ML17265A474
| ML17265A474 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/01/1998 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-01, GL-97-1, TAC-M98567, NUDOCS 9812090009 | |
| Download: ML17265A474 (8) | |
Text
CATEGORY 1 REGULA Y INFORMATION DISTRIBUTION. SYSTEM (RIDS)
ACCESSION NBR:9812090009 DOC.DATE: 98/12/01 NOTARIZED: YES FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G
AUTH. fA(4E '.
AUTHOR AFFILIATION MECRECY,R.C.
Rochester Gas E Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION VISSING,G.S.
DOCKET I 05000244
SUBJECT:
Forwards response to NRC 980901 RAI re degradation of CRDM/
~
CEDM nozzle E other vessel closure head penetrations,per GL 97-01.Response addresses issues which are deemed to be generic in nature.
DISTRIBUTION CODE: A075D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: GL-97-01 Degraduation of Control Rod Drive Mechanism 6 Other NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72)
T'essel E
05000244 G
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1 NOTE TO ALL "RIDS" RECZPIENTS:
PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)
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aNfy ROCHESTER GAS ANDELECTRIC CORPORATION 89EASTAVENUE, ROCHESTER, N Y. 14649 0001 AREA CODE 716-546-2700 ROBERT C. MECREDY Vice President Ntscteor operations December 1, 1998 U. S. Nuclear Regulatory Commission Document Control Desk Attn:
Guy S. Vissing Project Directorate 1-1 Washington, D.C.
20555
Subject:
Ref. (1):
Ref. (2):
Response
to Request for Additional Information (RAI) Related to Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations (TAC No. M98567)
R. E. Ginna Nuclear Power Plant Docket No. 50-244 s
Letter from Guy S. Vissing (NRC) to Robert C. Mecredy (RG&E),
SUBJECT:
GENERIC LETTER (GL) 97-01, "DEGRADATIONOF CRDM/CEDM NOZZLE AND OTHER VESSEL CLOSURE HEAD PENETRATIONSn REGARDING THE ROCHESTER GAS AND ELECTRIC CORPORATION RESPONSES TO GL 97-01 FOR THE R. E. GINNA NUCLEAR PLANT (TAC NO. M98567), dated September 1, 1998 Memorandum from Jefferey F. Harold (NRC) to S. Singh Bajwa (NRC),
SUBJECT:
SUMMARY
OF THE MEETING ON SEPTEMBER 29, 1998, BETWEEN THE NRC STAFF AND MEMBERS OF THE PRESSURIZED WATER REACTOR (PWR) OWNERS'ROUPS AND NUCLEAR ENERGY INSTITUTE (NEI) TO DISCUSS THE STATUS OF MULTI-PLANT ACTION TAC NOS. OPENED IN REGARD TO ISSUANCE OF GENERIC LETTER (GL) 97-01, dated November 5, 1998
Dear Mr. Vissing:
By Reference ili, the NRC staff requested additional information regarding degradation of CRDM/CEDM nozzle and other vessel closure head penetrations.
By Reference (2), the NRC staff reported on discussions that had been held between the staff and an industry working group, PWR Materials Reliability Project (MRP), that had been formed to address the issue of GL 97-01. It was agreed that a generic, integrated response from the PWR MRP that would address the NRC's generic information requests in the RAIs on behalf of NEI, EPRI, and all of the participating members of the CEOG, B&WOG, and WOG would be submitted to the NRC in mid-December 1998.
The attachment to this letter provides the requested plant specific information and reference to the PWR MRP response for those issues which are deemed to be generic in nature.
~ or 1.A 98i20'F0009
'st8120i PDR ADOCK 05000244 P
It should be noted that RGB has decided to perform an inspection of the reactor vessel head pehetratio'ns during the next refueling outage, scheduled to begin March 1, 1999.
Very truly yours, Robert C. Mecredy Attachment Subscribed and sworn to before me on this 1st day of December, 1998 Notary Public CHRISTINAK. SARDOU Notary Public, State of New York Replstratlon No. 01 SA6015061 Genesee County Commission Expires October19, 20 oo xc:
Mr. Guy S. Vissing (Mail Stop 14B2)
Project Directorate I-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector
'l)90f5" i 4> ll"Pl'
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RESPONSE TO NRC REQUEST FOR ADDITIONALINFORMATION(RAI)
REGARDING THE RG&E RESPONSES TO GENERIC LETTER (GL) 97-01 FOR THE R. E. GINNA NUCLEAR PLANT
-Indicate whether RGE is participating in the latest industry integrated program developed for the assessment of VHP nozzles in domestic PWR designed reactors. If so, please provide the followinginformation with respect to the industry's integrated program for assessing PWSCC in domestic PWR VHP nozzles:
a.
Identify all PWR Owners Groups diat RGE is a member of.
Response
RG&E is a member of the Westinghouse Owners Group and as such has participated in the Owners Group and NUMARC (NEI) Activities related to the CRDM penetration issue since its inception in the early 1990's.
b.
Indicate whether RGE is participating in the latest integrated programs developed by these Owners Groups for assessing PWSCC in PWR VHP nozzles, and ifso, whether or not any ofthe followingtopical reports are applicable to the assessment of VHP nozzles in the Ginna reactor design:
Topical Report BAW-2031, developed for member utilities and plants in the Babcock & Wilcox Owners Group (B&WOG)
Topical Report CE NPSD-108$, developed for member utilities and plants in tIre Combustion Engineering Owners Group (CEOG)
Topical Report WCAP-14901, Revision 0, andlor WCAP-14902, Revision 0, developed for member utilities and plants in the Westinghouse Owners Group (WOG)
Response
RG&E is currently participating in the Westinghouse Owners Group and PWR Materials Reliability Project (MRP) activities.
Topical reports WCAP-14901 and WCAP-14902 are applicable to Ginna.
- c. Ifany ofthe topical reports listed in Item I.b. above are applicable to the assessment of VHP nozzles in the Ginna reactor design, provide a description ofthe probabilistic susceptibility model being endorsed to assess the VHP nozzles in the Ginna reactor design.
Include the following information with respect to the description ofthe susceptibility model:
Response
A description of the Westinghouse susceptibility model is included as Enclosure 7 of the PWR MRP integrated response.
(I)
Provide the model's relative susceptibility ranking for the VHP nozzles in the Ginna reactor design to the rankings compiled by other utilities applying the model to the assessment ofthe VHP nozzles in their plant designs.
In addition, ifone ofthe Westinghouse Topical Reports is applicable to the assessment ofthe VHP nozzles in the Ginna reactor design, compare the susceptibility ranking of VHP nozzles in the Ginna reactor design to that of WOG member plants applying the other model to the assessment ofthe VHP nozzles at their plants.
Include the basis for establishing the ranking ofyou plant relative to the others.
Justify why the crack initiation and growth susceptibility model used foryour VHP nozzles is considered to yieM as reasonable a ranking as would application ofthe other probabilistic failure model being used be WOG member utilities.
Provide a composite susceptibility ranking ofall WOG member plants and any conclusions which may be drawn from such a ranking with respect to the assessment of PWSCC in the Ginna VHP nozzles relative to those in other WOG member designs.
Response
The industry ranking of plants is provided by the Industry Histogram in Enclosure I of the PWR MRP integrated response.
Ginna is included in the histogram using the Westinghouse'model.
(2)
Describe hoiv the probabilistic susceptibility model used for the assessment ofthe VHP nozzles in the Ginna design ivas bench-marked, and provide a list and discussion ofthe standards the model ivas bench-marked against.
Response
The benchmark of the Westinghouse model is discussed in Enclosure 2, Question 2, Response 2.b. of the PWR MRP integrated response.
(3)
Provide additional information regarding hoiv the probabilistic susceptibility model being used for the assessment ofthe VHP nozzles in the Ginna design willbe refined to allow the input ofplant-specific inspection data into the model's analysis methodology.
Response
The Westinghouse probabilistic susceptibility model can be refined for plant specific inspection data as described in Enclosure 2, Question 3, Response 3.b. of the PWR MRP integrated response.
(4)
Describe how the variabilityin product forms, matenal specifications, and heat treatments used to fabricate eacli VHP nozzle in the Ginna reactor design is addressed in the probabilistic susceptibility model being used for the assessment of these nozzles.
Response
Variability in product forms, material specifications, and heat treatments and how they are addressed are contained in Enclosure 2, Question 1, Response 1.b. of the PWR MRP integrated response.
2.
Indicate whether any final conclusion has been made to commence with the volumetric examinations ofthe VHP nozzles at the Ginna Nuclear Station in 1999. Ifthe decision has been made not to commence with the examinations of the VHP nozzles in the Ginna design, justify ivhy RGE considers it to be safe to operate the Ginna plant wiNsout performing volumetric examinations ofthe Ginna VHP nozzles.
Response
In support of the industry Integrated Assessment Program, the Ginna reactor vessel head penetrations are scheduled to be inspected during the 1999 refueling outage.
3.
Identify all other plants wJrere volumetric examinations were performed on the VHP nozzles in the reactor design.
Include a more in-depth summary ofthe inspection results at the other plants which RGE considers to be applicable to the evaluation of VHP nozzles at Ginna, and identify the heats ofmaterials for the Ginna VHP nozzles which are considered to be bounded by the examination results at these plants.
Response
The listing of plants that have performed examinations willbe provided as part of the submittal to be issued by the PWR MRP.
Per Westinghouse Proprietary Class 2 Report, WCAP 13493, and RG&E review of the report, Ginna contains the same three material heat numbers as the Point Beach Unit 1 reactor vessel head.