NG-98-1261, Forwards Addl Info Related to Response to GL 92-01,suppl 1, Reactor Vessel Structural Integrity, as Requested in

Forwards Addl Info Related to Response to GL 92-01,suppl 1, Reactor Vessel Structural Integrity, as Requested in

ML20236T388
Person / Time
Site:	Duane Arnold
Issue date:	07/20/1998
From:	Peveler K IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:	NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, NG-98-1261, TAC-MA1188, NUDOCS 9807280150
Download: ML20236T388 (5)

Text

_ _ _ _ _ _ _ _ - - _ . .__ _ _ - _

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ALLI ANT g UTILITIES as nam ant.

IES Utilities EIh",iNiN)T**

Palo IA 52324 9785 a: Ifl N July 20,1998 ~"*'""""'"""'

NG-98-1261 Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station 0-PI-17 Washington, D. C. 20555-0001

Subject:

Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 Response to Request for AdditionalInformation Regarding Reactor Pressure Vessel Integrity - Duane Arnold Energy Center (TAC No.

mal 188), dated April 21,1998 File: A-10lb, A-275, B011

Reference:

1) NRC Generic Letter (GL) 92-01, Revision 1, Supplement 1 (GL 92-01, Rev.1, Supp.1), " Reactor Vessel Structual Integrity", dated May, 1995

2) L. J. Tilly,"Duane Arnold RPV Surveillance Materials Testing and Analysis," General Electric Nuclear Energy, San Jose, C A, July,1997, (GE-NE-B1100716-01)

3) T. A. Caine,"Duane Arnold Energy Center Reactor Pressure Vessel Surveillance Materials Testing," General Electric Nuclear Energy, San Jose, CA, July 1986,(GE Report NEDC-31166)

4) Letter, K. Peveler (IES) to NRC," Reactor Pressure Vessel Surveillance Materials Testing and Analysis," NG-97-1834, dated October 17,1997

5) Letter, J. Franz (IES) to NRC," Request for Technical Specification Change (RTS-200): " Reactor Vessel Pressure - Temperature Curve Update", /

NG 98-0493, dated April 15,1998

/

Dear Sirs:

In the subject letter, the NRC staff requested that Alliant-IES Utilities Inc. provide additional information related to its response to Generic Letter 92-01, Supplement 1 (reference 1). The j enclosed attachment provide. the information requested.

9807280150 DR 980720 P p ADOCK 05000331 ,

PDR l

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l NG-98-1261 July 20,1998  ;

Page 2 of 2 i l

l 1

Shculd you have any questions regarding this matter, please contact this office.

Since ely, j l

W

'enneth E. Peveler Manager, Regulatory Performance t

l

Enclosure:

Response to Request for Additional Information Regarding Reactor Pressure Vessel integrity - Duane Arnold Energy Center (TAC No. MA1188) cc: L. Sueper K. Putnam E. Protsch J. Franz t

D. Wilson C. Paperiello (Region Ill)

R. Laufer (NRC-NRR)

M. Boyle (NRC-NRR)

L NRC Resident Office DOCU 1

L. .

i

l 1

l Enclosure to NG-98-1261 Page1of3 l

Response to Request for AdditionalInformation Regarding Resetor Pressure Vessel Integrity - Duane Arnold Energy Center (TAC No. MA1188) r l

l Request 1: [ Provide] an evaluation of the bounding assessment in (BWRVIP-46," Update of Botnding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity Issues"] and its applicability to the determination of the best-estimate chemistry for all of your RPV beltline l welds. Based upon this reevaluation, supply the information necessary to completely fill out the data requested in Tabic i for each RPV beltline weid material. If the limiting material for j j your vessel's P-T limits evaluation is not a weld, include the information requested in Table i for the limiting material also.

Response: Table 1 below provides the information requested.

The conner and nickel chemistry for each weld wire heat is taken from the Fabrication Report of CBI Nuclear Company dated October 11,1972. The beltline weld chemistry variability is less than 0.05% and therefore per BWRVIP-46 does not impact the beltline P-T limits.

EOL ID Fluence is calculated using information from the surveillance capsule removed from

, the RPV at 14.7 EFPY The peak surface fluence at 32 EFPY is the same as the nominal 2

value (3.6 x 10" n/cm ) that is calculated from the first capsule dosimetry as reported by GE (Reference 3).

Assiened Material Chemistry Factor (CF) and Method of Determining CF. The CF is based on a least squares fit of surveillance data using the actual shifts of the Charpy specimens within the surveillance capsule (Reference 2).

Initial RTg (RTunrq)* is taken directly from mechanical properties of beltline weld i materials from test data information frorn General Electric Quality Assurance records (Reterence 2).

n; is the standard deviation on initial RT NuT, which is taken to be O' F (Reference 2). j i

g3 is the standard deviation on DRTuur (Reference 2) l Marcin* is based on riandard deviation, and is found by the following equation:

l (Reference 2). i i

1-Margin = 2]cr/ .cr/

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l l

l , , Enclosure to NG-98-1261 Page 2 of 3 ART at EOL* is found using the specific relationship from Reg. Guide 1.99, Rev. 2.

d

• Multiple values may appear in the table columns associated with these parameters for a given heat because these weld wire heats were used on more than one weld.

The limiting material for Duane Arnold is beltline plate 1-20 (heat number B0436-2).

l Request 2: If the limiting material for your plant changes or if the adjusted reference temperature for the limiting material increases as a result of the above evaluations, provide the revised RTwor value for the limiting material. In addition, if the adjusted RTynr value  ;

increased, provide a schedule for revising the P-T limits. The schedule should ensure that compliance with 10 CFR 50 Appendix G is maintained.

l Response: Neither the limiting material nor the adjusted reference temperature for the limiting material for the DAEC changes as a result of the above evaluations. By letter dated October 17,1997 (Reference 4), the methods and results of testing performed on the second l materials surveillance capsule, removed from the DAEC RPV in October 1996, were d transmitted in accordance with 10 CFR 50, Appendices G and II. Included in the report were revised P-T curves valid to 32 effective full power years. On April 15,1998, the DAEC submitted a request for an amendment to the Technical Specifications to update the P-T limits based on these results (Reference 5).

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