Skirt  :

Rigid

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Figure IR-042-1 1

Attachment 2 PY-CEUNRR-2310L Page 68 of 85 Sheet 1 of 2 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-043, Rev. O I. Identification of Components Class 1, Category B-M-1, Item No. B12.30, Valve Body Welds (valves, less than NPS 4), Perry component ID No. 1G33-F101-SEAM, ISI Isometric drawing SS-305-671-108.

ZZ. ASME B&PV Section XI Requirements ASME XI, 1989 Edition, Table IWB-2500-1, Category B-M-1 requires examination of the valve body welds as defined by Figures IWB-2500-17. Note 3 of the table states that the examinations are limited to at least one valve within each group of valves that are of the same size, construction design (such as globe, gate, or check valves), and manufacturing method, and that perform similar functions in the system (such as containment isolation or overpressure protection).

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested from examining the subject valve body weld due to radiological concerns. The relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief The subject valve is a 3" maintenance isolation valve within the Reactor Water Cleanup System. It is a Borg Warner gate valve with a forged body welded to a forged neck. The valve is located downstream of the reactor bottom head drain and has had contact dose rates ranging from 10 REM to in excess of 100 REM dependant upon system operational and flush status. As such, the valve represents a severe radiological hazard and it is heavily shielded during refueling outages. It is the only 3" maintenance isolation valve in the Reactor Water Cleanup System. As such, and interpreting Note 3 of Table IWB-2500-1, Category B-M-1 literally, it is in a group by itself and its body weld requires examination.

.The structural integrity of the pressure boundary was demonstrated during construction by meeting the requirements of the ASME Code Section III. The subject weld was examined in accordance with the appropriate Code requirements, seld j techniques and welders were qualified in accordance with Code requirements, and {

materials were purchased and traced in accordance with the appropriate Code and NRC requirements and guidelines.

The pressure boundary passed the required hydrostatic test and first interval I inservice system pressure tests, and has operated for a total of about 2,662 equivalent full power days between November 1987 and March 1998, without leakage .

indication attributable to the subject weld.

l Attachment 2 PY-CEI/NRR-2310L Page 69 0f 85

)

Sheet 2 of 2 l

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #IR-043, Rev. O Design, procurement and operational provisions against nil ductile failure of the subject weld remain as described in the Perry USAR.

There are two other maintenance isolation valves within.the Reactor Water Cleanup i

System that are also Borg Warner gate valves with welded necks. They are 1G33-F100 and 1G33-F106. At 4", they are only slightly different in size than the 1G33-F101 valve. The contact dose rates on these valves are considerably less than those of the 1G33-F101 valve. Therefore, in lieu of categorizing the 1G33-F101 valve in a group by itself, it will be grouped with the 1G33-F100 and 1G33-F106 valves. Of these 3 valves, 1G33-F100 has exhibited the lowest contact dose rates and will be the valve selected for examination in accordance with Note 3 of Table IWB-2500-1.,

Category B-M-1. Also in accordance with the table, the 4" body weld of 1G33-F100 will receive a volumetric examination. In accordance with IWB-2430, Additional Examinations, should examination of the body weld of 1G33-F100 reveal indications exceeding the acceptance standards of Table IWB-3410-1, the examinations will be extended to the 1G33-F106 and 1G33-F101 valves.

In summary, because of the acceptable initial condition, successful preservice hydrotest and first interval inservice pressure tests, protection against brittle failure, and the ability to fully examine another valve that is of essentially the same design, manufacturer and function it is concluded that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

V. Alternate Examination In lieu of categorizing the 1G33-F101 valve in a group by itself, it will be grouped with the 1G33-F100 and 1G33-F106 valves. Of these 3 valves, 1G33-F100 has exhibited the lowest contact dose rates and will be the valve selected for examination in accordance with Note 3 of Table IWB-2500-1, Category B-M-1. Also in accordance with the table, the 4" body weld of 1G33-F100 will receive a volumetric examination. In accordance with IWB-2430, should examination of the body weld of l 1G33-F100 reveal indications exceeding the acceptance standards of Table IWB-3410-l 1, the examinations will be extended to the 1G33-F106 and 1G33-F101 valves.

l l

Attachment 2 PY-CEl/NRR-2310L Page 70 0f 85 Sheet 1 of 7 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-001, Rev. 1 I. Identification of Components Class 2 systems / components attached to the Reactor Coolant Pressure Boundary (Class

1) which are not provided with either pressure or test isolation (i.e.,

instrumentation, drain, vent, and test piping). A list of valve numbers identifies the affected components (i.e., valves, piping systems and instruments).

'II . 'ASME B&PV Section XI Requirements The subject systems / components are required to operate during normal plant operation. Thus, In accordance with ASME XI, 1989 Edition, IWC-5210 (a) (2) , Test, the' system pressure test prescribed would be a system inservice test. Furthermore, in accordance with IWA-5211(c), Test Condition Holding Time, there is no hold time required, provided the system has been in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

III. Relief Request Pursuant to 10 ' CFR 50.55a (a) (3) (i), relief is requested from operating l the system for four hours before commencing the VT-2 examinations for those Class 2 systems / components that are non-isolable from the Reactor Coolant Pressure Boundary (Class 1). These components shall be examined (VT-2 Visual Examination) during the Class.1 Reactor Coolant Boundary System Leakage Test at the frequency intervals specified within l Subsection IWC. Thus, this relief request proposes substituting IWA-5213(a) for IWA-5213(c) and IWB-5210 (a) (1) for IWC-5210 (a) (2) . The relief is requested for Perry's second 10-year inspection interval IV. Basis for Relief Numerous components attached to the reactor coolant pressure boundary are covered by the provisions of 10CFR50.55a(c) Reactor Coolant Pressure Boundary. The following excerpt from 10CER50a(c) is provided:

"(2) Components which are connected to the reactor coolant system and are part of the reactor coolant pressure boundary as defined in Section 50.2 need not meet the requirements of paragraph (c) (1) of this section, Providod:

(1) In the event of postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system; or L____.__________.____ _ _ _ _ - -

Attachment 2 PY-CEl/NRR-2310L Page 71 of 85 Sheet 2 of 7 Perry Naclear Power Plant 1 RELIEF REQUEST #PT-001 (ii) The componant is or can be isolated from the reactor coolant system by two valves in series (both closed, both open, or one closed and the other-open). Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only."

The piping systems and their associated components connected to the reactor coolant pressure boundary and less than 1 inch in diameter were constructed to the requirements of ASME Code,Section III, Subsection NC, and identified as Safety Class 2 for inservice inspection. The associated components and component parts are identified by valve number and listed below. These piping systems shall be pressurized during the Class 1 reactor coolant pressure boundary System Leakage Test and a VT-2 Visual Examination will be performed. The System Leakage Test frequency and pressure will be that required for a Class 2 System Inservice Test.

Although the system will not have been in operation for four hours prior to commencing the examinations, the time required to bring the reactor coolant system up to test pressure will allow for the detection of leakage.

Within ASME Section XI the test conditions (i.e., pressure, temperature and hold time) between the reactor coolant pressure boundary and other safety systems are different. Although there are differences, all the system pressure tests ensure leak tightness. Therefore, the substitution of IWA-5213(a) for IWA-5213(c) and the substitution of IWB-5210 (a) (1) for IWC-5210 (a) (2) satisfies the intent of the Code.

The NRC (reference TAC No. M75334, dated 2/14/92) approved revision 0 of this relief request, which requested the same relief for Perry's first 10-year inspection interval.

V. Alternate Examination For those Class 2 systems / components attached to the Reactor Coolant Pressure Boundary (Class 1) which are not provided with either pressure or test isolation, pressure testing will be conducted in accordance with IWA-5213(a) and IWB-5210(a)(1) in lieu of IWA-5213(c) and IWC-5210 (a) (2) .

Attachment 2 PY-CEI/NRR-2310L Page 72 of 85 Sheet 3 of 7 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-001, Rev. 1 Valve No. Description P&ID No.

1B33-F068A/B Recirc Pump A/B Discharge Valve Vent D-302-601, 602 1B33-F070A/B Recirc Pump A/B Discharge Valve Drain D-302-601, 602 1B33-F065A/B Recirc Loop A/B FCV Drain D-302-601, 602 IB33-F647A/B Recirc Loop A/B FCV Vent D-302-601, 602 IB33-F686A/B Recire Loop A/B FCV Drain D-302-601, 602 g 1B33-F027A/B Recirc Pump A/B Suction Valve Drain D-302-601, 602 1B33-F503A/B Instrument Isolation Valves for dPT-NO15A/B, D-302-602

-F504A/B Respectively 1B33-F505A Instrument Isolation Valves for FT-N014C/D D-302-602

-F506A 1B33-F505B Instrument Isolation Valves for FT-N011B and D-302-602 1

-F506B FT-N024C/D 1B33-F507A Instrument Isolation Valves for FT-N011A and D-302-602

-F508A FT-N014A/B 1B33-F507B Instrument Isolation valves for FT-N024A/B D-302-602

-F508B 1B33-F512A/B Recirc Pump A/B Diff Pressure Instrument Vent D-302-602 1B33-F513A/B Recirc Pump A/B Diff Pressure Instrument Vent D-302-602 1B33-F577 Recirc Loop B Flow Instrument Vent D-302-602 1833-F578 Recirc Loop B Flow Instrument Vent D-302-602 1B33-F579 Recirc Loop A Flow Instrument Vent D-302-602 1D33-F580 Recire Loop A Flow Instrument Vent D-302-602 1B33-F581 Recirc Loop B Flow Instrument Vent D-302-602 1B33-F582 Recirc Loop B Flow Instrument Vent D-302-602 1B33-F583 Recirc Loop A Flow Instrument Vent D-302-602 1B33-F584 Recirc Loop A Flow Instrument Vent D-302-602 1833-F059 Recirc System Sample Isolation D-302-602 1B33-F019 Reactor Water Sample Isolation D-302-602 1B33-F110 Rx Recirc Sample Line Drain D-302-602 1G33-F507 Instrument Isolation Valve for FT-NO37 D-302-671 1G33-F523 RWCU Bottom Head Flow Instrument Vent D-302-671 1E32-F506A Instrument Isolation Valves for PT-N051A, D-302-341

-544A PT-N061A 1E32-F506E Instrument Isolation Valves for PT-N051E, D-302-341

-F544E PT-N061E 1E32-F506J Instrument Isolation Valve for PT-N051J, D-302-341

-F544J' PT-N061J IE32-F506N Instrument Isolation Valve for PT-N051N D-302-341

-F544N PT-N061N 1B21-F596 1B21-F016 Test Connection Root Valve D-302-121 1R21-F017 MST Drain and MSIV Bypass Line Drain D-302-121 IN27-F551A/B/C Feedwater Header A Branch Test Isolation D-302-082 1N27-F551D/E/F Feedwater Header B Branch Test Isolation D-302-082 IN27-F557A/B Feedwater Header A/B First Test Connection D-302-082 1G33-F508A/B Instrument Isolation valves for PT-N076A, D-302-671, 962 PT-N076B 1G33-F108 Pen 131 INBD Test Conn First Isolation Valve D-302-671 1E31-F541B RWCU Diff Flow LD High Side Test Connection D-302-962 1E51-F528A/B/C/D Instrument Isolation Valves for PT-N084A/B, D-302-632, 961 PT-N085A/B L___-______---______________

Attachment 2 PY-CEl/NRR-2310L Page 73 of 85 Sheet 4 of 7 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-001, Rev. 1 Valve No. Description P&ID No.

1E31-F542A/B RCIC/RHR ST Supply LD Low Stby Test Conn D-302-961 1E31-F543A/B RCIC/RHR ST Supply LD High Stby Test Conn D-302-961 1E31-N084B-G Cross-Tie Low Side PT-N084A/B D-302-961 1E31-N084B-R Cross-Tie High Side PT-N084A/B D-302-961 1E31-F519 Instrument Isolation Valve For PT-N080A D-302-705, 962 1E31-F545A RHR A to LPCS LD High Side Test Connection D-302-962 1E31-F523 Instrument Isolation Valve for PT-N081 D-302-701, 962 1E31-F547 HPCS to SLC Ref Diff Pressure Test Connection D-302-962 1E31-F520, Instrument Isolation Valve for PT-N080A D-302-642, 962 1E31-F544A RHR A to LPCS LD Low Side Test Connection D-302-962 1E31-F521 Instrument Isolation valve for PT-N080B D-302-642, 962 1E31-F522 Instrument Isolation valve for PT-N080B D-302-642, 962 1E21-F502 LPCS to Rx Line Test Connection D-302-705 l 1E22-F501 HPCS to Rx Line Test connection D-302-701 1C41-F501 SLC Discharge Line Inboard Drywell Drain Vlv D-302-691 )

1E12-F508A LPCI From RHR A Inbd First Test Connection D-302-642 1E12-F508B LPCI From RHR B Inbd First Test Connection D-302-642 IE12-F508C LPCI From RHR C Inbd First Test Connection D-302-642 1E12-F501 Shutdown Cooling Suction Hdr Inbd First Conn D-302-642 1E51-F072 RHR & RCIC Steam Supply Line Test Connection D-302-632 IB33-F514 Recire Jet Pump 15 Flow Instrument Vent D-302-604 IB33-F515 Recirc Jet Pump 12 Flow Instrument Vent D-302-604 1B33-F516 Recirc Jet Pump 18 Flow Instrument Vent D-302-604 IB33-F517 Recirc Jet Pump 19 Flow Instrument Vent D-302-604 IB33-F518 Recirc Jet Pump 15 Flow Instrument Vent D-302-604 1933-F519 Recirc Jet Pump 16 Flow Instrument Vent D-302-604 i 1833-F520 Recirc Jet Pump 11 Flow Instrument Vent D-302-604 1833-F521 Recirc Jet Pump 17 Flow Instrument Vent D-302-604 l 1B33-F522 Recire Jet Pump 13 Flow Instrument Vent D-302-604 1B33-F523 Recirc Jet Pump 20 Flow Instrument Vent D-302-604 1833-F524 Recirc Jet Pump 20 Flow Instrument Vent D-302-604 IB33-F525 Recirc Jet Pump 14 Flow Instrument Vent D-302-604 1833-F526 Recire Jet Pump 15 Flow Instrument Root D-302-604 FT-NO38B, LT-N044D 1833-F527 Recirc Jet Pump 12 Flow Instrument Root D-302-604 FT-N037F 1B33-F528 Recirc Jet Pump 18 Flow Instrument Root D-302-604 FT-NO37M 1933-F529 Recirc Jet Pump 19 Flow Instrument Root D-302-604 FT-NO37S 1B33-F530 Recirc Jet Pump 15 Flow Instrument Root D-302-604 FT-N037U, FT-N038B 1833-F531 Recirc Jet Pump 16 Flow Inst Root FT-NO37D D-302-604 1833-F532 Recirc Jet Pump 11 Flow Inst Root FT-N037B D-302-604 1933-F533 Recirc Jet Pump 17 Flow Inst Root FT-N037H D-302-604 1833-F534 Recire Jet Pump 13 Flow Inst Root FT-N037K D-302-604 1833-F535 Recirc Jet Pump 20 Plow Inst Root FT-NO38D D-302-604 1B33-F536 Recirc Jet Pump 20 Flow Inst Root FT-N037W, D-302-604 FT-N0380 l

l l

Attachment 2 PY-CE!/NRR-2310L

. Page 74 of 85 Sheet 5 of 7 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-001, Rev. 1 Valve No. Description P&ID No.

1B33-F537- Recirc Jet Pump 14 Flow Inst Root FT-N037P D-302-604 1B33-F646' Jet Pump Post Accident Sample Isolation D-302-604 1P87-F001 Reactor Recirc B Sample Isolation Valve- D-302-431 1B33-F538' Recirc Jet' Pump 7 Flow Instrument' Vent D-302-603 IB33-F539 Recirc Jet Pump 9 Flow Instrument Vent D-302-603 1B33-F540 Recirc Jet Pump 10 Flow Instrument Vent D-302-603

.1B33-F541 Recirc Jet Pump 1 Flow Instrument VentJ D-302-603 1B33-F542 Recirc Jet Pump 2 Flow Instrument Vent D-302-603 1B33-F543 Recire Jet Pump 5 Flow Instrument Vent D-302-603

-1B33-F544' Recirc Jet Pump 3 Flow Instrument Vent D-302-603 1B33-F545' Recirc Jet Pump 10 Flow Instrument Vent D-302-603 IB33-F546 Recirc Jet Pump 5 Flow Instrument Vent D-302-603

-1B33-F547 Recirc Jet Pump 4 Flow Instrument Vent D-302-603-

'1B33-F548 -Recirc Jet Pump 6 Flow Instrument Vent D-302-603

-1B33-F549 Recirc Jet Pump 8 Flow Instrument Vent D-302-603 1933-F550 Recirc Jet Pump 7 Flow Instrument Root D-302-603 FT-N037G 1B33-F551 Recire Jet Pump 9 Flow Instrument Root D-302-603-FT-NO37R 1B33-F552 Recirc Jet Pump 10 Flow Instrument Root D-302-603 FT-NO37V, ET-N038C 1B33-F553 Recirc Jet Pump 1 Flow Instrument Root D-302-603 FT-NO37A 1B33-F554- Recirc' Jet Pump 2 Flow Instrument Root D-302-603 FT-NO37E 1B33-F555 Recire Jet Pump 5 Flow Instrument Root D-302-603 FT-N038A, LT-N044C

-1B33-F556 'Recirc Jet Pump 3 Flow Instrument Root D-302-603 FT-N037J 1B33-F557 Recirc Jet Pump 10 Flow Instrument Root D-302-603 FT-NO38C 1833-F558 Recirc. Jet Pump 5 Flow Instrument Root D-302-603 FT-NO37T, FT-NO38A 1B33-F559 Recire Jet Pump 4 Flow Instrument Root D-302-603 FT-NO37N 1833-F560 Recirc Jet Pump 6 Flow Instrument Root D-302-603 FT-NO37C 1833-F561 Recirc Jet Pump 8 Flow Instrument Root D-302-603 FT-N037L-1833-F570 Jet Pump Flow Instrument Vent D-302-603 1B33-F571 Jet Pump Flow Instrument Isolation FT-N037G, D-302-603 FT-NO37R, FT-N037V, FT-N037A, FT-N 037 E, FT-N037J, FT-NO37T, FT-NO37N, FT-NO37C, FT-NO37L 1833-F645 Jet Pump Post Accident Sample Isolation D-302-603 1P87-F007 Reacter Recire A Sample Isolaton Valve D-302-431 1E31-F503- Instrument Isolation Valves for PT-N003A, D-302-961

-F504 PT-N086A, PT-N086B j

u__ - .

Attachment 2 PY-CEl/NRR-2310L Page 75 of 85 Sheet 6 of 7

' Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-001, Rev. 1 Valve No. Description P&ID No.

1E31-F505 Instrument Isolation Valves for PT-N086C, D-302-961

-F506 PT-N086D 1E31-F507 Instrument Isolation Valves for PT-N003B, D-302-961

-F508 PT-N087A, PT-N087B 1E31-F509 Instrument Iaolation Valves for PT-N087C, D-302-961

-F510 PT-N087D 1E31-F570 Main Steam Line A Flow Instrument Test Conn D-302-961 1E31-F571 Main Steam Line A Flow Instrument Test Conn D-302-961 1E31-F572 Main Steam Line A Flow Instrument Test Conn D-302-961 1E31-F573 Main Steam Line A Flow Instrument Test Conn D-302-961 1E31-F574 Main Steam Line B Flow Instrument Test Conn D-302-961 1E31-F575 Main Steam Line B Flow Instrument Test Conn D-302-961 1E31-F576 Main Steam Line B Flow Instrument Test Conn D-302-961 1E31-F577 Main Steam Line B Flow Instrument Test Conn D-302-961 1E31-F511 Instrument Isolation Valves for PT-N088A, D-302-961

-F512 PT-N088B 1E31-F513 Instrument Isolation Valves for PT-N003C, D-302-961

-F514 PT-N088C, PT-N088D 1E31-F515 Instrument Isolation Valves for PT-N089A, D-302-961

-F516 PT-N089B 1E31-F517 Instrument isolation Valves for PT-N003D, D-302-961

-F518 PT-N089C, PT-N089D

-1E31-F578 Main Steam Line C Flow Instrument Test Conn D-302-961 1E31-F579 Main Steam Line C Flow Instrument Test Conn D-302-961 1E31-F580 Main Steam Line C Flow Instrument Test Conn D-302-961 1E31-F581 Main Steam Line C Flow Instrument Test Conn D-302-961 1E31-F582 Main Steam Line D Flow Instrument Test Conn D-302-961 1E31-F583 Main Steam Line D Flow Instrument Test Conn D-302-961 1E31-F584 Main Steam Line D Flow Instrument Test Conn D-302-961 1E31-F585 Main Steam Line D Flow Instrument Test Conn D-302-961 1821-F512 Instrument Isol Valve for LT-N027, LT-N017 D-302-606 1B21-F514 Instrument Isol Valve for LT-N095B, PT-N403B, D-302-606 PI-R004B, PT-N058, PT-N403F, PT-N068B, PT-N008B, PT-N068F, PT-N040, PT-N078B, PT-N062B, PT-N004B, LT-N0808, LT-N490, LT-N091B, LT-N402B, LT-N091F, dPI-R0098, LT-N081B 1821-R011 B-H Reference Leg Fill Line D-302-607 IB21-R011 B-G Reference Leg Fill Line D-302-607 IB21-F510 Instrument Isolation Valve for PT-N078D, D-302-606 LT-N080D, LT-N073L, LT-N073R, LT-N081D, LT-N402F, LT-N044D 1821-R011 D-H Reference Leg Fill Line D-302-607 IB21-R011 D-G Reference Leg Fill Line D-302-607 IB21-F542 RPV Level Instrument Line Drain D-302-606 1B21-F511 Instrument Isolation Valve for LT-N080D, D-302-606 dPI-R005 IB21-F544 RPV Level Instrument Line Vent D-302-606 IB21-F546 RPV Level Instrument Line Drain D-302-606 L

Attachment 2 PY-CEUNRR-2310L Page 76 0f 85 Sheet 7 of 7 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-001, Rev. 1 Valve No. Description P&ID No.

1B21-F515 Instrument Isolation Valve for LT-N080B, D-302-606 LT-N004, LT-N017, LT-N027, LT-N095B 1821-F551 RPV Level Instrument Line Vent D-302-606 1B21-F540 RPV Level Instrument Line Drain D-302-606 1B21-F545 RPV Level Instrument Line Vent D-302-606 1B21-F509 Instrument Isolation Valve for LT-N073L, D-302-606 LT-N073R, LT-N081D, LT-N402F l 1821-F548 RPV Level Instrument Line Drain D-302-606 1B21-F549 RPV Level Instrument Line Vent D-302-606 1B21-F513 Instrument Xsolation Valve for LT-N081B, D-i42-606 LT-N091F, dPI-R009B, LT-N402B, LT-N091B 1B21-F583 Instrument Isolation Valve for PT-N081, D-302-606, 962 l

dPT-N032  :

1B21-F582 Jet Pump Instrument Line Vent D-302-606 1821-F585 Instrument Isolation Valve For dPT-N011, D-302-606, 872 dPT-N008 1821-F523 Instrument Isolation Valve for Flow D-302-606, Instruments P009, dPI-R005, LT-N490, 604, 671~

dPT-N032, FT-NO37, FT-N032, dPI-R005 1B21-F584 Jet Pump Instrument Line Vent D-302-606 1B21-F553 Instrument Isolation Valve for LT-N095A, D-302-606 PT-N403A, PI-R004A, PT-N403E, PT-N005, PT-N068A, PT-N050, PT-N068E, PT-N006, PT-N008A, PT-N078A, PT-N062A, LT-N004A, LT-N080A, LT-N010, LT-N091A, LT-N402A, dPI-R009A, LT-N091E, LT-N081A 1821-R011 A-H Reference Leg Fill Line D-302-607 1B21-R011 A-G Reference Leg Fill Line D-302-607 IB21-F505 Instrument Isolation Valves for LT-N080C, D-302-606 PT-N078C, LT-N004C, LT-N073G, LT-N402E, LT-N073C, LT-N081C, LT-N044C IB21-R011 C-H Reference Leg Fill Line D-302-607 1821-R011 C-G Reference Leg Fill Line D-302-607 1821-F536 RPV Level Instrument Line Drain D-302-606 1B21-F506 Instrument Isolation Valve for LT-N080C, D-302-606 LT-N004C 1821-F539 RPV Level Instrument Line Vent D-302-606 1821-F528 RPV Level Instrument Line Drain D-302-606 1821-F552 Instrument Isolation Valve for LT-N080A, D-302-606 LT-N004A, LT-N095A 1821-F533 RPV Level Instrument Line Vent D-302-606 1821-F535 RPV Level Instrument Line Drain D-302-606 1821-F504 Instrument Isolation Valves for LT-N081C, D-302-606 LT-N073C, LT-N402E, LT-N073G 1821-F534 RPV Level Instrument Line Vent D-302-606 1821-F529 RPV Level Instrument LIne Drain D-302-606 1B21-F555 Instrument Isolation Valve for LT-N081A, D-302-606 LT-N091E, dPI-R009A, LT-N402A, LT-N091A, LT-N010 1B21-F531 RPV Level Instrument Line Vent D-302-606 l

L______._

Attachment 2 PY-CEI/NRR-2310L l Page 77 of 85 l Sheet 1 of 2 Perry Nuclear Power Plant Unit 1

]

RELIEF REQUEST PT-006, Rev. 1 l Code Case N-546 I. Identification of Components l i Class 1, 2, and 3 systems subject to pressure testing.

II. ASME Boiler & Pressure Vessel Code Section XI Requirements l

Tables IWB-2500-1, IWC-2500-1 and IWD-2500-1 of the 1989 Edition of Section XI require the VT-2 examination method for pressure testing.

IWA-2312 requires that personnel performing visual examinations (e.g., VT-2 examination) be qualified and certified to comparable levels of competency I as defined in SNT-TC-1A and the Owner's written practice.

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (i), relief is requested from requiring that the personnel performing the VT-2 examinations be qualified and certified to comparable levels of qualification as defined in SNT-TC-1A.

The relief is requested for Perry's second 10-year inspection interval. l IV. Basis for Relief I The use of Code Case N-546 will eliminate the need to treat VT-2 examination personnel as NDE personnel. The Abstract of SNT-TC-1A states, "This standard applies to personnel whose specific tasks or jobs require appropriate knowledge of the technical principals underlying nondestructive testing (NDT) methods for which they have responsibilities within the scope of VT-1 and VT-3 examination methods." VT-2 requires no special knowledge of technical principals underlying its performance. It is simply the straight forward examination for leakage. No special skills or technical training are required in order to observe water dripping from a component or bubbles forming on a joint wetted with leak detection solution. As such, qualification in accordance with the provisions of the Code Case does not present any reduction in quality or safety. In fact, it will facilitate the qualification of those personnel most familiar with the walkdown of plant systems.

Additionally, there is a cost benefit of approximately $12,000 per operating cycle' realized by eliminating the formal certification of Perry and contracted VT-2 examination personnel.

Attachment 2 PY JEI/NRR-231OL Page 78 of 85 Sheet 2 of 2 i Perry Nuclear Power Plant Unit 1 i RELIEF REQUEST PT-006, Rev. 1 Code Case N-546 In summary, approval of this request would be in accordance with 10 CFR 50.55a (a) (3) (i), as compliance with Code Case N-546 will provide an essentially equivalent alternative to the IWA-2300 requirements. It would also provide relief from the administrative and financial burdens of certification, which do not provide any compensating increase in the level of quality or sa'fety.

V. Alternate Examination CEI proposes to perform VT-2 examinations utilizing personnel qualified in accordance with the provisions of Code Case N-546 in lieu of personnel who are qualified and certified to comparable levels of qualification as defined in SNT-TC-1A. The qualification provisions specified in Code Case N-546 are as follows:

(1) Personnel must have at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of plant walkdown experience, such as that gained by licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel.

(2) Personnel must receive at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of training on Section XI requirements and plant specific procedures for VT-2 examination.

(3) Personnel must meet the vision test requirements of IWA-2321, 1995 Edition.

Training and qualification of VT-2 personnel will be documented and the j

records will be maintained. l Additionally, to provide for consistent, quality VT-2 visual examinations, the examinations will be performed using standard procedures. An independent review and evaluation of the VT-2 visual examination results will be performed and documented on the examination records.

l I 1 i I

Attachment 2 PY-CEl/NRR-2310L Page 79 0f 85 Sheet 1 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST PT-007, Rev. 1 I. Identification of Components 1

[

Class 3 safety or relief valve piping which discharges into the containment pressure suppression pool.

II. ASME Boiler & Pressure Vessel Code Section XI Requirements Hydrostatic pressure tests per ASME XI, 1989 Edition I, section IWD-5223(f).

III. Relief Requested Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested frem performing the Code required hydrostatic tests. An alternative pressure test is proposed. This relief is requested for Perry's second 10-year inspection interval.

IV. Basis for Relief ASME Code,Section XI, Table IWD-2500-1, requires hydrostatic testing of Class 3 pressure retaining components. For safety or relief valve piping which discharges into the containment pressure suppression pool, section IWD 5223(f) states that a pneumatic test (at a pressure of 90% of the pipe submergence head of water) that demonstrates leakage integrity shall be performed in lieu of system hydrostatic tests. l The reactor coolant system has a total of 19 safety / relief valve discharge lines. These lines are used to direct steam from the main steam lines to the suppression pool allowing removal of the latent heat through condensing of the steam within the suppression pool. The valve discharge piping and associated valves (i.e., vacuum breakers) are designated ISI Safety Class.3 from the safety / relief valve discharge ports to the end of the submerged quenchers. Each line's multiple vacuum breakers (i.e.,

simple check valves) eliminate the pressure differential created between the drywell atmosphere and piping following a safety / relief valve actuation. Therefore, the discharge piping pressure retaining boun'ary d is not leaktight. j The safety / relief valves are routinely (i.e., each refueling outage) used during valve testing and expecter' to see service during unplanned plant j transient conditions -(i.e., reataor scrams). The proper operation of the relief system ensures the integrity of the piping to perform its design function.

l Attachment 2 i

PY-CEl/NRR-2310L Page 80 0f 85

. Sheet 2 of 3 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST PT-007, Rev. 1 The Cleveland Electric Illuminating Company (CEI) has determined that performing hydrostatic testing results in a hardship without a compensating increase in the level of quality and safety. The following hardships would be encountered with the performance of hydrostatic testing in accordance with the Code. First, the hydrostatic. test pressure conditions are unique to these discharge lines, and therefore, special test equipment will need manufacturing (e.g., blank flanges with test ports). Additionally, the VT-2 examination during the pressurization of the pneumatic test pressure boundary would include the application of leak checking solution (i.e., snoop) to over 35 welds and mechanical connections for each of the 19 discharge lines. To perform the snooping approximately 35 feet of scaffolding would be needed for each of the 19 discharge pipes within the drywell structure. Portions of the drywell scaffolding would have to be erected in high radiation fields. The total effect on radiation exposure can not be easily estimated due to the massive task being undertaken. However, dose calculations have been estimated as an additional 5.0 manrem, for the staging (i.e., scaffolding) work, for the examinations and test equipment, for the performance of the VT-2 exams, and for restoration. Therefore, preparation and performance of a system hydrostatic tests at 90% of submergence pressure (i.e., l approximately 5.4 psig) involves considerable time, expense, manpower and radiation dose without a compensating increase in the level of quality or safety.

Since the 1989 Edition, the ASME Subcommittee XI concluded that the l requirements of section IWD-5223(f) served no useful purpose and the pressure test has been exempted.Section XI, section IWD-5223(f),

1992 Addenda, states that open ended Class 3 safety or relief valve discharge lines including safety and relief valve piping which discharges into the containment pressure suppression pool, are exempt from hydrostatic test.

Industry experience has demonstrated that inservice leaks are not discovered as a result of hydrostatic pressures propagating an existing flaw through-wall. Also, since the purpose of these discharge lines is to direct steam flow, and not provide a leaktight barrier, determining the location of flaws would not provide a compensating increase in the level of quality and safety.

The safety / relief valve discharge lines are basically open ended piping, that function to direct the steam flow to the quencher, and are not a leaktight pressure retaining boundary. Therefore, rather than performing the 90% submergence pressure test of Section XI, section IWD-5223(f),

1989 Edition, performing the requirements for open ended portions of l discharge lines as stated in section IWD-5223(d) of the same Code Edition would be appropriate. Section IWD-5223(d) requires confirmation of adequate flow during system operation in lieu of performing system hydrostatic testing.

I ,

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f Attachment 2 I PY-CEI/NRR-2310L  !

Page 81 of 85 Sheet 3 of 3

)

l Perry Nuclear Power Plant Unit 1 RELIEF REQUEST PT-007, Rev. 1 V. Alternate Examination As an alternative to the hydrostatic / pressure test requirement of section IWD-5223 ( f) , CEI proposes to perform the hydrostatic / pressure test requirements of section IWD-5223(d). Confirmation of adequate flow in accordance with section IWD-5223(d) will satisfy the inspection requirements of: extent of examination (pressure retaining material), i examination method (visual, VT-2), and frequency of examination (each I inspection interval).

l

1 l

I Attachment 2

! PY-CEl/NRR-2310L i

1 Page 82 of 85 )

Sheet 1 of 4 I

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-008, Rev. 0 l

I I. Identification of Components All Class 1, 2, and 3 pressure retaining bolted connections that receive system pressure tests in accordance with IWA-5000, IWB-5000, IWC-5000 and IWD-5000. l II. ASME B&PV Section XI Requirements i ASME XI, 1989 Edition, IWA-5250 (a) (2) , Corrective Measures, requires that when leakage at a bolted connection during the conduct of a system pressure i j test, the bolting shall be removed, "T-3 visually examined for corrosion, j and evaluated in accordance with IWA-3100.

III. Relief Requested I i

Pursuant to 10 CFR 50.55a (a) (3) (i), relief is requested from the removal of bolting and performance of a VT-3 visual examination for corrosion, when )

leakage at a bolted connection is identified within a system that does not contain a fluid that would likely cause a corrosion-related failure. The j relief is requested for Perry's second 10-year inspection interval. l IV. Basis for Relief j IWA-5250 (a) (2) states that if a leak is identified at a bolted connection the l bolting shall be removed and VT-3 visually examined for corrosion. Paragraph  ;

IWA-5250 (a ) (2) was not in the 1983 Edition, Summer 1983 Addenda of ASME XI; i which was the Code of record for Perry's first 10-year inspection interval.

It was incorporated into ASME XI in the Winter 1984 Addenda.

i During the first inspection interval, and in the absence of specific ASME XI requirements, Perry established acceptance criteria and procedures to address leakage at bolted connections and incorporated them into Perry's Inservice Pressure Testing Program. They required that all leakage from bolted connections, both pressure retaining bolting and non-pressure retaining bolting (e.g., packing gland bolting), be recorded during the conduct of system pressure tests. Leakage in excess of 5 drops per minute required the initiation of corrective action. As a minimum, the corrective action would be the initiation of a work request. In accordance with Perry's Work Prioritization System, work requests for correcting a condition that has resulted in or has a high potential to result in spills, leaks, or releases of radioactive or hazardous substances are given a high priority. Correction was typically accomplished by re-torquing the bolted connection. There were no cases in which leakage from bolted connections was found to be caused by a corrosive failure of the bolting or where the leaking connections were found to have bolting that was significantly degraded due to corrosion.

I Attachment 2 {

PY-CEI/NRR-2310L l Page 83 of 85 l Sheet 2 of 4 l

Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-008, Rev. 0 l The IWA-5250 (a) (2) requirement implies that a system must contain a fluid, chemical additive or impurity that will induce a corrosive environment upon leakage from a bolted connection and could cause a corrosive failure of the bolting (i.e., loss of structural integrity). Furthermore, the summary of Subcommittee XI action item ISI 80-13, which was the action that incorporated the new IWA-5250 (a) (2) requirement into ASME XI, stated, " Considerable experience dictates that more stringent requirements must be imposed for the examination of systems containing borated water." In other words, the impetus for the new requirement was to address degradation or failure of bolting due to the chemical addition of boric acid.

A review of Class 1, 2, and 3 systems was performed to evaluate which could contain unwanted impurities, process fluids, natural impurities, or chemical I additives that would be likely to create a corrosive failure. First, the  !

likelihood of unwanted impurities was considered. Perry's programs for Chemical Protection, Materials Control, Housekeeping / Cleanliness Control, i Foreign Material Exclusion, and Plant Chemistry Control preclude inadvertent transportation or intrusion of unwanted impurities (chemicals) in Class 1, 2 and 3 systems. Then, the types of process fluids and their natural impurities were considered. The process fluids within the Class 1, 2 and 3 systems are water, steam, fuel oil, lube oil, air and hydrogen. By themselves, these fluids do not have a chemical composition or contain ,

natural impurities that are likely to cause corrosive failures. Finally, the chemical additives to the Class 1, 2, and 3 systems were reviewed. The i safety class systems with chemical additives are as follows:  !

• Standby Liquid Control (lC41) System uses boric acid and sodium tetraborate to form a sodium pentaborate solution for controlling reactivity

• Feedwater (N27) System uses depleted zinc for oxide layer / dose control

• Emergency Closed Cooling (P42), Nuclear Closed Cooling (lP43), and Control Complex Chilled Water (1P47) Systems use ammonium hydroxide for pH control and hydrazine for control of dissolved oxygen; nitric acid is also may used to reduce pH

• Emergency Service Water (P45) System uses sodium hypochlorate as a biological growth inhibitor

• Diesel Fuel Oil and Transfer (lR44/1E22B) System uses the blocide BIOBOR JR as a biological growth inhibitor

• Diesel Generator Jacket Water (1R46/1E228) System uses Nalcool 2000 (an aqueous solution of borates, silicates, nitrates, nitrites, and mercaptobensothiazole) as a corrosion inhibitor and Nalcool 7330 (isothiazolone) to kill bacteria and inhibit microbiological fouling

• Diesel Lube Oil (1R47) System uses Dow Corning 200 as an antifoaming agent 1

Attachment 2 PY-CEl/NRR-2310L .

Page 84 of 85 J

Sheet'3 of'4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-008, Rev. 0

'Of'the above additives, only boric acid is considered likely to cause corrosive degradation.or' failures of pressure retaining bolted connections (although the sodium pentaborate solution formed by the combination of boric acid and sodium tetraborate results in a pH value of.7.9 within the 1C41 syctem and is not likely to cause corrosive degradation). Thus, an equivalent measure of safety can be obtained by following the requirements of IWA-5250(a) (2) for borated systems only.

ASME Subcommittee XI recently approved Intent Inquiry IN97-33; which addresses the intent ' of paragraph IWA-5250(a) (2) . The inquiry states, "Is l- it the intent of IWA-5250 (a) (2) that bolting be removed and a VT-3 visual examination be' performed only when the system is borated for the purposes of l controlling reactivity?". The reply was, "Yes." Following ASME Code Committee policy, an attending Code action (ISI 97-55) was passed (by Subcommittee XI and Main Committee) to revise paragraph IWA-5250 (a) (2) to state,-"If leakage occurs at a bolted connection in a system borated for the l purpose of controlling reactivity, one of the bolts shall be removed, VT-3 L examined, and. evaluated in accordance with IWA-3100." .These Code actions ji rainforce the conclusion of Perry's review of its safety related systems for

.likely susceptibility to corrosive failure at pressure retaining bolted connections.

In some instances, compliance with IWA-5250(a) (2) as written in the 1989-Edition would cause extreme hardship. For example, leakage from the bolted flange connections of Control Rod Drive (CRD) Housings is a common occurrence in BWRs. 'The leakage is identified during the conduct of the Class 1 System Leakage Tests performed by all BWRs at the end of each refueling outage. The I

tests are performed at the normal operating pressure, but at lower than operating temperatures. General Electric has informed BWR owners that, i n

most instances, the leakage. stops within eight hours of being pressurized to 1000 pounds at operating: temperatures. This has been observed at Perry. In every outage,. leaking CRD flanges are identified and as the plant starts up

.and.comes to full operating pressure and temperature the in-leakage to the drywell' floor drain sump goes down. Also, entries into the drywell have been made at 15% power and it was noted that the leakage had stopped. In accordance with Category B-G-2, the bolting of all disassembled CRD housings-is VT-1 examined. In RF04, Perry replaced a minimum of 4 out of 8 of the bolts in every CRD flange connection with an new GE design with higher strength and' slotted washers to. facilitate drainage. Of the bolts that were replaced, none of them were found.to be corroded. If the bolting of these CRD flanges - that are known to leak, but have never exhibited signs of corrosive damage - had to be disassembled to perform VT-3 examinations, the overall duration of every refeuling' outage could be extended by a day or more.

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Attachment 2 PY-CEI/NRR-2310L Page 85 of 85 Sheet 4 of 4 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST #PT-008, Rev. O V. Alternate Examination In lieu of removing bolting and performing a VT-3 examination for all pressure retaining bolted connections found to be leaking during the conduct of a system pressure test, only the bolting of borated systems will be removed and VT-3 examined. For other than borated systems, the acceptance criteria and corrective action procedures of Perry's Inservice Pressure

! Testing Program will be applied.

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