ML20206B656

Corrected SER Re Fracture Toughness Requirements for Protection Against PTS Events (10CFR50.61).Util 860123 Submittal Re Matl Properties & Fast Neutron Fluence of Reactor Vessel Acceptable

ML20206B656
Person / Time
Site:	Rancho Seco
Issue date:	03/13/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20206B654	List: ML20206B651 ML20206B656
References
NUDOCS 8704090270
Download: ML20206B656 (4)

Text

_ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _. _ _. _ ____. _ _

_ /joaergIo UNITED STATES g.

i-r

'.g NUCLEAR REGULATORY COMMISSION 3

,w j

WASHINGTON. D. C. 20555

%...../

SAFETY EVAL'UATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS-(10 CFR 50.61)

SACRAMENTO MUNICIPAL UTILITY DISTRICT-RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 By letter dated January 23, 1986, the Sacrainento Municipal Utility District, the licensee for the Rancho Seco plant, submitted information 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 (References 1, 2 and 3).

Our evaluations of the pressure vessel material properties, and fast neutron fluence for fracture toughness requirements for protection-against pressurized thermal shock events (10 CFR 50.61) are as follows:

Material Properties The controlling beltline material from the standpoint of PTS susceptibility was identified to be a longitudinal weld in the. lower shell course, weld WF70 (Weld Wire Heat No. 72105). This submittal states that there is a small probability that the controlling weld contains " atypical weld metal,"

an off-specification material that B&W found in 1978 in a test piece welded with wire hea t number 72105. This was the subject of Topical Report BAW 10144-A, which was reviewed by the staff on December 12, 1979. Based on a limited amount of test data developed by 8&W, the submittal states that the screenino criterion will not be reached before end of life even if the atypical weld metal is present. The staff does not disagree with the B&W evaluation of this material, although their evaluation does not follow the procedure for calculating RT given in the PTS rule.

(The formula for RT was developed withoukThansideration of the atypical weld metal.) The skNf i>elieves that the probability of occurrence of atypical weld metal in vessel welds is low enough and its properties are such that the results of a plant-specific probabilistic risk analysis would not be affected significantly if the atypical weld metal was considered to be present. Thus, the materials input to the calculation of RT will be evaluated without further consideration of the atypical weld Ntal..

p The material properties of the controlling material and the associated margin and chemistry factor were reported to be:

1 r

8704090270 870313 PDR ADOCK 05000312 P

PDR

..:l./g :. l'

~

A

~2 1

Iltii iy S utune t t. I S t.: t f l'v.!lu.t t sin

~

0. n

0. h Cu (copper content. :)

0.59 0.59; Ni (nickci conterit.

)

0 0

I (Initi.at Ri

, "F)-

5'l -

-59 H '(Margin.'*F) a 720 0 CT (Chemistry Factor. "F)

Ile'3"SEiII

• E The contenlling material has been properly identified.,

g given for the Copper and nickel contents and the Ittii1a p

,heOaks

,} q5_

The margin has been derived from consideration O,.

t following the PTS P.ule. Section 50.61 of 10 CFR Part 50.

East Neutron Fluence Detailed calculations ilavq been performed with the assistance of our contracter.

orookhaven National (Mb6catory (,8NL).for the cycle dependcet orcssure'ves.

sci inner-wall (E > 10 MeV) fluxes for the Rancho Seco plant.

The calculated fluxes were used t6' determine the vessel (inner diameter) accumulated fluence, f r specific welds _ in the reactor which was then used to predict the RTPTS The analysis was based on a 00T calculation for the Rancho Seco plaat.

vessel.

An 80% load factor, low leakage loading and plant specific data were assumed.

.or the extrapolation.

The following table summarizes the comparison of the applicant and the BNL results.

The BNL estimates arc in substantial agreccent wt th the applicant values, therefore. they are acceptabic.

The RT values pyg at the expiration of the current license arc below the applicable criteria spect ficd in 10 CFR 50.61 and are acceptable.

RANCHO SECO RT CALCULATED FOR OCTOBER 11, 2008 (LICENSE EXP! RATION) p73 Up t s(b)(og )

NYTx10.is (a)

(a) tle i d lica t#

CU N[

[

M SMUD ONL Screen ( F) SMUD lint

1. 03:

NL GLI Fv4023

.15

.68 10.

48.

5.4

5. 5 210 14 0.

139.9 0.07 U Si(ELL C50621

.12

.60 4

48.

7.1

7. 2 270 118.

111.5

0. 3 U SHELL C50622

.12

.60

-10.

48.

7.1 7.2 270 104 103.5 0.43 L SHELL C50701

.10

.58

-20.

48.

7.1

7. 2 270 81.

80.43 0.71 L SHELL C50702

.10

.58

-20.

48, 7.1

7. 2 270 81.

80.43 0.71 WF.233

.29

.68 0.

59.

5.4 5.5 300 225.

225.2

-0.09 U CIRC H CIRC WF-154.

.31

.59 0.

59.

7.1

7. 2 300 241.

241.7

-0.29 l

L CIRC

.W-233

.29 68 0.

59.

5.4

.5.5 300 225.

.225.2 0.09 u tx.

W -29'g.: 2.'s.1 $ 31

0.{ 59.

6_;8.,

. 6.9 2., 70. -

193..

193.7

.-0.*2 s

e r

f, r-

= n -- _ -

n

"~

^

. pc :

- -3.

WF-29

. 23'

.63 0.

59.

6.9 7.0

-270 194.

.194.11

-0.05

?L LN

~

-L LN ID WF-70

35

.59 0.

59 "6. 9 -

7.0 270 265.

264.9

-0.04

~L LN OD,WF-29

.23

.63 0.

59.

6.9' 7.0

'270 194.

_194.1

-0.05~

a)

Ref.-l-PTS =I+M[-10+470Cu+350CuNi]f.270 0

b) RT 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 fact that the RT value is_readily available from.the calculation PTS of the Pressure Temperature limits, and (c) the staf f desire to be informed on the. current value of the RT for all PTS

PWRs, and a compari-we, request that the licensee submit a reevaluation of the.RTPTS son to-the prediction of Reference 1 along with:the Pressure-Temperature operat-ing limits which are required by 10 CFR 50 Appendix G.*' It should be noted that this reevaluation is a requirement by 10 CFR 50.61, whenever core loadings, surveillance measurements, or other information indicate a significant change in projected values.

l

,;,. f. - -

REFERENCES 1.

Letter from R. J. Rodriguez, Sacramento Minicipal Utility District, to

^

Director, NRR, dated January 23, 1986.

2.

Letter from R. J. Rodriguez to Director, NRR, dated February'13,1986, and attached report BAW-1884,:"'lessel Fluence Reduction Fuel Cycle Study,"

dated December 1985.

3.

Letter from J. F.' Walters, B&W Owners Group, to Lambros Lois, NRR, dated'-

August 13, 1986.

4.

Memorandum from D. Crutchfield to C. E. Rossi,-NRR, " Rancho Seco Material' Properties for Fracture Toughness Requirements for Protection Against-Pressurized Thermal Shock Events, 10 CFR 50.61," dated April 22, 1986.

5.

Memorandum from A. Prince, BNL, to J. Carew, BNL, " Review of the Sacramento Municipal Utility District, Ra,1cho'Seco PTS Submittal," dated November 4, 1986 (Draft).

i i

).