ML20212K863

Safety Evaluation Re Requirements for Protection Against Pressurized Thermal Shock Events

ML20212K863
Person / Time
Site:	Zion File:ZionSolutions icon.png
Issue date:	08/14/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20212K861	List: ML20212K856 ML20212K863
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8608250094
Download: ML20212K863 (5)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION kY RELATED TO PROTECTION AGAINST i

PRESSURIZED THERMAL SHOCK EVENTS l ZION NUCLEAR POWER STATION, UNIT I DOCKET NO. 50-295 1

FAST NEUTRON FLUENCE ASPECT By letter dated January 17, 1986, the Commonwealth Edison Company, licensee for the Zion Unit 1 plant, submitted information on the material properi.ies and the i

fast neutron fluence (E>1.0 MeV) of the reactor pressure vessel in compliance with the requirements of 10 CFR 50.61 (Reference 1). The Engineering Branch of OPWRL-B

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reviewed the submitted information and detemined the acceptable values of the material properties and chemical composition (Reference 2). The following 7

evaluation concerns the estimation of the fluence to the inside surface of the i pressure vessel for 32 effective full power years of operation and to the end of license and the corresponding values of the RTPTS*

i The neutron transport calculation was performed by Westinghouse using the bench-

marked DOT discrete ordinates code. Plant and cycle specific sources were used

- to obtain the exposure due to each fuel cycle. The SAILOR cross section set was used, which is ENDF/8-IV based, with a P 3 scattering approximation.

Results of j l the calculation are within 115% of the measured flux values at the surveillance capsule locations. The fluence calculation methodology is acceptable. In. par-f j

ticular, future low leakage cycles are conservatively estimated. The maximum value occurs at 45' in the azimuthal distribution. For the lower shell

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circumferential weld the peak values of 1.47 x 101' n/cm2 and 1.79 x 101' n/cm2 i' are applicable. They are based on 25.0 and 32 effective full power years of

! operation respectively and low leakage loadings for cycle 7 and subsequent 1

l cycles. <

The applicable equation for the estimation of RTPTS as specified in 10 CFR 50.61 is:

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1 07/08/86 l ZION FRACTURE TOUGHNESS RQMTS l osoeaccog4 080u14

! PDR ADOCK 09000?9S PDR <

p

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RTPTS= I + M +(-10 + 470 Cu + 350 Cu Ni) f Where: I = initial RT NDT

= 0'F M = uncertainty margin = 59'F Cu = w/o copper in weld WF70 = 0.35 Ni = w/o nickel in weld WF70 = 0.59 f = peak fluence in units of 1018 n/cm2 = 1,79 Then:

o,27 RTPTS = 59 + (-10 + 470 x 0.35 + 350 x 0.35 x 0.59) x 1,79

= 59 + 226.8 x 1.17 = 59 + 265.4 = 324.4*F which exceeds the applicable criterion of 300*F.

For the end of the current license i.e., 25.8 effective full power years the RT PTS value is:

o,27 RTPTS = 59 + 226.8 x 1.47 = 59 + 251.7 = 310.7'F which also exceeds the 300'F screening criterion.

Therefore, we will require that the licensee submit their plans to satisfy the

. provisions of 10 CFR 50.61.

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ZION FRACTURE TOUGHNESS RQMTS 2 07/14/86 L

, t References 4

1. Letter from G. L. Alexander, Commonwealth Edison Company to H. R. Denton, dated January 17, 1986. .

2. Memorandum from C. E. Rossi to J. Norris, " Zion Station Unit 1, Material Properties for Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,10 CFR 50.61,", dated June 24, 1986.

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l ZION FRACTURE TOUGHNESS RQMTS 3 07/14/86 t

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. 1 MATERIAL PROPERTIES ASPECT The controlling beltline material from the standpoint of PTS susceptibility was identified to be intermediata to icwer shell circumferential weld WF70 (weld wire heat number 721055 The material properties of the controlling material and the associated margin and chemistry factor were:

Utility Submittal Staff Evaluation Cu (copper content, %) = 0.32 0.35 Ni (nickel content, %) = 0.56 0.59 I (Initial RTNDT, F) = 0 M (Margin,"F) ,

= --

59 CF (Chemistry Factor, 'F) = --

226.8 Discussion The controlling material has been properly identified, but the copper and nickel i

contents given for weld wire 72105 are not acceptable. In the staff evaluation, the copper and nickel contents tabulated above were taken from a B&W weld chemistry study,* which reported more than 50 measurements on this weld wire.

! The licensee's submittal is based on a Westinghouse report,** which presented 57 values from the B&W report plus 30 additional values from recent surveillance reports and elsewhere.

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l K. E. Moore, A. S. Heller, "B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study," BAW-1799, July 1983.

• • E. L. Furchi, S. L. Anderson, K. R. Balkey, V. A. Perone, and M. A. Weaver,

" Zion Units 1 and 2 Reactor Vessel Fluence and RTPTS Evaluations," WCAP10962 December 1985.

A histogram of the copper contents listed for heat 72105 in Table B.1-4 of the licensee's submittal shows that most of tha values added by Westinghouse fall in the range 0.22 - 0.28% copper, whereas most of the data taken from BAW 1799 range from 0.28% upward. Clearly, the values added by Westinghouse constitute a different population of weld wire from heat 72105, or the difference is a con-sequence of the different method used for chemical analysis. The latter is probably the case because most of the values added to the data base by Westinghouse came from two surveillance reports which gave x-ray fluorescence measurements on irradiated broken Charpy bars - a technique for which calibra-tion is more difficult. In addition, three of the added values are from early measurements on weld qualification samples which are now known to be in error on the low side.

The large number of copper measurements result from widespread use of wire heat 72105 in studies of low upper shelf toughness, in the investigation of " atypical weld metal" (BAW10144-A, 1979) and in surveillance programs. Heat 72105 was used with one weld flux to make welds designated WF70, which was used in several vessel welds. It was used with another flux lot to make weld WF209-1, which was used as t'he surveillance weld for several plants. Copper measurements for these two weld designations do not appear to be one homogeneous population, for reasons unknown. Weld flux does not affect the copper content.

Conclusion The primary observation to be made from the studies of welds made from heat 72105 is the broad range of copper contents represented -- for example, 9 of the 87 values are in the range 0.40-0.49% copper. One standard deviation is about 0.07%. In light of these facts, it seems prudent to use the higher average values of copper and nickel content tabulated above. Based on the staff's evaluation, assuming the reported fluence values are correct, Zion 1 will rea-h the screening criterion before the end of licensed life. (The staff's evalua-tion of fluence issues will be forthcoming from L. Lois, Division of pWR Licensing-A.) Hence, the Licensee should be asked to submit an analysis of possible flux reduction programs as required by paragraph 50.61(b)(3).

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