ML20215K962

Forwards Evaluations Supporting 860115,0402,05 & 0710 Responses to 10CFR50.61 Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,Based on 32 EFPYs

ML20215K962
Person / Time
Site:	Mcguire, McGuire
Issue date:	10/21/1986
From:	Hood D Office of Nuclear Reactor Regulation
To:	Tucker H DUKE POWER CO.
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR TAC-59962, TAC-59963, NUDOCS 8610280400
Download: ML20215K962 (6)

Text

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Docket flos.: 50-369 and 50-370 007 2 1 586 Mr. H. B. Tucker, Vice President l Nuclear Production Department Duke Power Company 422 South Church Street Charlotte, North Carolina 28242

Subject:

Reactor Vessel Fracture Toughness - McGuire Nuclear Station, Units 1 and 2 By letters dated January 15, April 2 and 5, and July 10, 1986, you submitted information for McGuire Nuclear Station, Units 1 and 2, in response to the requirements of 10 CFR 50.61, " Fracture toughness requirements for protection against pressurized thermal shock events." The information is based upon 32 effective full power years (EFPY) of operation.

The NRC staff has completed its review of these submittals and finds that the pressure vessel of each respective McGuire unit meets the toughness require-ments of 10 CFR 50.61 for 32 EFPY of operation. Our evaluation for Unit 1 is attached as Enclosure 1; Enclosure 2 is our corresponding evaluation for Unit

2. These enclosures also request that periodic future re-evaluation of RT and corresponding comparisons with the predictions of your January 15 lettE[S',

be submitted at the time your pressure-temperature operating limits required by 10 CFR 50 Appendix G are forwarded to NRC.

Contact me at (301) 492-8961 if you have questions regarding the enclosures.

Sincerely, Darl Hood, Project Manager PWR Project Directorate #4 Division of PWR Licensing-A

Enclosure:

As stated cc: See next page

. DISTRIBUTION:

1iDocket F11e. PWR#4 Rdg ACRS (10) NThomrson G.Vissing NRC~PDR~ ~ ^MDuncan JPartlow J Lanbois Local POR DHood Edordan CDSellers PRC System OGC/Bethesda BGrimes PNRandall Phk#/DPWR-A PW PWR-A PW 4 PW -A DHood/mac VDu BJYoungblood 10/y /86 10/Jl/86 10/77 /86 8610280400 861021 9 DR ADOCK 0500

3 Mr. H. B. Tucker Duke Power Company McGuire Nuclear Station cc:

Mr. A.V. Carr, Esq. Dr. John M. Barry Cuke Power Company Department of Environmental Health P. O. Box 33189 Mecklenburg County 422 South Church Street 1200 Blythe Boulevard Charlotte, North Carolina 28242 Charlotte, North Carolina 26203 County Manager of Mecklenburg County 720 East Fourth Street Charlotte, North Carolina 28202 Chairman, North Carolina Utilities Comission Mr. Robert Gill Dobbs Building Duke Power Company 430 North Salisbury Street huclear Production Departirent Raleigh, North Carolina 27602 P. O. Box 33189 Charlotte, North Carolina 28242 Mr. Dayne H. Brown, Chief Radiation Protection Branch J. Michael McGarry, III, Esq. Division of Facility Services Bishop, Liberman, Cook, Purcell Department of Human Resources and Reynolds 701 Barbour Drive 1200 Seventeenth Street, N.W. Raleigh, North Carolina 27603-2008 Washington, D. C. 20036 Senior Resident Inspector

.' c/o U.S. Nuclear Regulatory Comission Route 4, Box 529 Hunterville, North Carolina 28078 Regional Administrator, Region II U.S. Nuclear Regulatory Commission, 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 L. L. Williams Area Manager, Mid-South Area ESSD Projects Westinghouse Electric Corporation PNC West Tower - Bay 239 P. O. Box 355 Pittsburgh, Pennsylvania 15230

3 Enclosure 1 Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock, McGuire Unit 1 By letters dated January 15, April 5, and July 10, 1986, the Duke Power Corrpany, licensee for the McGuire Nuclear Station, Unit 1, submitted inforn.ation on the material properties and the fast neutron fluence (E>1.0 MeV) of the reactor pressure vessel in compliance with the requirements of 10 CFR 50.61. The in-formation provided and the following evaluation are based upon the material properties and the fluence to the pressure vessel for 32 effective full power years of operation. The corresponding value of the reference temperature for pressurized thermal shock (RTPTS) is determined and compared to screening criteria of 10 CFR 50.61.

a. Material Properties For Unit 1, the controlling reactor vessel beltline material from the stand-point of pressurized thermal shock (PTS) susceptibility is the lower longitudinal weld, Weld M1.34, weld wire heat number 305424.

The material properties of the controlling material and the associated un-certainty margin and chemistry factor as reported by the licensee and evaluated by the NRC staff are:

Licensee Submittal Staff Evaluation Cu (Copper content, %) = 0.28 0.28 Ni (Nickel content, %) = 0.63 0.63 I (Initial RTNDT, F) = -56 -56 M (Uncertainty margin, F) = 59 59 CF (Chemistry factor, *F) = -

183.3 Based on its review of material properties, the staff concludes that the con-trolling material has been properly identified. The justifications given for the copper and nickel contents and the initial RT are acceptable. The margin has been derived frorn consideration of theEses for these values, followin fluence,gEquation Section 50.61 1 of 10 ofCFR 10 CFR 50.61Part 50. Based governs, and upon the reported the chemistry values factor is as of shown above,

b. Fluence and RT

_. PTS--

The methodology of the fluence calculation by the licensee was based on the discrete ordinates code COT using SAILOR, an ENDR-B/IV based cross.section set.

The scattering is treated with a P, approximation, plant specific sources were used and the code has been benchmarked by Westinghouse. As noted above, the longitudinal weld M1.34 is the controlling material. The peak axial fluence

9  !

at- the location of weld M1.34 is the aoplicable value. The comparison of the calculated to surveillance capsule measured values shows them to be conservative by up to 27%.

Based upon its review, the staff finds that the methodology, the cross sections and the approximations used are acceptable.

The equation specified in 10 CFR 50.61, as apolicable for the McGuire Unit 1 is:

0 RTPTS = I+M(-10+470xCu+350xCuxNi)xf 27 where: I = Initial RT = -56 F M = Uncertaint((argin = 59 F Cu = w/o Copper in longitudinal weld M1.34 = 0.28

' Ni = w/o Nickel in longitudinal weld M1.34 = 0.63 f = peak aximuthal fluence for 32 effective full power years (E>102 MeV) in units of 10ig n/cm = 2.89 Therefore:

RTPTS = 247.1"F From 10 CFR 50.61, the applicable screening criterion is 270 F. The value of RT acEbtable,for Unit 1 is less than the screening criterion and, therefore, is 4

c. Future Re-evaluation In view of:

(a) the pressure-temperature updating requirements for the fracture toughness of the beltline material in 10 CFR 50 Appendix G, and (b) the lationfact thatpressu of the the RT@ temperature limitsvalue

,and is readily available from the ca (c) the RT NRC staff's desire to be infonned of the current value of the for all PWRs, PTS the staff requests that the licensee submit a re-evaluation of the RT p and a comoarison to the prediction in the letter of January 15, 1986. Thesdbmittal may be made at the tinie that the future pressure-temperature operating limits, required by 10 CFR 50 Appendix G, are submitted to NRC. This request is in addition to the re-evaluation required by 10 CFR 50.61 whenever core loadings, surveillance measurements, or other information indicate a significant change in projected values.

This request for information is covered under 0MB clearance No. 3160-0011.

S g Enclosure 2 ,

Fracture Toughness Requirements for Prctection Against Pressurized Thermal Shock, McGuire Unit 2 By letters dated January 15, April 2, and July 10, 1986, the Duke Power Company, licensee for the McGuire Nuclear Station, Unit 2, submitted information on the material prorert hs and the fast neutron fluence (E>1.0 MeV) of the reactor pressure vessel in compliance with the requirements of 10 CFR 50.61. The in-formation provided and the following evaluation are based upon the material properties and the fluence to the pressure vessel for 32 effective full power years of operation. The corresponding value of the reference temperature for pressurized thermal shock (RTPTS) is determined and compared to screening criteria of 10 CFR 50.61.

a. Material Properties For Unit 2, the controlling reactor vessel beltline materiai from the stand-point of pressurized thermal shock (PTS) susceptibility is the intermediate shell forging 05 (Heat No. 526840).

The material properties of the controlling material and the associated un-certainty margin and chemistry factor as reported by the licensee and evaluated by the NRC staff are:

Licensee Submittal Staff Evaluation Cu (Copper content, %) = 0.16 0.16 Ni (Nickel content, %) = 0.85 0.85 I (Initial RTNDT, F) ' -4 -4 M (Uncertainty margin, F) = 48 48 CF (Chemistry factor, "F) = -

112.8 Based on its review of material properties, the staff concludes that the con-trolling naterial has been properly identified. The justifications given for the copper and nickel contents and the initial RT g are acceptable. The margin has been derived from consideration of the Nses for these values, following Section 50.61 of 10 CFR Part 50. Based upon the reoorted values of fluence, Equation 1 of 10 CFR 50.61 governs, and the chemistry factor is as shown above.

b. Fluence and RT PT.S-The methodology of the fluence calculation by the licensee was based on the di crete ordinates code 00T using SAILOR, an ENDR-B/IV based cross,section set.

The scattering is treated with a P 3approximation, plant specific sources were used and the code has been beichmarked by Westinghouse. As noted above, the intermediate shell forging 05 is the controlling material. The peak axial

9 fluence is the applicable value. The comparison of the calculated to surveil-lance capsule measured values shows them to be conservative by up to 27%. Based upon its review, the staff finds that the methodology, the cross sections and the approximations used are acceptable.

The equation specified in 10 CFR 50.61, as applicable for the McGuire Unit 2 is:

RT 0 PTS = I+M(-10+470xCu+350xCuxNi)xf .27 where: I = Initial RT = -4 F M=Uncertaint[kargin = 48 F Cu = w/o Copper in intermediate shell forging 05 = 0.16 Ni = w/o Nickel in intemediate forging 05 = 0.85 f = peak axirruthal fluence for 32 effective full power years yg (E>10 2 MeV) in units of 10 n/cm, for intemediate shell forging 05 = 2.9 Therefore:

RTPTS = 194.4 F From 10 CFR 50.61, the applicable screening criterion is 270 F. The value of

, RT for Unit 2 is less than the screening criterion and, therefore, is acNhtable,

c. Future Re-evaluation In view of:

(a) the pressure-temperature updating requirements for the fracture toughness of the beltline material in 10 CFR 50 Appendix G, and l (b) the fact that the RT value is readily available from the calcu-lationofthepressuN3 temperature limits, and (c) the NRC staff's desire to be informed of the current value of the RT f r all FWRs, PTS the staff requests that the licensee submit a re-evaluation of the RT a r.d a I

comparison to the prediction in the letter of January 15, 1986. TheNbmittal may be made at the time that the future pressure-temperature operating limits, required by 10 CFR 50 Appendix G, are submitted to NRC. This request is in addition to the re-evaluation required by 10 CFR 50.61 whenever core loadings, surveillance measurements, or other information indicate a significant change in projected values.

This request for infornation is covered under CPB clearance No. 31'50-0011.

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