ML20197C804

Informs That Braidwood Station Concurs with Proposed Encl Response to NRC Re RPV Integrity at Braidwood & Byron Stations

ML20197C804
Person / Time
Site:	Byron, Braidwood
Issue date:	09/02/1998
From:	Schwartz G COMMONWEALTH EDISON CO.
To:	Krich R COMMONWEALTH EDISON CO.
References
NUDOCS 9809140213
Download: ML20197C804 (9)

Text

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I Memorandum l

1 Date: September 2,1998 To: R. Krich, Regulatory Senices Vice President  ;

Subject:

Response to Request for Additional Infonnation Regarding Reactor Pressure Vessel Integrity at Braidwood Station, Unit Nos. I and 2, and Byron Station, Unit Nos. I and 2.

Byron Nuclear Power Station, Units I and 2 Facility Operating Licenses NPF-37 and NPF-66 NRC Docket Numbers: 50 - 454 and 50-455 l Braidwood Nuclear Power Station, Units 1 and 2 Facility Operating Licenses NPF-72 and NPF-77 NRC Docket Numbers: 50 -456 and 50-457

References:

Comed letter, R. M. Krich to U.S. NRC Document Control Desk," Response to Request for Additional infonnation Regarding Reactor Pressure Vessel Integrity at Braidwood Station, Unit Nos. I and 2 and Byron Station, Unit Nos. I and 2", dated September 2, 1998.

We have revic sed the attached Comed's response for additional information (RAl) regarding Reactor Pressure Vessel Integrity at Braidwood Station. Based on the infonnation contained in the attachment, Commonwealth Edison has concluded that the currently licensed Pressure Temperature and Low Temperature Overpressure Protection limits for Braidwood Station, Units I and 2 do not require revision.

The Braidwood Station concurs with this response as proposed for Comed. If there are any questions or comments concerning this letter, please contact T. W. Simpkin, Regulatory Assurance at extension 2980.

Respectfull ,

/

G. K. Schwartz Station Manager Braidwood Station Enclosure cc: D. Chrzanowski G. O'Donnell T. Simpkin f

f l 9909140213 990902 ?

PDR ADOCK 05000454 p PDRt Oks98030. doc

1.

i j September 1,1998 7

U. S. Nuclear Regulatory Commission 4

ATTN: Document Control Desk Washington. D. C. 20555 - 0001 1I Byron Station, Units 1 and 2 Facility Operating Licenses NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

. Braidwood Station, Units 1 and 2 Facility Operating Licenses NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457

Subject:

Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity at Braidwood Station, Unit Nos. I and 2, and Byron Station, Unit Nos. I and 2.

Reference:

J. B. Hickman (U. S. Nuclear Regulatory Commission) to O. D. Kingsley (Commonwealth Edison Company) letter dated June 1,1998.

The purpose of this letter is to provide the response to the referenced request for additional information (RAI). The information and tabular summary requested in Sections 1.0,2.0, and 3.0 of the RAI are contained in an attachment to this letter.

Based on the information contained in the attachment, Commonwealth Edison has concluded that the currently licensed Pressure Temperature and Low Temperature Overpressure Protection limits for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, do not require revision.

. Please address any comments or questions regarding this matter to Mr. David J.

Chrzanowski at (630) 663-7205.

Respectfully, R. M. Krich Vice President - RegulatorfServices

ATTACllMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING REACTOR PRESSURE VESSEL INTEGRITY REOUEST FOR ADDITIONAL INFORM ATION Based on information received in Reference 1, in accordance with the provisions of Generic Letter 92-01, Supplement 1, the NRC requested the followi..s ,aformation be provided for Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2.

Section 1.0: Assessment of Best-Estimate Chemistry

" 1. An evaluation of the information in the reference above and an assessment ofits applicability to the determination of the best-estimate chemistry for all your RPV beltline welds. Based upon this reevaluation, supply the information necessary to completely fill out the data requested in Table 1 for each RPV beltline weld material. Also provide a discussion for the copper and nickel values chosen for each weld wire heat noting what heat-specific data were included and excluded from the analysis and the analysis method chosen for determining the best estimate.

If the limiting material for your vessel's pressurized thermal shock / pressure-temperature (PTS /PT) limits is not a weld, include the information requested in Table 1 for the limiting material also. Furthermore, you should consider thc information provided in Section 2.0 of this RAI on the use of surveillance date when responding."

Section 2.0: Evaluation and Use of Surveillance Data "2.0 That (1) the information listed in Table 2, Table 3, and the chemistry factor from the surveillance data be provided for each heat of material for which surveillance weld data are available and a revision in the RPV integrity analyses (i.e., current licensing basis) is needed or (2) a certification that previously submitted evaluations remain valid. Separate tables should be used for each heat of material addressed. If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in the tables for the limiting material (if surveillance data are available for this material)."

l Section 3.0: PTS /PT Limit Evaluation j "3. 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 RTrrs value for the limiting material in accordance with 10 CFR 50.61. In addition, if the adjusted RTmr value increased, provide a schedule for revising the P-T and LTOP limits. The schedule should ensure that compliance with 10 CFR Part 50, Appendix G,is maintained."

RESPO'NSE

Comed has performed a re-evaluation of reactor pressure vessel (RPV) beltline material chemistry values previously submitted as part of the current licensing basis and presented i the results in each Table 1 for the units. These tables include the latest known best- 1 estimate chemistry information for all beltline materials in the four Comed Pressurized Water Reactor RPVs. The best-estimate chemistries provided in these tables were obtained from the information provided in support of the Byron Station Units 1 and 2, and Braidwood Station Units 1 and 2, Technical Specification Amendment request regarding pressure temperature curves (Reference 2). References 3 through 6 list the heat specific chemistry data considered, defined the material sources for these chemistry data, and determined the appropriate method for determining the best-estimate chemistry.

Additional input information has been provided in these tables to ease understanding of the calculated pressurized thermal shock reference temperature (RTns) values aad their comparison to the current licensing basis values. These tables include the calculation of l

the adjusted reference temperatures (ARTS) to assess the impact on the licensed pressure <

temperature (PT) and low temperature overpressure protection (LTOi') limits. The j footnotes, provided with each table, identify the references and the evalaation methods used to determine the best-estimate chemistry values. )

1 Chemistry data with multiple material sources was evaluated using the means-of-sources approach. Babcock & Wilcox has determined this approach to be most appropriate for determining the best estimate chemistries. In applying this approach to a given weld wire heat, all available test results from separate and distinct test or production welds are i I

averaged. The resulting average values for separate and distinct test or production welds are subsequently averaged to obtain the best estimate value. The mean-of-the-sources approach provides the most appropriate estimate of weld chemistry. This approach eliminates the inappropriate weighting efTect which can result from widely varying numbers of analyses performed on individual weld blocks.

The requested surveillance data are presented in each Table 2 for the units. All surveillance capsule resu;ts for the heats of material in the Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2 surveillance programs are included in these tables.

Surveillance capsule chemistry data are defined from the mean of all chemistry analyses performed on the capsule source materials. The irradiation environments, i.e., irradiation temperatures, of the surveillance capsules at Byron Station Units I and 2 and Braidwood Station Units 1 and 2 were identical except for the second fuel cycle at Byron Station Unit

1. The temperature difTerence, detailed in Reference 3, for this single fuel cycle was ,

considered to have an insignificant effect on the surveillance data. As confirmed by the i NRC in Reference 7, no temperature adjustment was necessary to the surveillance capsule data.

The surveillance data credibility is evaluated for each unit and presented in each Table 3 for the units. For the surveillance weld credibility evaluations, the measured ARTmr values were adjusted for differences in the surveillance capsule chemistry values by the ratio of the chemistry factor for the mean chemistry of the surveillance capsules to the chemistry factor of each surveillance capsule. Once the surveillance data were deermined j

to be credible, the weld chemistry factor was determined from the surveillance data, which had been normalized to the vessel weld chemistry by the ratio of the vessel chemistry factor to the chemistry factor of each surveillance capsule. These evaluations were consistent with the guidance provided in Reference 8.

The nozzle shell forgings and upper girth welds between the nozzle shell and the intermediate shell coarses were included in the evaluatione provided in each Table i for the units. Although these vessel components are above the active fuel zone, the fluence values used to determine the RTers values and adjusted reference temperatures (ARTS) were conservatively estimated by using the top of active fuel fluence. Using this conservative fluence value has yielded a conservative prediction of the RTers and ART for these materials.

Examining the results presented in Table I for Byron Station Unit 1, the limiting material, based on the ART used in the PT and LTOP limits, is the intermediate shell forging, heat number SP5933. The ART values for this material are the same as those used for determining the currently licensed PT and LTOP limits. However, the RTi>rs for the nozzle shell forging, heat number 123J218, is projected to be 2 F higher than the intermediate shell forging at end-of-life. The nozzle shell forging Rhs of 82 F remains well below the 10 CFR 50.61 screening criteria of 270 F for forgings.

Examining the results presented in Table 1 for Byron Station Unit 2, the middle girth weld, heat number 442002, remains the limiting material based on the RTers and ART for the PT and LTOP limits. The ART value for this material was determined to be less than the value used in the currently licensed PT and LTOP limits.

Examining the results presented in Table I for Braidwood Station Unit 1, the middle girth weld, heat number 442011, remains the limiting material based on the RTvrs and ART for the PT and LTOP limits. The ART value for this material was determined to be less than the value used in the currently licensed PT and LTOP limits.

Examining the results presented in Table I for Braidwood Station Unit 2, the middle girth weld, heat number 442011, remains the limiting material based on the RTers and ART for the PT and LTOP limits. The ART value for this material was determined to be equal to the value used in the currently licensed PT and LTOP limits.

In summary, the chemistry values for the beltline materials of the Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2 RPVs were evaluated using the information provided in ReQrences 3 through 6. Based on this information, the best-estimate chemistry values were determined using the mean-of-the-sources approach for vessel materials with multiple chemistry data sources and the mean chemistry values for vessel materials with chemistry data from a single source. The results of this evaluation are documented in each Table I for the units. The requested surveillance capsule data from the Byron Station Units I and 2 and Braidwood each Table 2 for the units. Units I and 2 surveillance capsule programs were provided in each Table 2 for the units and evaluated in

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each Table 3 for the units. The results of the surveillance data evaluations were included in each Table 1 for the units. As seen from these tables, the ART for the PT and LTOP limits was reduced or remained unchanged for all of the RPVs. The RTns limiting material for Byron Station Unit 2 and Braidwood Station Units 1 and 2 did not change.

The Byron Station Unit 1 RTns limitirig material changed from the intermediate shell forging, RTns = 80 F, to the nozzle shell forging, RTns = 82 F. This RTns is well below the 10 CFR 50.61 screening criteria of 270 F for a forging. Consequently, the currently licensed PT and LTOP limits do not require revision.  ;

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"eferences

1. M. DeVan letters to NRC dated June 6,1997,(INS-97-2262), June 19,1997 (INS-97-2450), and July 10,1997 (INS-97-2741).

2. II. G. Stanley letter to NRC dated January 8,1998, transmitting

" Supplemental Information Pertaining to Technical Specification Amendment Regarding Pressure Temperature Curves," Byron and Braidwood Nuclear Power Stations, Units 1 and 2.

Includes:

• Errata to " Byron Unit 1 IIeatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron

& Braidwood," WCAP-14824 Revision 2, (Westinghouse Letter CAE-97-233/CCE-97-316)

• Errata to " Byron Unit 211eatup and Cooldown Limit Curves for Normal Operation," WCAP-14940, (Westinghouse Letter CAE 233/CCE-97-316)

. Errata to "Braidwood Unit 2 lleatup and Cooldown Limit Curves for Normal Operation," WCAP- 14970, (Westinghouse Letter CAE 233/CCE-97-316)

3. J. B. Ilosmer letter to NRC dated December 3,1997, transmitting

" Byron Unit I lleatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron &  !

Braidwood," WCAP 14824, Revision 2 and Errata in Westinghouse Letters CAE-97-220/CCE-97-304

4. J. B. Ilosmer letter to USNRC dated November 18,1997, i transmitting:

• " Byron Unit 2 lieatup and Cooldown Limit Curves for Normal Operation," WCAP-14940, with Errata in Westinghouse Letter CAE-97-210/CCE-97-289.

. "Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," WCAP-14970, with Errata in Westinghouse Letter CAE-97-210/CCE-97-289.

5. "Braidwood Unit 1 Ileatup and Cooldown Limit Curves for Normal Operation," WCAP-14243.

6. NRC letter to O. D. Kingsley, dated January 21,1998,"Ilyron Station, Units I and 2, and Braidwood Station, Units I and 2,"

Acceptance for Referencing of Pressure Temperature L imits Report.

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, 7. NRC Memorandum from K. Wichman to E. Sullivan, dated November 19,1997, " Meeting Summary for November 12,1997 i Meeting with Owners Group Representatives and NEI Regarding Review of Responses to Generic Letter 92-01, Revision 1 Supplement 1 Responses."

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7 Braidwood Station Unit 2: Table 3 Lower Shell Forging 50D102-1-1150C97-1-1, Single Source Credibility Determination:

Measured FF x Surveillance (Measured -

Program Capsule Chem. Irradiation Fluence ARTwoT Predicted Predicted) Measured Source ID Cu Ni Factor Temp ( F) Factor ( F) ARTuor ( F) ARTuor ( F) ARTNor '

B R-2 U 0.056 0.767 34.6 (1) 0.741 0 9.9 -9.9 0.00 0.55

, B R-2 U 0.056 0.767 34.6 (1) 0.741 5 9.9 -4.9 3.71 0.55 BR-2 X 0.056 0.767 34.6 (1) 1.033 3 13.7 -10.7 3.10 1.07 B R-2 X 0.056 0.767 34.6 (1) 1.033 35 13.7 21.3 36.16 1.07 Sum: 42.97 3.23 Surveillance Chemistry Factor: 13.3 (1) Surveillance data are from this vessel. No adjustment for irradiation tenmerature.

Credibility Criteria of 17 F in Reg. Guide 1.99 Rev. 2 for base metal is exceeded, therefore the surveillance data is not credible.

Lower Shell Forging 50D102-1-1150C97-1-1 Assess Chemistry Factor for Conservatism:

Measured -

Surveillance Measured Predicted Predicted Program Capsule Chem. Fluence ARTuor ARTuoT ARTnor @)

Source ID Cu Ni Factor Factor ( F) ( F) (2)

B R-2 U 0.056 0.767 34.6 0.741 0 25.7 -25.7 BR-2 U 0.056 0.767 34.6 0.741 5 25.7 -20.7 B R-2 X 0.056 0.767 34.6 1.033 3 35.7 -32.7 B R-2 X 0.056 0.767 34.6 1.033 35 35.7 -0.7 Reg. Guide 1.99 Rev. 2, Table chemistry factor is considered conservative oecause these differences are less than +/- 34 F.

Reference:

" Byron U it 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld MetalIntegration for Byron & Braidwood,"

WCAP-14824, Revisioi. 2, with Errata in Westinghouse Letters CAE-97-220/CCE-97-304, CAE-97-231/CCE-97-314 and CAE-97-233/CCE-97-316.

"Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," WCAP-14970.

See Braidwood Unit 1 Table 3 for Middle Girth Weld WF-562, Heat # 442011 Credibility Determination and Vessel Weld Chemistry Factor:

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