ML20082H637
| ML20082H637 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 08/01/1991 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20082H635 | List: |
| References | |
| PROC-910801, NUDOCS 9108260197 | |
| Download: ML20082H637 (136) | |
Text
{{#Wiki_filter:. _ _ - _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ _ _ - - _ - _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ Attachment to NIS9100$27 dahd August 14, 19l August 1991 Addenda Date of Issue: August 1, 1991 O NEBRASPA PUBLIC POWER DISTRICT COOPER NUCIIAR STATION INSERVICE IhSPECTION PROGRAM POR ASME CLASS 1, 2. AND 3 COMPONENTS Revision 3 l This is an addenda to the loose leaf version of the Inservice Inspection Program for ASME Class 1, 2, and 3 Components, Revision 3, and is issued in the form of replacement or - cdditional pages, revisions, additions, or deletions, and are incorporated directly into t.he l affected pages. , I l S.ummary of_Chancesi I This is the eighth addenda to be published to the Inservice Inspection Program for ASME class 1, 2, and 3 compononts, Revision 3. This change affects only the Augmented ISI section of the CNS ISI Program. Changes given below are identified on pages 1 and 2 at the beginning of the section by a margin note (August 91) next to the affected item, i SECTION PAGE(s) DESCRIPTION _ 1 Augmented Inservice Inspections 1 and 2 Delete existing pages and replace with attached new pages. Updated the AISI listings for Tabs 5 and 8 and added Taba 11, 12, 13 and 14. b v Augmented Inservice Inspections Tab 5 all pages Delete existing paBes and replace with attached new pages. Updated CNS GL88 01 ISI requirements to reflect changes to RWCU system.
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Augmented Inservice Inspectionu Tab 8 - all pages Delete existing pnSes and replace with attached new pages. . Updated CNS shroud access cover inspection requirements to meet GE SIL No. 462S2R1. Augmented Inservice Inspections Tab 11 - all Add this now tah and attached' pages pages. Added Type 11 AISI, SRM and IRM dry tube inspections per CE SIL No. 409R1, Augmented Inservice Inspections Tab 12 - all Add this new tab and attached pages pages. Added Type 12 AISI, shroud head bolt inspections por GE SIL No. 433. Augmented Inservice Inspections Tab 13 all Add this new tab and attached pages pages. Documents rationale for updating RPV UT exam procedures per Reg. Guide 1.150 as recommended by GE SIL No. 515. Augmented Inservice Inspections Tab 14 all Add this new tab and attached O' pages page. Add new Volume 3 to ISI Program. Augment ISI Program enhancement as a result of U.S. NRC Inspection Report 90 15. 9108260197 910811 8 E.R ADOCK 0500
1 1 1 N " AEGJiG12.lhgERVICE Inrignigi
- O Augstuted Inservice Inspections (A181) are not ASME Section XI Code j requirements, but are 1) additional examinations areas or 2) increased inspection frequency or combinations of both which are requested by tha Nuclear Regulatory Commission, recouunended in General 01cetric Company Service Inforination Letters or added for other reasons.
When exarnination cotoponents fall into the scheduled testing of IS1 and are also AISI requirements, then credit for both requirements inay be taken (no double testing). The following are types of Augtnented Inservice Inspections requ.t red at July 89 Cooper Nuclear Station. The TAB number corresponds to tabbed pages that follow which contain information on the specific type of AISI. TAB TYPE DESCRIPTION REVISION DATE m- -_ 1 1 All ring girder bolting and ring girder anchor Original bolting is to be volutnetrienlly inspected each ten Release 3/85 ; year interval. The anchor bolting adjacent to the inboard MSIV is to bo visually inspected each ten year interval. (
Reference:
NRC DRO Bulletin #74-i -- 3 L_ 2 2 Ultrasente examination of the feedwater nozzle safe original ends, beres, and inside blend radii, liquid Release - 3/85 L penetrant exatnination of the feedwater nozzles, and visual inspection of the feedwater spargers as
} required per Table 2 and Section 4.3.2.4 of NOREG.
; 0619, 3 > Visual inspection of the Core Spray spargers and Original the Core Spray piping inside the RPV shall be Release - 3/85 conducted each rofueling outage. (
Reference:
IE Bulletin No. 80 13.) 4 4 Ultrasonic examinations, utilizing G.E. Procedure Original TP508.0654, Revision D, or equivalent, are Release - 3/85 conducted to assess the integrity of the jet pump hold down bea.as at the mid length ligament areas bounding the bearn bolt. These examinations shall be performed once during the second ten year interval. These examinations may be deferred to the end of the interval . 5 5 Ultrasonic 'exarninations per Generic Letter (G.L.) Aug. 91 88 01, 0.L. 88 01 applies to all BVR piping taade of austenitic stainless steel that is four inches or larger in nominal dinineter, and contains reactor coolant at above 200*F during power operation regardless of code classification. All accessible welds will be exarnined in accordance with CNS G.L.
,88 01 commitments.
O Page 1 of 2
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O TAB TYPE _ . _ . _ DESCRIPTION
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REVISION DATE Feb. 89 6 6 Visual inspection of stearn dryer channel welds during refueling outages (Reference concral E1cetric SIL No. 474.) 7 7 Visual inspection of jet pump nozzles and inixer July 89 inlets in conjunc* ton with jet purep inspection. (
Reference:
General E1cetric SIL No. 465.) 8 8 Ultrasonic exatnination of the shroud support access Aug. 91 hole covers once every three years beginning with the Spring 1993 Refueling OutaBe, Visuni exarnination VT 1 of the shroud support access hole , j covers during the 1991 Retuoling Outage, (
Reference:
General Electric SIL No. 462, S2, R1, and G.E. :nemo RCll9143, dated April 4,1991.) 9 9 Visual inspection of the Core Spray T junction box July 89 welds inside the reactor vessel. (
Reference:
General Electric SIL No. 289, R1, St.) 10 10 visual examination of the Reactor Recirculation July 89 (RR) pumps' shaf ts, pump covers, inopeller/shaf t attachment region (includin6 bolts), and hydrostatic bearings (including baffic plate). : Frequency of examination sna11 coincide wit.h the regularly scheduled RR pumps inspection. O (
Reference:
General Electric SIL No. 459 and RICSIL No. 038.) 11 Visual exarnination of all accessible areas of the Aug. 91 11 Interinediate Range Monitor (IRH) and Source pange Monitor (SRM) dry tubes each refueling outage. (
Reference:
General Electric SIL No. 409 R1.) Ultrasonic (UT) exarnination of all remaining old Aug. 91 12 12 design creviced Inconel 600 Shroud licad Bolts (SilBs) each refueling outage. (
Reference:
Cencral Electric SIL No. 433.) Rationale for updating of General Electric reactor Aug. 91 13 13 pressure vessel ultrasonic inspection procedures to incorporate U.S. NRC Regulatory Guide 1.150 require ments. (
Reference:
General Elcet.ric SIL No. 515.) 14 Augmented Inservice Inspection (AISI) Pro 5 ram for Aug. 91 14 Service Water (SW) and Reactor Equipment Cooling (REC) pipe supports outside the scope of the CNS ASME Section XI ISI Prograin. This AISI Pro 6 ram requires only VT-3 and/or VT 4 exarninations of selected supports. I OlBBWW_ O Page 2 of 2
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l a-i nt, 3 (Revised Text)
-COOPgR NUCLEAR STATION 181 PROGRAH f
O Augmented Insatvice Inspection Progra's i in Accordance with U.S. NRC Ge.teric Letter 88 01 i
References:
- 1. NRC position on ICSCC in BWR austenitic stainicas steel piping (Generic Letter 88 01), ;
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- 2. Letter, C. A. Trevors (NPPD) to U.S. NRC, dated l October 9, 1990.
Subject:
Gencr!c letter 88 01, Cooper , Nuclear Station. .
- 3. Letter, P. W. O'Connor to C. R. llorn (NPPD), d.ned f May 24, 1991.
Subject:
Reactor Water Cleanup (RWCU) Pipo Weld Inspection. As a renuit of previous cormitments to Reference 1, NPPD replaced the majority of the Category D Intergranular Stress Corrosion Cracking (ICSCC) susceptible piping and welds with Category A'IGSCC resistant piping and wolds during the 1990 i Refueling Outage. The only Category D piping and welds that were not replaced consists of the RWCU return piping frorn the regenerative heat exchanger outlet to the RWCU/RCIC attachment to the feedwater inlet line. In heference 2, NPPD proposed to revise its previous cormitment to replace this piping and instead conduct continued inspection of this piping in accordance with the requirements of CL 88 01. This request was approved and future inspection requirements for all RWCU piping veld inspections were documented in keierenco 3. The following pages document the CNS Augmented ISI require:sents for all of the applicable RWCU piping welds in accordance with Referena 1, All irmcceasible Category D welds have been reclassified as Category C vQds in accordance with CL 88 01. Also, weld RCA 11F 1 (CRD nozzle cap weld) har been reclassified as Category D and included in the attached prograrn. NPPD will conform to the NRC staff position on reporting uquirements as stated in CL 88 01. The NRC will be notified of any flaws identifhd that do not meet IWB.3500 criteria froin Section XI of the Code for continued operation without evaluation or a change found in condition of welds previously known _ to be cracked. The NRC will be notified of any flaw evaluation rhquired for continued operation and/or flaw repair plans. O.
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- O- O O I-GENERIC LEITER 88-01 ADQENTED 151 EIAMINATIONS CATEG0KY A RUCU PIPE WEIDS WELD ID CONF. SIZE ISO- ELEV. MAT. UT PROC. INTERVAL ,.f.A UZARK5 RUCU-82 P-V 4* 2605-2A 936'2" P-21/F-27 No. 110 18 IGSCC Cat. A per C.L. 88-01 RWCU-83 'V-T 4* 2605-2A 936'2" F-27/F-26 No. 110 18 -
ICSCC Cat. A per G.L. '88-01 RWCU-64 T-P 4* 2605-2A 936'2' P-21/F-26 No. 110 18 ; ICSCC Cat. A per G.L. 88-01 (
;, RWCU-85 P-E- 4* 2605-2A 940'6" P-21/F-26 No. 110 18 IGSCC Cat. A per G.L. 88-01 l IGSCC Cat. A per G.L. 88-01 i RWGU-86 E-P 4* 2605-2A 940'6" P-21/F-26 No. 110 18 RWCU-87 P-N 4* 2605-2A 937'5" P-21/(1) No. 110 18 ICSCC Cat. A per C.L. 88-01 4
RWCU-88 V-T 4* 2605-2A' 935'10" F-26/F-27 No. 110 18 IGSCC Cat. A per G.L. 88 i RUCU-89 P-V 4* 2605-2A' 934*6* F-27/P-21' No. 110 18 IGSCC Cat. A per G.L 88-01 l 'RWCU-90 E-P 4" 2605-2A 932'6" P-21/F-26' No. 110 18 IGSCC Cat. A per G.L. 88-01 i ! RWCU-91 P-E 4* 2605-2A 932'6" P-21/F-26 No. 110 18 ICSCC Cat. A per G.L. 38-01 1 I RWCU E-P 4" 2605-2A 932*6* P-21/F-26 No. 110 18 2 3/F94 ICSCC Cat. A per G.L. 88-01 i o I RUCU-93 N-E 4" 2605-2A 933'6" F-26/(1) No. 110 18 IGSCC Cat. A per C.L. 68-013 - RWCU-94 T-P 4* 2605-4A 946'0" F-27/P-21 No. 110 18 IGSCC Cat. A per G.L. BS-01 RWCU-95' P-CAP 4* 2605-4A 946'0* P-21/F26 No. 110 18 ICSCC Cat. A per C.L. 88-01 i' CUA-CF-45 V-P 6* 2605-4 946'0" (2)/P-21 No. 48 18 ICSCC Cat. A per C.L. 88 CWA-CF-46' P-T 6* 2605-4 946'0* P-21/F-26 No. 48 18 ICSCC Car. A per G.L. 58-01 i CWA-CF-47 T-P 6* 2605-4 946'0" F-26/P-21 No. 48 18 IGSCC Cat. A per G.L. 56-01' , i ,
- CWA-CF-48 P-E- 6* 2605-4 946'0* P-21/F-26 No. 48 18 2 2/F91 ICSCC Cat. A per G.L. 38-01 !
CWA-CF-50 E-P 6* 2605-4 946'0* F-26/P-21 No. 48 18 ICSCC Cat. A per C.L. 83 01 i ! Page 2 of 3 i
[: + , 1 \ i GENERIC LEITER 88-01' AUQtENIED ISI EIlJtINATIONS CATEC0KY A RUCU PIPE WEIDS
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i i
- J PERIOD / REMARKS.
LTELD ID SIZE ISO - EI1V. MAT. UT PROC. INTERVAL ! CONF. CNSCAL}- STD. NO OUTAGE l. . CUA-CF-51 P-E ' 6" 2605-4 940'0* P-21/F-'26 ~ No. 48' 18 ICSCC Cat. A per G.L. 88-01 CVA-CF 52 E-P 6" 2605-4 940'0* F-26/P-21 No. 48 18 IGSCC Cat. A per G.L. 28-01' . i CUA-CF-54 P-E 6" 2605-4 940'0 - P21/F-26 No. 48 18 '2 3/S93 ICSCC Cat. A per G.L. 88-01 i CWA-CF-56 E-T '6" 2605-4 940'0" F-26/F-26 No. 48 18 IGSCC Cat. A per G.L. 88'-01 [ CWA-CF-57' T-R- 6' 2605-4 938'10" F-26/F-26 No. 48 18 IGSCC Cat. A per G.L. 88 31 CWA-CF-58 T-R 6" 2605-4 938'10" F-26/F-26; No. 48 18 IGSCC Cat. A per G.L. 88-01 l _
- NOTES: ~(1) Material - ASTM A351 Gr. CF3.
(2) RWCU-M07-M018 Material - ASTM A351 Gr. CF8M. I 1 i l ( -- l 1 k J'I s i Page 3 of 3 _- - . . , _ . _ _ , . . . _ . - - _ . - , , _ - _ _ . . . , , , . . _ , _ . _ _ _ _ . . , _ . - . _ , , . . , _ . ~ . - . - _ _ . . , _ . . . - . _ _ . . , _ _ . . _ . . _ _ _ - . . . _ . - _ . - _ . ~__..__....!
O O CENERIC LETTER 88-C1 AUGMENTED ISI EXAMINATIONS O CATECORY D AND G RWCU PIPE WEIES INTERVAL ! REMARICS WELD ID CONF. SIZE ISO ELEV. MAT. pg , GE RWCU-10 N-P 4" NPS 2605-3 940' P-12 10 18 2 3/F94 IGSCC Cat. D per CL E8-01 RWCU-11 P-E 4" NFS 2605-3 948' P-12 10 18 2 3/S93 IGSCC Cat. D per GL 88-01 RUCU-12 E-P 4* NPS 2605-3 948' P-12 10 18 2 3/S93 IGSCC Cat. D per CL 88-01 RWCU-13 P-E 4" NPS 2605-3 948' P-12 10 18 2 (1) IGSCC Cat. D per GL 88-01 RWCU-14 E-P 4" NPS 2605-3 948' P-12 10 18 2 2/F91 ICSCC Cat. D per GL 88-01 RUCU-15 P-P 4" NPS 2605-3 948' P-12 10 18 (2) IGSCC Cat. G per GL 89-01 j RWCU-16 P-E 4" NPS 2605-3 948' P-12 10 18 (2) IGSCC Cat. G per CL 88-01 RWCU-17 E-P 4" NPS 2605-3 948' P-12 10 18 (2) IGSCC Cat. G per GL 63-01 RUCU-18 P-E 4" NPS 2605-3 948' P-12 10 18 (2) ICSCC Cat. G per GL 88-01 RWCU-19 E-P 4" NPS 2605-3 947'6" P-12 10 18 (2) IGSCC Cat. G per GL 88-01 RWCU-20 P-E 4" NPS 2605-3 947' P-12 to 18 (2) IGSCC Cat. G per GL 88-01 RWCU-21 E-P 4" NPS 2605-3 946' P-12 10 18 (2) IGSCC Cat. G per GL 88-01 4 RWCU-22 P-E 4" NPS 2605-3 946' P-12 10 18 (2) IGSCC Cat. C per GL ES-01 ! RWCU-23 E-P 4' NPS 2605-3 Above 931' P-12 10 18 (2) IGSCC Cat. G per GL 68-01 RUCU-24 P-P 4" NPS 2605-3 Below 931' P-12 10 18 (2) IGSCC Cat. G per GL 88-01 i RUCU-25 P-P 04" NPS 2605-3 Above 903' P-12 10 18 IGSCC Cat. D per GL 88-01 l RWCU-26 P-E 4" NPS 2605-3 899'9" P-12 10 18 2 (1) IGSCC Cat. D per GL 88-01 I RWCU-27 E-P 4" NPS 2605-1 899'3" P-12 10 18 2 2/F91 ICSCC Cat. D per GL 88-01 RWCU-28 P-E : 4" NPS 2605 899'3" P-12 10 18 2 3/F94 IGSCC Cat. D per GL 88-01 l l RWCU-29 E-P 4" NPS 2605 899'3" P-12 10 18 2 3/F94 IGSCC Cat. D per GL 88-01 , RWCU-29A P-T 4" NPS 2605-1 899'3" P-12 10 18 2 3/F94 ICSCC Cat. D per GL 88-01 RWCU-30 T-P 4" NPS 2605-1 899'3" P-12 10 18 ICSCC Cat. D per GL 88-01 l Page 1 of 3
0 % CENERIC LeLAra 88-C1 AUGMENTED ~ E1 NATIONS CATEGORY D AND C RWCU PIPE WEIRS WF.LD ID CONF. SIZE ISO ELEV. ?!AT. INTERVAL RE KS
. . FROC. 0 A RWCU-31 P-E 4* NPS 2605-1 899'3" P-12 10 18 ICSCC Cat. D per GL 68-01 RWCU-32 E-P 4" NPS 2605-1 898' P-12 10 18 ICSCO Cat. D per CL 88-01 RWCU-33 P-E 4* NPS 2605-1 894' P-12 10 18 IGSCC Cat. D per GL 88-01 RWCU-34 E-P 4* NPS 2605-1 893'4" 1 P-12 10 18 IGSCC Cat. D per GL 88-01 RECU-35 P-E 4" NPS 2605-1 893'4" P-12 10 18 ICSCC Cat. D per CL 88-01 I RWCU-36 E-P 4" NPS 2605-1 893'4" P-12 10 18 3/S93 I IGSCC Cat. D per CL 88-01 RWCU-37 P-E 4" NPS 2605-1 893'4" P-12 10 18 3/593 ICSCC Cat. D per GL 88-01 IGSCC Cat. D per GL 88-01 RWCU-38 E-P 4* NPS 2605-1 893'4" P-12 10 18 RWCU-39 P-E 4* NPS 2605-1 893'4" P-12 10 18 ,
ICSCC Cat. D per GL 88-01 RWCU-40 E-P 4" NPS 2605-1 893'4" P-12 10 18 IGSCC Cat. D per GL 88-01 RWCU-41 P-E 4' NPS 2605-1 893'4" P-12 10 18 2 3/F94 IGSCC Cat. D per GL 88-01 RWCU-42 E-P 4* NPS 2605-1 894' P-12 10 18 2 3/F94 IGSCC Cat. D per GL 88-01 3
. RWCU-43 P-P 4* .4PS 2605-1 902'6" P-12 10 18 IGSCC Cat. P per GL 88-01 RWCU-44 P-E 4" NPS 2605-1 903' P-12 10 18 IGSCC Cat. D per GL 88-01 i RWCU-45 E-P l 4" NPS' 2605 904*6* P-12 10 18 2 l 2/F91 l IGSCC Cat. D per GL 88-01 1 RWCU-46 P-E 4" NPS 2605-1 904*6* P-12 10 18 2 2/F91 IGSCC Cat. D per GL 88-01 1 .
l RWCU-47 E-V 4* NPS 2605-1 904*6" P-12 10 18 2 l 2/F91 IGSCC Cat. D per GL 88-01 l RUCU-48 V-P 4" NPS 2513-1 904'6* P-12 10 18 2 3/593 IGSCC Cat. D per CL 88-01 I RWCU-49 P-P 4" NPS 2513-1 904'6* P-12 10 18 2 3/593 IGSCC Cat. D per GL SS-01 i CW3-BF-8 P-V 4* NPS 2513-1 904'6" P-12/CS 10 18 ICSCC Cat. D per GL 88-01 RCA-BF-1 C-N 4* NPS (4) (4) (3) 20 6 2 3/F94 IGSCC Cat. D per GL 88-01 j l j Page 2 of 3
O O CENTRIC IEITER 88-Cl ADCMENTED ISI EXAMINATIONS O CATECORY D AND C REGU PIPE WEIDS NOTES: (1) This veld failed UT exa:mination during the Spring 1989 Cutage. Weld was repaired and passed second UT examination; will be re-examined every two refueling outages starting with the 1991 Fall Refueling Outage. . (2) These velds are considered inaccessible. (3) Weld RCA-BF-1 is the CRD nor:le cap veld; the materials are no le-carbon steel, cap-NiGrFe, and Inconel 182 and Inconel 82 veld filler metal. (4) The CRD return line nor:le is located on the reactor pressure vessel adjacent to feedvater no::le N45. 1 4 i j . t l ! i i Page 3 of 3
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OENERAL El.ECTRIC COMPANY SIL KO. 462, SUPPL.EMENT 2. REV.tSlaN 1 SilROUD SUPPORT ACCESS Il01.E COVER CRACKS REVISED CNS ENGINEERING RI'SPONSE TO RECOMMENDATIch No. 1 [ l EUJ199ht1%VMID1LLtoa 1 G.E. Nucicar Energy recommetvis that owners of all G.E. BWRs that have creviced, Alloy 600 access hole covoro which havo not yet ban exarnined ultaasonically do so during the next outage. Re exatninationo nhould be perforuod at least once every three years. At G.E. lWRs operating on two year cycles, the to-examina. tions should be performed during overy refuelin6 outago. If ultrasonic oxaminations of shroud head bolts reveal live or anoro bolts with ne,w cracks, access hole covers should be exs.tnined teoro f requent.ly than once overy three years. Access hole covers which aircady havo been exatnined with no crackin6 found also should be re examined every throa years irotn the date of the first. exatnination, At 0.E. INRs operating on two year cycles, the re examinations should be performed ovary refueling outage. Owners of 0.E. INRs in this category which are oporatinty, with hydrogen water chemistry taay justify longer inspection intervals consistent with the plant inservice inspection prograts following the first access hole cover exatnination9. CNS RESPONSE (Revised) The augtmented section of the CNS ISI Prograta will be revised to requiro UT of the access cover to shroud support veld (including base metal areas adjacent to the weld) every tbreo years. CNS performed a UT exatnination of those areas in April, 1989 with no reportable indications, and the next examinat lon would be duo in April, 1992. Based in part upon recommendations in memo RCil 9143, R. C. Ilooper (0.E.) to G. E. lilcks_, Jr. (NPPD), dat ed Apr.114,1991 CNS will perfortn the next UT examination of the shroud access cover-to ohroud support. wold and adjdicotit base metal areas during the Spring 1993 Refueling Outage. A visual exainination of this weld and adjacent base inctal areas will be performed during the Fall 1991 Refueling Outage. In addition, a requirement will be added to the Augmented Section of the CNS ISI Prograin that if five or more shroud head bolts are found to havo UT crack indications, a UT exainination of the access cover to shroud wold areas will be immediately performed. CNS performs UT examinations of all shroud head bolts each outsgo. All other CNS Engineering ror;ponses t.o G.E. SIL No. 462, Supplement 2, Rovi-sion 1, retnain unchanged. , T f o 1 G/k7/i( G. E. Ilicksl, J r.
#hM.
'J'. dp[encerb c4 t_.<..tv t
! IS1 Engineer Engincoring Programs Supervisor O l 2- - - . .
GE Nacicar Enotay In 'MiZ10!'[.L. u mn b) Decemter 31,1990 Gl! Nuclear F,nergy has issued the Sllidentined below, a copy of which is enclosed with this letter for your information. ; SIL No. 462 Supplomont 2 Revision 1 Shroud Support Access Holo Cover Cracks
'the applicat.llity desi;: cations t< low are based on infonnation availabic to Gli Nuclear !!nergy at the titue this SIL wns issued. 'lhe applicability designations are explained as follows:
X Oli belleves this SIL applies to this IlWit. I Gli was unable to assess applicability of this SIL to this llWit. 'Ihls SIL may or may not apply to this llWit O Gli belleves this SIL does not apply to this llWit. Please read the Notice contained in this SIL concverning determination of applicability and implementation of recommended action,if any, if you have questions about this Sll or if you wish to change the address to which Gli distsibutes SILs and it!CSILs to you, please contact your local Gli Nuclear Services hianager. If you wish to infonn Gliof the 51atus of f mplementing this or any other SIL at your llWit, please complete form S im (Service Information letter Status itesponse) at:d snail it to J. G. hiocre hi/C 385, Gli Nuclear linergy. San Jose, CA 95125. Plant Name Applicability Plant Name Applicability liig Itock Point O Laguna Verde 1 & 2 X
- Ilrowns Feny 1,2 & 3 X LaSahe 1 & 2 X lhunswick 1 & 2 X leibstadt X Caon..o. . . _ . _
X Limerick 1 & 2 X Clinton i X hionticello X CNV I & 2 X Nine hille Point 1 0 Cofrentes X Nine hille Point 2 X Cooper X Oyster Creek O Dresden 2 & 3 m X Peach llottom 2 &. 3 O Duane Arnold X Perry 1 X Fermi 2 X Pilgrim 1 X hti. Patrick X Quad Cities I & 2 X Fukushima i X ltiver llend X Fukushima 2 X Santa h1 aria de Garona X Fukeind X Shoreham X GKN Dalewaard O Susquehanna 1 & 2 X Grand Gulf X Tarapur 1 & 2 O l Ilatch I & 2 .;' Tokai2 X l Ilope Cicek J Tsuruga O KKh1 X Vennont Yankee X l Kuo Sheng 1 & 2 X WNP-2 X bl kI \ J G h!oore, Customer Service Communications hianager l
/. . ' .
- _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _SIL _.___._-.--.-
NJ December 19,1990 Sil, No. 462 Supplement 2 Ites hinn 1 Categosy 1 l Shroud Support Access Hole Cracks In Febniary 1988, SIL No, 462 r eported that cracking cracking had an average depth of about 45% of the had been detected in the shroud support access hole thickness for die temninder of the cover. 'ihis crack-covers at a GilllWlu4 and of fer ed speci fic inspection ing is very similar to an:llocated in the same area as recommendations. In February 1989, Supplement 1 the cracking that was detected at another GlillWlU4 to SIL No. 462 provided more information on the and reported in SIL 462 and Sil,462 Supplement 1. location and the extent of the original encking fol- 1hc access hole cover design also was similar to that lowing a reinspeedon at that GliIlWlU4 using im- in which the previously reported cracking ocemted. proved equipment and procedures. Supplement I also contained inspection recommendations. Fol- As 1,hown in Figme 1, the cracking initiated in die lowing successful inspections at a number of GI! vertical uevice and pmparated up the vertical weld llWRs, SIL No. 462 Supplement 2, issued in August fusion line pn the f.hroud support ledge side of the 1990, stemphasized the inspection recommenda- weld. 'lhe cracking is believed to be crevice assisted tions and rec nnmended a ten year hequency for intergranular stress cornwlon aucking (IGSCC). *lhls
/O reexaminations, cracking initiates in a cmvice on the underside of the d support led;.r,e and could initiate at several h> cations
'Ihis llevision 1 to SIL 462 Supplement 2 discusses amund the circumference of the weld at the same a new occurrence of circumferential cracking ob- time. 'lhe uncks cannot he seen by visualirspections served recently at a Gl!IlWlU4 located outAlde of the until they have propagated through the wall.
United States. Cracking was ob- 's - o, ., served in both of the access hole
' [ '.N.' W$, h.N N9NO1, i
awer to shroud sugwit kdge v. cit during a roudne visual examina-tion and has been confirmed by an y 9 -CrackI cations g , ultrasonic (UT) examination to be condnuous over the entlic circum- Access llole Cover ference of both covers. This Itcv0 Shroud 1.cdge M sinn 1 'voics Sil,ivu. *ut aupple-ment 2.
, , w , % Discussion * < ' 7 W If access hole cover cracking is 1cycre, complete separution of a covers fmm the shroud ledge could
'lhrough wall cracking adjacent to the access hole occur during plant operation. Although severe crack-cover was detected over about 29% of the circumfer- ing is not a safety concern, separation would result in ence ofone cover. 'Ihe cracking had an average depth damage to internals and cause a forced outage. The of about $5% of the cover thickness for the remainder results provided in SIL No. 462 of the safety evalu-of the circumference, 'Ihrough wall cracking also ation of a Gl!IlWlU4 showed that access hole cover was detected adjacent to the opposite access hole cracking does not represent a safety concern because 9 cover over about 17% of its circumference. 'lhis of the margins inherent in the design and the case in pago1
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, +
1 i Q b detecting a discarpancy between indicated cote flow and indicated core power. No. 462 also apply to thch plants,
- 3. Prepale a cofitingenCy sepalt plograin to suppott I Original design slu oud head bolts in inany Gl! !! Wits access hole covci seplacernent ut the next sched. I are tuade of thuaine ruaterial t s the shroud ledge and uled outage if c.vcks are detected.
are similarly rieviced. Although the water environ-ment above the corc is morc oxidid ng then that in the 4. Until htitiat inspections have been made. conduct plenutn, shroud head bolt cracking can be used as a periodic operational checks to ldentify evhlence predictor of potential cracking in access hole covers. of byle flow through cadni sms tole covers.
. .. , .,. f- ; 'Ihc following operating guideline is suggested Reconinicnded Action for,GE UWRs whh Ior perfunning the pcshni checks.
, CrevicedfIloy s 600 Ac' cias "'Holo Covers -
~
Basin of Sugges'Ich Rcactor' Operating
- 1. Gl! Nuclear linergy recommends that owners of (Guldclino' @
all Gl! IlWits that have creviced. Alloy 600 access hole covers which have not yet been exam- 'Through-wall cracking in the heat affected rone of ined ultrasonically do so during the next outage. the shroud support access hole cover attachment wchl iteexaminations should be performed at least couhl provide a flow path which partially bypaues once every three years. At Gl!ItWits operating the reactor core. Although the reduced recirculation on two year cycles, the r eciarninations shouhl be system hydraulic iesistance can allow the total tecn-performed during every refueling outage. culation flow toincicase, a pottion of the flow would bypass the core. 'Ihus, the actual core flow wouhl If ultrasonic examinations of shroud head bolts decrea',e and lead to a reduction in core power covis-reveal flve or more bolts with new cracks, access tent with the normal power flow inap. hole covers should be examined more frequently ( than once every three years, llecause most access holes are approximately 19 inches in diameter, the cf fcct of sue.len separation of Access hole covers which already have been ex- the covers on secirculation flow and core power amined with no cncking found also should le would be significant. If severe enough the efIcci on reexamined every threc ycars from the datc of the core flow couhl be sufficient to cause the plant to first examination. At GlillWits otrrang on enter regions of potential core thennal. hydraulic two year cycles, the tecxaminations shouhl be instability (please refer to Sil No. 380 for related performed during every refueling outage, infonnation). With bypass flow through a leating access hole cover, the cor e flow measurement system Owners of GlillWits in this category which are and, therefore, the indicated stability rone, may not operating with hydrogen water chemistry may be reliable. Justify longer ins;rction intervals consistent with the plant inservice inspection program following l'or most plants sudden fallule of an access hole cover the first access hole cover examinations, would te indicated by an increase in the total indi. cated core flow with a cettesponding power seduc.
- 2. llecause Gl! Nuclear !!ncigy performed the re- tion. Such changes in power to flow ratio would be lated safety evaluation only for a Gl! IlWlV4 as apparent to the operator, who is tiained to take stated above, owners of GlillWit/3s and IlWly appropdate actions. "Ihus, monitoring core power 5s should perfonn individual evaluations to ver- versus flow is one way of detecting separation of an ify that the safety conclusions presented in Sil, access hole cover.
e pagoP SIL No 4G2 Supplemer,!2 Revision 1 Category 1
.p. ..
. :Rc~a ctor Operating Guidelino w Notico .- ,
Gli Nuclear Energy recommends that the following 'this SIL pestains only to GE llWits. GE Nuclear reactor operating guideline te implemented until the linergy prepared this Sil. exclusively as a service for in)tial cncking examination of the access hole covers owners of Gl! 11 Wits. Gl! Nuclear Enc 4gy has not is performed and results of the examination evalu- considered or evaluated the applicability, if any, of ated. Information contained in this SIL to any plant or facility other than GE IlWits. Iktermination of A. Monitor steady state core power verses total core applicab!11ty of infonnation containco in this SIL to flow periodically and plot Ihe trend of the col. a specific llWit and implernentation of nwmmertted lected data. action are the responsitilities of the owner of that IlWit.
- 11. If there is an unexplained decrease of rnore than 5% in core power with no conesponding change No wanunty or representation e x pr essed or implied is in core flow or an increase in the total core flow, made with respect to the accuracy, completeness or immediately reduce reactor power telow about usefulness of this infonnation. General Electric 50% by Iriserting control rods. When power has Company assumes no responsibility for liability or been reduced below abou t 50%, reduce recircula- damage which may result from the use of this infor+
tion pump speed to ininimum and continue with a normal shutdown. C. Monitor the plant continually for any indication . of core instability in accordance with USNitC IB Dulletin 88-07. Supplement 1. dated Decemter O 1988. Manually scram the plant at the first indication ofinstability. To receive additional information on this subject or for assistance in implementing a recommendation, please contact your local Gli Nuclear Energy Service llepresentative,
&$ }.l Product Refercnce; , , ,
1111 - lteactor Pressure Vessel gp l .u;fNTechnical Sources t issued by J. G. Moore l Customer Service Cormnunications Manager J. P. Clark GE Nuclear Energy G. L Son) 175 Curtner Avenue, San Jose, CA 95125 l l l l O SIL No. 462 Supplomont 2 novlsion 1 P 90 3 Calogory 1
/ \j GE Nucle:r [trergy l
/_ cera,m n,nm ; April 4,1991 b) RCil-9143 airt mur z.wt Sr. a m tu tw it erzt m m na Mr. Gerald 11. Ilicks, h, Mc< har.ical Engineer Coopt Nuclear Station Nebruka Public Power District P.O. Ilox 98 Brownville, Nebraska 68321 '
Subject:
Schtdule for RenamlLIAll011_Qf_ Cit 9 Pet'1Shmud.StiPPat13 Cit $1]htlLC91tu
Reference:
OB-Nil Services Information letter (SIL) No. 462 Supplement 2 Revision 1, ' Shroud Support Accen llote CracLa*
Dear Mr. Ilicks,
in response to your inquiry, OB Nil has conducted an initial review of the recommendations made in the referenced SIL with regards to the schedule for the ultraronic (UT) rectamination of the access hole covers (AllC) at Cooper Nuclear Station. %e exact recommendation from this $11. shat applys to Cooper is,
' Access hole covers which already have teen examined with no eracking fM also should be reexamined every three years from the date of the first examination. At all llWRa opsteg on two year cycles, the reexaminations khould be performed during every refueling outage.' Our recorda indicate that the first examination of the AllCs at Cooper was performed on April 27,1989, and that no evidence of cracking was detected. Therefore, in Strict accordance with this recomnwndation, a reexamination Should be performed on or before April 27,1992. It is understood that you have an outage scheduled for the fall of 1991, but that you would prefer to perform the reexamination at your next outage, which is scheduled for the spring of 1993. This results in a four year period letween examinstions, in reviewing this Gil-NB considered the following two facts,
- l. The Cooper AllCa have a thicknea of 2 inches, while the AllCs that have been reported as cracked have a thicknea of about $/8 of an inch. It is believed that it may take more time for cracLa to initiste in these thicker covers and that the crack growth rates may be r.omewhat slower due to the differencea in streu levels.
- 2. The incidence of cracking of 6hroud head tolts at Cooper has been relatively low (1 in 1986,4 in 1988,2 in 1989 and I in 1990). It is believed, based on the field data, that this indicates that Cooper has relatively gcxxl water chemistry, which should slow down the initiation and the
. growth of cracks.
Dascd on this initial review, GII-Nil believes that it is acceptable for Cooper to perform the reexamination of the AClla in the spring of 1993, rather than the fall of 1991. Ilowever, Oll Nil does recommend that Cooper perform a visual examination (VT I) of the AllCs and an ultrasonic cramination (UT) of the $hroud head tolta during the fall of 1991 outage. If suspect indications are identified during the VT of the AllCa, or if 5 or more Shroud head bolts are found to be cracked, then the rectamination should be performed before going back to power. Sincerely yoms, o k V t> = Richard C. IlooperA, _\ Manager - Inspection Servicca Central Territory RC110PC
a cc: R. Joffe MIC TSC J. Self MIC ARO
- 11. Meys D. Erster
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CENERAL ELECTRIC COMI'ANY SIL NO. 409, REVISION 1 INCORE DRY TUBE CRACKS General Electric Company Service Information Letter (SIL) No. 409, Revision 1, dated July 31, 1986, documents cracking observed in Intermediate Range Monitor (IRM) and Source Rango Monitor (SRM) dry tubes. The cracking occurred at the top portion of the dry tube assembly adjacent to either the weld between the tube and Suide plug or the weld between the tube and the primary pressure boundary. The cracks are primarily in the perforated tube that to not part of the pressure boundary. Metallurgical exainination of a cracked dry tube assembly determined the cracking to be intergrannular with no evidence of sensitization. The cracking mechanism is believed to be a combination of crevice cot rosion cracking and Irradiation Assisted Stress Corrosion Cracking (IASCC). General Electric first performed visual exarnination of all IRM and SRM dry tubes at CNS during the 1986 Refueling Outage and no cracking of dry tubes was observed. Subsequent IRM and SRM dry tube visusi exarninations were performed by
.CNS Operations personnel during the 1989 and 1990 reiuoling outages under Preventive Maintenance (pM) No. 04642. An indication evaluated as non relevant was observed during visual exanination on SRM "D" during the 1989 Refueling Outage; re exarnination of the indication during the 1990 Refueling Outago did not reveal any visual changes. Video tapes of SRM visuni examinations f rom the 1989 and 1990 refueling outages are on filo with and available for review from CNS Operations Engineering.
CNS will perform a visual exarnination of all accessible arcu of the IRM and SRM Os dry tubes each refueling outage in accordance with SIL 409, Revision 1. General Electric Invessel Visual Inspection Procedure VT 6 or equivalent will be employed for this exatnination. The results of these examinations will be Jocumented in the NDE services report for each outage. A copy of SIL No. 409, Revision 4 is attached for additional infortnation. l l l O
EdBPA8 l INFORMATION LETTER SAN JOSE, CALIFORNIA July 31, 1986 SIL No. 409 File Tab C Revision 1 Category 2 NCORE LRY TUDE CRACKS Tine purpasc of this SIL 409, Revision 1, is to provide new information and recommendations on the crccks found in Intermediate Range Afonitor (IRhl) and Source Range Afonitor (SRhl) dry tubes. This SIL 409, Revision 1, supersedes SIL 409. Discussion Underwater tel < ,a nd video tape laspections of IRH/SRM dry tubes have
, en performed . m y -two BWRs. Cracking; or crack indications have been O de t ea 1 served cracks are in
'"ser irr tete t r ertee er tweee 91 ete-t'.e top portion of the dry tube assembly adjacent either
^11 r tae e6-l to the weld between .ae tube and the guide plug or the weld between the tube j and the primary pressure boundary. The attached Figure 1, a schematic of the ;
top two feet of the dry tube assembly, shows the locations of the observed j cracks. The cracks are primarily in the perforated tube, which is not part of , the pressure boundary. Some cracks have propagated a short distance (less ) than or equal co 0.02 in.) into the pressure boundary. ! No instruments have failed to function as a result of these cracks, none of the cracks has caused any detected leakage, and there has been no reported penetration of the primary pressure boundary. The inspections also show thut no loose pieces have been generated. Data from the visual examinations indicate a strong correlation between con-ductivity of the water and the amount of time that passes before cracking be-gins: the lower the average water conductivity, the longer the time to crack initiation. A significantly longer time to crack initiation is observed if the water chemistry satisfies the BWR Water Chemistry Guidelines published in EPRI NP 3589 SR LD, April 1985. General Electric has performed metallurgical examinations on cracked dry tube assemblics from one reactor. The following are the significant results of those examinations. O GEN ER AL (S ELECTRIC N0 W ARRAN f Y OH HFFHESEN I A flON I MHESMD OH IMPI If D 19 MADE WifH HTMLCT TO 7Hf ACCUHACV. COMetE T E NL% OH UM FUt HESS Of THIS It #CW AtA D' GE NE HAL ELE C f HIC COMPANY V6UME S NO Hi SPO*GEnuiY f OR UAMiTY OH DAMAGE WHICH MAV HCSUL T IHOM THE USE OF iHIS iNF OHMATKW
O SIL No. 409 Revision 1 llh Category 2
- 1. The cracks begin in crevice regions within approximately one-half inch of the welds shown in Figure 1.
- 2. The cracking is intergranular, and the microstructure is free of sensiti-zation. Eecause of these findings, the cracking is considered to be caused by a combination of crevice corrosion cracking and irradiation assisted stress corrosion cracking (IASCC).
- 3. This cracking is similar to that discussed in RICSIL No. 002 (Crevice Corrosion Cracking of Advanced Test Control Rod) and SIL No. 433 (Shroud llend Bolt Ctacks).
GE has incorporated design improvements into the upper portion of its replace-ment IRH/SRM dry tube assemblies and into the corresponding locations of its new Wide Range Neutron Monitors. The improvements consist of elimination of crevices exposed to reactor water and a change to a more IASCC-resistant ma-terial in the region of the cracks. The improved dry tube assemblica satisfy all the reactor interface requirements and are intended as direct replace-ments. Recommended Action GE recommends that owners of BWR/2a chrough BWR/6s implement tF* following ac-tions.
- 1. Perform visual Jnspections at the intervals shown in Table 1 of this SIL No. 409, Revision 1. The inspections should concentrate on the upper two feet of the dry tube. If crack indications are observed, GE is available to assist the utility in determining the need, if any, for replacing the dry tube. GE also can furnish procedure. and equipment needed for dry tube replacement.
- 2. Operators of plants which have been in service for several years may want to make contingency plans which consider the long lead times involved in obtaining replacement dry tubes.
- 3. Care should be taken to avoid bumping a dry tube during fuel movement.
- 4. Flow-induced vibrations should be minimized. (Please refer to SIL No.
406, "In-core Instrumentation Protection".) If you want additional information on this subject, please contact your local ' General Electric Service Representative. 7echnical Source: J. E. Charnley Issued by: ' B.11. Eldridge, M'anager Services Information and Analysis Product
Reference:
C51: Neutron Monitoring _2_ ,
e i i SIL No. 409 Revision 1 ) Category 2 l
- 1. The cracks begin in crevice regions within approximately one-half inch of I the velds shown in Figure 1. ;
- 2. The cracking is intergranular, and the microstructure is free of sensiti- !
ration. Because of these findings, the cracking is considered to be e I caused by a co:nbination of crevice corrosion cracking and irradiation assisted stress corrosion cracking (IASCC). I
- 3. This cracking is similar to that discussed in RICSIL No. 002 (Crevice !
Corrosion Cracking of Advanced Test Control Rod) and SIL No. 433 (Shroud l Ilead Bolt Cracks). ; GE has incorporated design improvements into the upper portion of its replace- i ment IRM/SRH dry tube assemblies and into the corresponding locations of its ' new Wide Range Neutron Monitors. The improvements consist of elimination of i crevices exposed to reactor 'fater and a change to a more IASCC-resistant ma- ; terial in the region of the cracks. The improved dry tube assemblies satisfy
. all the reactor interface requirements and are intended as direct replace-ments.
Recommended Action GE recommends that owners of BWR/2s through BWR/6s implement the following ac- ) - tions.
- 1. Perform visual inspections at the intervals shown in Table 1 of this SIL '
No. 409, Revision 1. The inspections should concentrate on the upper two i feet of the dry tube.- If crack indications are observed,- GE is availabic ! to assist the utility in determining the need, if any, for replacing the dry tube. GE also can furnish procedures and equipment needed for dry ; tube replacement. ; ~
- 2. Operators of plants which have been in service for several years may want to make contingency plans which consider the long lead times involved in [
obtaining replacement dry tubes.
- 3. Care should be taken to avoid bumping a dry tube during fuel movement.
4.- Flow-induced vibrations should be minimized. (Please refer to SIL No. ' 406, "In-core Instrumentation Protection".) If you want additional information on this subject, please contact your local , Gereral Electric Service Representative.
- l Technical Source: J. E. Charnicy issued by: #)b L B.11. Eldridge, ffanager -
Services Information , and Analysis I l Product
Reference:
CSI: Neutron Monitoring l r l . l l
. +
I 4
/~') SIL No. 409 U ~ Revision 1 Category 2 FIGURE 1 SCHEMATIC OF TOP PORTION OF DRY TUBE l
Fits into Top Guide f l i Adaptor l Collar l l
- . _ - Sha f t Guide Plug VT
/
1 Locations } l~[ V l Ind ations Perforated Tube t l Spring ( i l l /~ - 1 Primary Pressure Boundary (Dry Tube) l b Instrument Cavity l
/
.s .
SIL No. 409 ' Revision 1 Category 2 TABLE 1 RECOMMENDED INSPECTION INTERVALS k'ater Conductivity Dry Tube Design Meets EPRI Cuide11nes Does Not Meet EPRI Cuidelines Original 4/2 2/1 Equipment Replacements 6/3 3/1 with Crevice Elimination and Material Change EPRI water conductivity guidelines appear in EPRI NP 3589 SR LD for the cumulative service of dry tubes.
"X/Y" means the recommended inspection should be performed during the "Xth" refueling outage af ter dry tube installation. Follow on inspec-tions then should be performed every "Yth" refueling outage.
O _4_
- -- - . = . ._ - - .-- - - .
,.e . _
4 i SIL No. 409 Revision 1 Category 2 , TABLE 1 RECOMMENDED INSPECTION INTERVALS , i Water Conductivity Dry Tube Design Meets EPRI Guidelines Does Not Meet EPRI Guidelines Original 4/2 2/1 Equipment Replacements 6/3 3/1 l ' l with Crevice l Elimination and
- i. Material Change EPRI water conductivity guidelines appear in EPRI NP 3589 SR LD for the -
cumulative service of dry tubes. i "X/Y" means the recommended inspection should be performed during the "Xth" refueling outage af ter dry tube installation. Follow on inspec- , tions-then should be performed every "Yth" refueling outage. ! l t l'
- - - . , . -- - - - --. - - - , - . - , - - - . - - . - - - - - - , - - - - - - - , - - - - - , . . . . ------,--c----- - - -- - - - - . - - - ,-. - . - - . - , - - - - - - - - - - . - - , .- , . , . , -- - - - - - - - , . - - - - - , - - , - - - , - - -
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GENERAL ELECTRIC COMPANY SIL NO, 433 h V. Sl3t0UD llEAD BOLT CRACKS General Electric Company Service Information Letter (SIL) No. 433, dated February 7,1986, documenta the discovery of cracked shroud head bolts (SilBs) at several domestic Bk'Rs. The cracking mechanism was identified as crevice assisted Intergrannular Stress Corrosion Cracking (IGSCC) and occurred on the Inconel 600 shaft of the SHB in a creviced region formed.by a 304 stainless steel sleeve / , collar welded to the bolt shaf t. SIL No. 433 recommended that all creviced, old I design SHBs be ultrasonically inspected during the next refueling outage and all l cracked SilBs be replaced with a new improved design SilB without the sleeve / collar crevice. General Electric performed ultrasonic (UT) examination of the installed, old design SilBs at CNS during the 1986,1988,1989, and 1990 refueling outages. The following SHBs were found to be cracked and were subsequently replaced: , ' 1986 Refueling Outage SHB #31 1988 Refueling Outage SHB #3, #25, #33, #34
- SHB #10, #11 1989 Refueling Outage
- 1990 Refueling Outage SilB #28
- Total Replaced (As of End of Spring 1990 Outage)
- 8 SilBs l
.CNS will continue to perform UT examination of all installed, old design, Q creviced Inconel 600 SHBs each refueling outage until all the old design SilBa are C/ replaced with improved design non creviced Sil!,s. General Electric UT examination procedure UT-48 or equivalent will be employed for this examination. The results of these examinations will be documented in the NDE services report for each outego. A copy of SIL No. 433 is attached for additional information.
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. .. l j RDRlE l INFDRMATION LETTER !
j O "uc'e^n svereus & senvices oren 1'o~s - sa" aose. c^' iron ~ia l l l l 1 1 February 7,1986 SIL No. 433 i File Tab B_ Category 1 j i l SHROUD HEAD BOLT CRACKS l l l Cracking of ';hroud head bolts (SHB) has been observed at four BWR/4's and one BWR/3. The cr,.: king occurs 'In the NiCrfe alloy 600 shaf t of the SHB in a creviced region formed by a 304 SS sleeve welded to the bolt shaft. The BWR/6 uses a shroud head stud design differen t than j the BWR/2-5 design and is not susceptible to the failure mode addressed by this Service Information Letter (SIL), Complete failure of a SHB is normally detected during assenroly following shroud head removal and replacement. Complete failure has only been observed at one plant. Cracking at other plants was found by ultrasonic examina-tion (UT). The purpose of this Service Information Letter is to discuss the bolt cracks, bolt inspections and results, possible conse-quences of SHB failure and General Electric recommendations. DISCUSSION Design Descript.ig The shroud head bolt is a bi-metallic device (See Figures 1 & 2) designed to allow ren.ote assembly and disassembly that utilizes dif-ferential thermal expansion for loading of the shroud-to-shroud head flange joint. The SHB is a non-safety related component that is part of the non-safety related shroud head and separator assembly. The SHB is designed to keep the shroud head in place on the shroud during normal operation and during transient and accident conditions. The SHB's are loosened and unlatched each time the shroud head is removed
-from the vessel. The removal operation, using tooling sp?cially designed for the purpose, allcws the joint to be unloaded and the bolt disengaged from the shroud SHB lugs. When the SHB's are unloaded and disengaged, the shroud head and separator assembly can be removed from the reactor pressure vessel (RPV). Installation of the shroud head and separator assembly is performed using the same special tooling.
During installation, the bolts are latched (i.e., rotated 90 into engagement with the SHB lugs) and tightened to a torque of approxi-mately 50 f t.-lbs. A broken bolt is not capable of developing 50 f t.-lbs. and it is in this way that a failed bolt is detected. Fail-ure of the shroud head bolt does not result in loose parts. The lower
-O- part of the failed bolt cannot drop away from the sleeve and become loose because the alignment oin protrudes through the window in the sleeve and the broken segment is thus captured.
l GENERAL $ ELECTRIC NO WAAHANTY OR F4 PRESENT ATION DPfl[SSED OH IMPLIEO IS MADE WITH HESPECT TO lHE ACCUHACY, COMPLETENESS OR USE FULNESS OF THIS INFOHMATiON GENTRAL ELLCTR6C COMPANY ASSUMES NO RESPON9fullTY FOR LIAfHUf Y OH DAMAGE WHICH MAY RESUL T FHOM THE USE OF THIS INFOHMATlON
^ - ---
Sil No. 433 Category 1 Cracking Found By UT Examination An ultrasonic examination (UT) procedure has been developed to examine SHB's. A straight beam UT examination is made from the bottom of the bolt with the bolt.in place on the shroud head and separator asserrbly while the assembly is in the equipment pool. In most cases, this examination method is capable of differentiating between severely cracked or separated bolts and those which are partially cracked. Cracked SHB Examination Results The fracture surface of one of the failed SHB's has been subjected to extensive metallurgical examination. The cause of failure has been confirmed to be crevice accelerated intergranular stress corrosion cracking (IGSCC). The failure location is shown on Figure 1 and is just above the connecting weld of the collar to the shaft, within the All shroud crevice formed between the SHB collar and the SHB shaft. head bolts used on BWR/2-5's are potentially susceptible to this type of cracking. Other Considerations g
" Time in use" is one factor in predicting when cracks will initiate However, " water and grow. Variation in " loading" is also a factor.
quality" is the major factor that controls time to crack initiation and crack growth. Additional data is being obtained from on-going UT examination, review of existing data and some additional laboratory work. This on-going work will help determine the frecuency of examin-ations that will assure maintaining the integrity of the shroud-to-shroud head flange joint. At this time, there is no known safety concern. However. . if some bolts are found to be cracked in a given reactor, it should be expected that other tolts may also crack. Therefore, consideration should be given to replacement of creviced bolts to allow future operation without inspection of SHB's for cracks. RECOMMENDATIONS .
- 1. It is recommended that all BWR/2-5's perform a UT examination of all shroud head bolts the next time the reactor vessel head is removed and the shroud head and separator assembly is moved' to the equipment storage pool.
- 2. All bolts that are found to be cracked should be replaced with new bolts that do not have a crevice, gI l
1
SIL No. 433 Category 1 (3-V
- 3. If cracked bolts cannot be replaced due to unavailability of
- spare _ bolts, they should remain in place until replacement bolts can be obtained. Some structural strength is retained until the time of complete severance of the bolts. Failed bolts do not result in lost parts.
- 4. If' cracked bolts cannot be replaced or the bolt status is un-known, an evaluation should be performed to confirm that there are no safety concerns and to assess the potential risk of damage to reactor internals and Balance of Plant equipment.
Parts Availability An improved design bolt (the collar crevice has been eliminated and other product improvements incorporated) is _ now available. These parts are carried in stock in limited quantities. If demand has used-up existing stock, the factory delivery cycle is 28 weeks from receipt of order. , Please contact your local General Electric service representative for O additional information. (/ Prepared by: R.E. Legate Issued by: A // M s d_ v B.H. Eldridge, Manager Serv d = Information and Analysis Product
Reference:
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CENERAL ELECTRIC COMPANY SIL NO. 515 REACTOR FRESSURE VESSEL HEAD LINEAR INDICATIONS Ceneral Electric Service Information Letter (SIL) No. 515, dated May 30, 1990, documented the discovery of significant indications -during ultrasonic (UT) examination of a Reactor Pressure Vessel (RPV) head weld as part of a routino ASME Section XI examination. Subsequent evaluation and analysis of the indications determined that the indications were a result of an original construction welding condition; also present were laminar inclusions near the center of the dollar plate of the RPV head. The RPV head was approved for return I to service with augmented inspection requirements. During the course of the above evaluation, the actual indication was sized using both ASME Section XI Code criteria and supplemental requirements of U.S. NRC t Regulatory Guida 1,150. Using only ASMR Section XI Code criteria, the indication did not require reporting or additional evaluation. When the sizing requirements of Regulatory Guide 1.150 were applied, the indication was required to be r dispositioned by analysis (fracture mechanics). i CNS performs all required RPV examinations in accordance with the requirements of ASME Section XI. Au recommended by SIL No. 515, CNS will apply Regulatory - Guide 1.150 requirements to all RPV examinations beginning with the 1991 Refueling Outage. All affected UT procedures used at CNS will be revised prior , to the 1991 Refueling Outage to include Regulatory Guide 1.150 requirement.s. A ( copy of SIL No. 515 is attached for additional information. ! O l l b F e t
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Gf Nuclear fncigy p . . -- g v, . ,, S I L s, + ,,,m _ .,uom May 30, 1990 SIL No. 515 Category 1 REACTOR PRESSURE VESSEL llEAD LINEAR INDICATIONS Please note that this SIL No. 515 closes out RICSIL No. 051, "Re-actor Pressure Vessel llend Linear Indications", issued by CE Nu-clear Energy on May 3, 1990. Background During recent routine ASME Section XI examinations, significant ultrasonic test (UT) indications were detected in the reactor pressure vessel (RPV) top head at a GE BWR/4 located in the United States. The top head at this plant is unclad. The indications were in one circumferential weld of the head top dome. A These indications are different from those reported in RICSIL No. ()) - 050, " Reactor Pressure Vessel llead Clad Cracking", issued by CE Nuclear Energy April 23, 1990, and the results of evaluations have confirmed that the issues reported in RICSIL No. 050 are not re-lated to those addressed in this SIL No. 515. Discussion The RPV top head is fabricated by welding a domed plate (called a
" dollar" plate) to vertical, toroid shaped plate sections to form the hemispheric top dome of the head assembly. The circumferen-tial weld joining the dollar plate to the vertical sections is called the dollar plate weld. As reported in RICSIL 051, UT indi-cations were discovered in the area of the weld and the heat af-forted zone on the dollar plate. The initial UT data revealed a subsurface indication 12 inches long and approximately 2 inches deep, possibly connected to an intermittent surface indication about one inch long. Figure 1 shows the construction of the RPV top head and the location of the indications.
Following the initial observations, manual and automated ultra-sonic testing, radiography and review of the fabrication records and construction radiographs were performed. These analyses confirmed that the indication is not a crack but is an original construction welding condition that shows no evidence of in-service growth. The final UT sizing, based on a conservative combination of manual and automated examination data, shows the g
indication to be a subsurface flaw with a maximum depth of 0.5 inches and maximum length of 5 inches. Figure 1 DOL 1AA PLATE DOLLAR PtATe WELD ARCA Or NDCATKW s r j . convtD
-e TOP.US SE Guim s
{ O ooooooooooooooa_ w > NE RPV Top Head Assemoly Non-destructive evaluation tests also showed the presence of ac-ceptable laminar inclusions near the center of the plate. These inJ1usions, typically called segregates, are metallurgical condi-tions from the plate's original manufacturing. These segregates were located in the area of the original indications which caused the ultrasonic reflections that initially were evaluated as a more extensive flaw. Examinations of the inner surface of the head by liquid penetrant and magnetic particle techniques also confirmed tuat the indication is sub-surface. The originally reported linear indication on the inner surface was determined to be an ac-l ceptable surface condition where weld build up was applied to the l dollar plate to correct a misantch at the joint. The final evaluated size o'f the indication was determined in ac-cordance with ASME Code Section XI and the supplementary require-l ments of Regulatory Guide 1.150. Based solely on ASME Code Sec-tion XI criteria, the indication did not require reporting or ad-ditional evaluation. However, when the supplementary sizing requirements of Regulatory Guide 1.150 were applied, the indica-tion was required to be dispositioned by analysis. A tracture-1 SIL No. 515 Category 1 Page 2
i
' indication to be a subsurface flaw with a naximum depth of 0.5
. inches and maximum length of 5 inches.
Figure 1 DOLLAA PLATE I DOLLAR PLATE WELD AREA 0F NNCAT10tJ N .
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-Non destructive evaluation tests also showed the presence of ac-ceptable laminar-inclusions near the center of the plate. These
' inclusions,otypically called segregates, are metallurgical condi-tionn.from_the-plate?s original manufacturing. <These segregates ,
-were located in the area of the origina1' indications which caused, *
- the ultrasonic reflections.that initially were. evaluated as'a more extensive flaw. Examinations of the inner surface of the head by.
- liquid penetrant _and magnetic particle _ techniques also confirmed that the indication is sub-surface. The originally reported linear indication on the inner surface was determined to be an ac-ceptablo_ surface condition where weld build up was-applied to the
' dollar p' late. to correct a mismatch at the- joint.
'The final evaluated size o'f the indication ~was determined in ac-cordance with AS.ME Code Section XI-and the supplementary require-ments'of Regulatory Guide 1.150. Based solely on ASME Code Sec-tion XI. criteria, the indication did not require reporting or ad-ditional evaluation. However, when the supplementary' sizing requirements.of Regulatory Guide 1.150 were applied, the indica.
tion was required to be dispositioned by analysis. A fracture SIL No. 515
- Category 1 Page'2-
_ _2 . . - . _ , _ _ -
mechanics evaluation was performed which showed that the indica. s - _ tion is acceptabic without repair for operation for the next-oper-ating cycle. .The top head has been approved for return to service 1 with augmented in service examination performed in accordance with ASME Code Section XI requirements.
~
Recommended Based on the experience at this GE BWR/4, GE Nuclear Energy recom. Action mends that owners of GE BWRs which have unclas! RPV top heads fol. Iow the ASME Section XI ultrasonic examination procedure, supple-mented by Regulatory Guide 1.150 requirements, for top head exami-nations. To receive additional information on this subject or for assis-tance in-impicmenting-a recommendation, please contact your local GE Nuclear Energy Service Representative. Technical T. L. Chapman Source Notice This SIL pertains only to GE BWRs. GE Nucicar Energy prepared-this.SIL exclusively as a service for owners of CE BWRs, GE tiu-clear Energy has not considered or evaluated the applicability, if any,_of information contained in this SIL to any plant or facility other than GE BWRs. Determination of applicability of information contained in this SIL to a specific BWR and implementation of rec-ommended action are the responsibilities of the owner of that BWR, O) No warranty or representation expressed or implied is made with respect to the accuracy, completeness or usefulness of this infor-j mation. General Electric Company assumes no responsibility for. _ liability or damage which may result from the use of this informa-tion,
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Issued by i ! -J. G. Moore
-Customer Service. Communications Manager l' ' Product Bil - Reactor Pressure Vessel' l~ Reference SIL No. 515 Og g Category 1
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& COOPER NUCLEAR STATION ISI PROGRAM V Augmented Inservice Inspection Program For Service Water (SW) and Reactor Equipment Cooling (REC) Pipe Supports
References:
. Letter NPPD (C. A. Trevors) to U.S. NRC dated November 15, 1990, i Subj ect; NPPD Response to NRC Inspection Report 50 298/90 15 Cooper Nucicar Station l In the above reference letter, NPPD/CNS committed to create an augmented ISI program for component supports (including their associated integral or nonintegral attachments) of the safety related portions of the service water and reactor equipment cooling systems, The program was intended to provide VT-3/4 visual-examination of the component supports for inspection.
The Augmented ISI Program for SW and REC Pipe Support has been added to the CNS j ISI Program as a separate, new Volume 3 to the existing program. A copy of the 1 above reference letter is attached for additional information. ) l l () t O
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U.S. Nuclear Regulatory CommisuMC WPV . Document Contral Desk f0PS ENG SUPV 3 l Vashington, DC' 20555 TDA MCR
Subject:
NPPD Fesponse to NRC Inspection Report 50 298/90 15 Cooper Nuclear Station . Docket No. 50 298
References:
.1, Letter from S. J. Collins (NRC) to G. A. Trevors (NPPD), dated April 30, 1990, Transmittal of Inspection Report 90 15. -. .
- 2. Letter from G. A. Trevors (NPPD) to S. J. Collins (NRC), dated ~
May 30,1990, NPPD Response to Inspection Report 50 298/90 15.
- 3. Letter from S. J. Collins (NRC) to G. A. Trevors (NPPD), dated"-
A September 17, 1990. __,, i Q) 4 Lotter from G. A. %"rors (NPPD) to NRC, dated October 12, 1990. l - , . . l l Gentlemen: i This letter is written in response to your letters dated April 30, 1990, and l' September 17, 1990, concerning Inspection Report 50 298/90 15. Therein you l indicated that one of our activities was in violation of NRC requirements. I l Following is a statement of the violation and our response. I STATEMENT OF VIO1ATION Failure to Include ASME Class 3 Nonintecral Component Suenorts in the IST Procram Technical Specification 4.6.G for Cooper Nuclear Station states, in part, that inservice inspection of Am Code Class 1, 2, and 3 components shall be performed in accordance with S6erion XI vf the ASME Boiler and Pressure Vessel Code and ' applicable Addenda. l Paragraph IUD 2620, "Viwoal Ex minat. ion. VT 3" , of Section XI of the ASME code l states, in part, that the corputy : supp9rts and restraints within the boundary of each system specified in the excnnacien categories of Table IUD-2500-1 shall be
. subject to the visual examination nf VT-3 and shall be performed at the frequency l
-specified in Table IVD 2500 1 (which is each inspection interval).
9 M M w i m m q r g is M 2 :!; W m ; w Y ? R w a n N % V n M E & M M A
U.S. Nuciser Rsgulatory Commission Nove bar 15, 1990
- Page 2 (q) - '
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Contrary to the 3 nonintegralycomppnent supports from within_" the - boundary of eacl{ ,4,b..oypg,,Classsys, s tem specified"in Table IWD 2500/1 vere not inc and second 10 yuGSI ' Program for VT 3 visual examinations.
- l . -. .. .y Reason for Viol'atioE ' - -~
Q Since the issuahu,,,,of, Inspection Reportthorns
-- . .l '
90 ;1hl have; been several discussions j between the NRQn,d_ g?D concerning-the -interpytation of ASME Section XI code requirements fcr inspecting Class 3 component'euppprts. This issue has been further complicated byAbg,,,ap[ ent ambiguity ofQiidfademption criteria of the code for i Class .3 componerne[(s'.~.Co~op'er Noclear-Stati'dn's current 10-year ISI ins interval is bass ed on :the 1980' Edition-Witite@M Addenda of Section XI. . g@ The NRC's interpretation of the code differs from NPPD's in the area of the selection process for supports to be examined. Due to the complexity of the code requirements NPPD made a good faith effort to obcain clarification by submitting two inquiry questions to the ASME Section XI Code Committee at its quarterly meeting , in Nashville, Tennessee, on May 14, 1990. These inquiry questions and the Code , Committee repliec, which were submitted in Reference 2, supported NPPD's original conclusion. Af ter further review of this issue, including re evaluation of all available information, the District has again concluded that there may not have been a violation of NRC requirements, since NPPD's interpretation of the code did not violate the intent of ASME Section XI or Technical Specification 4.6.G. It is NPPD's understanding from Reference 3 that the NRC intends to submit inquiry q questions to the ASME Section XI Code Committee meeting in December,1990 to further tg clarify its position. NPPD supports this action and is willing to assist in oevelopment of additional inquiry questions to further clarify the intent af the code concerning examination requirements for Class 3 component supports. Corrective Stens Ubich Hnve Beet' Taken and Results Achieved At this time, the District is not yet convinced that a violation existed. However, , as a result of our re evaluation of the CNS ISI Program selection criteria for supports, certain Class 3 component supports associated with integral attachments weres added to the CNS ISI Program. The addition of these originally exempted supports is based upon using an "and" in lieu of an "or" requirement in the District's interpretation of IWD-1220.2(a) and (b). These supports and their associated integral attachments were inspected during the 1990 Refueling Outage and all were found to be satisfactory. Correerive Stens Uhich Will Be Taken to Avoid Further Violations . As a _ result of our extensive re-evaluation of code requirements in response to this Notice.of Violation, NPPD considers it prudent to further supplement the CNS ISI Program .for component supports (beyond ASKE code requirements) . This augmented inspection program vill balance the CNS ISI Program with the addition of VT-3/4 examination of selected component supports (including their associated integral or nonint.egral attachments) of the safety related service water and reactor equipment coo. ling systas, c. :. :- - 4 Furthermore, a representative sample of supports associated with noninte5ral /G attachments will be added to the augmented inspection program for non-exempt , d portions of Class 3 piping.
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- U.S. Nuclear Regulatory Commission
_Nevsmbsr 15, 1990 Page 3 fT U This action is consistent with current CNS management philosophy that a representative sample of safety related supports should be routinely inspected, regardless of code requirements. Date Vhen Full Compliance Vill Be Achieved All program enhancements mentioned herein will be included in an augmented inspection program by the completion of the 1991 Refueling Outage. The District L of the opinion that a violation of NRC requirements may not have occurred. The December 1990 code inquiry sheuld clarify this position. If not, a supplement to this response vill be issued addressing all remaining outstanding issues. Please contact me if you have any questions or require additional information. Sincerely. G. . revors - Senior Staff Advisor Nuclear Power Group CAT: Cell:sa i
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cc: Regional Administrator U.S. NRC Region IV NRC Resident Inspector Cooper Nuclear Station bc: NRC Distribution
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DESCRIPTIDN OF THC AIS1 PROfFAN f1ANtJAL SEl_CCT1DN OF SUPHIHTS (Cont.I The AISI pr ogr ama marma l is efivided into 4 sectlons. The first of supports are unique to the AISI program and not to be section is the test arwf gener a l description. The setorwJ confused ev i th ASMC Sectio, XI support categories F-A, F-H, etc. section irec ludes the tables Iisting the supports arid the Osirg the N-491 rules for ASf1E Class 3 piping a. a schedule of examination. The third sect ion contain s the guideline for selection, approximately 10% of the supports sn piping isometric drawinas showing the locations of the each category e ere selected in both the REC and SW system for supports listed in the AISI program. The fourth section contains "rone drawirwys" eehich are simply plant layout routine examination for the AISI program (r ef . TABLE T1. drawings marked with AISI rene ruebers which are tesed for locating supports. The zone rumber s are Iisted along ith the I n sel ec t i rw; supporis for examination, consideration was given supports in the second sect ion of the AISI program manual. to the location of the support on a Isne, distributinn of the sample throughout the= system, acc ess f or exaeination and GErdRfC EDITORIAL NOTES FOR TARLES IN THF AISI PROINAM per sonne l radiation exposure. With r epect to location, some pr ef er enc e meas g a ven to selection of supports near valves or Wher e hanger tar aw i ngs ha ve mu l t i p l e shes t s ne i t h di f f er ent fittings, however, that was balanced try consider at ion for 1. revisions, the revisions for each sheet are Iisted sample distr ibut non, access and per sonnel esposure. sequentially, e.g., R1,0,2 for Sheet 1, Rev. 1, Sheet 2, Rev. O and Sheet 3. He*v . 2. An "Rt)* designatinn indicates that the revision block on the draoeing was twst filled, TADLC 1. Number of Supports arwf Sample Sirms a prac t sce on older GrinelI dra ings irwf s c a t i ng that the System X-A X-H X-C X-D 1-E Al1 dr a*e i rwJ has wwt been aevised. 79 107 6 f2 21 223* 2. The hot load tHL) and cold l oa(f (CLI for spr a rw; can REC Total bargers are expr esse-d a n uni t s of pourwis . REC Sample 9 12 1 2 3 27 H2 16 16 11 2868 3. The designation "(11" follo.eing the deu rsption of a SW Total 162 suppert n ewf ic a tes that the support is i ntegr al l y attached, SW Sastple IO 9 3 3 3 36 i.e., eeelded to the pe essur e boundary.
*The total rumber as slightly le ss than the total of all the supports in each category because some indtvidual supports 4 The desagnatson *(21" f o f l owi rwJ the descript on of a rod or had components of more than one category.
strut type supper t s rwf i t.s tes that the suppor t has 2 rods ( tr apere type) or has 2 '.t r u t s. SCHEDa I S. Res t ra s nt s at ** generally descrabed as "oow" for bow-frame The examinatnon srhedule is i n t erwied to gener a l l y foilow Code type restraints or o ther =e s se= becadly riesc r i be=d as "br ac e, " C.sse N-491, Table 2410-2, inspec t n an Pr ogram B, ise. rwir aa l l y i rwi s c at n rwJ a sangle sneeber .A i c h e.e y or may tw t have a clamp or U-tin ! t at ourwt the pipe, approxamately I/3 of ti e suppor ts in the sample for each c a tegor y wall be emaanned during ear.h period of the i nter v a l . 6. The words "same- assemb l y as" in the r emar k s c o l umn are used F or the= HEC system only, categorles X -C arwf X -D eer e c omb i red to acidicate that two restrasnts share a c omewyn framo. For for whedu l a ng due to the small rm Wier of supports an these example, the reearks for SW-S-76 state "Same Ass *% I y as catagornes. Seeslarly, a restraint a rw) a y od harwJer may be S'-S-R3." Because at o.a s ries i red that the a nspec t s on interval for the d .crabed an r s= mar k s as "a t t ac herf to* one another. AISI pr ogr am be cormastent with that of the Cooper fbcfrar The designation "tAl" followarwl " Insulation" sn the r emar k s Station Section XI ISI prograce, all of the exami na t n ers f or 7. pernods I and 2 (through year 7 of the intervall have been column t rw1:c a tes that the insulatten material is asbestos. schedulad for per s od 2, outage 3, the last outage i n per s od 2. S**ee t %mhet sa 1
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