ML20059C340

Provides Response to Request for Addl Info Re Rev 1 to GL 92-01, Reactor Vessel Structural Integrity

ML20059C340
Person / Time
Site:	Point Beach
Issue date:	11/01/1993
From:	Link B WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-93-115 GL-92-01, GL-92-1, VPNPD-93-186, NUDOCS 9311010115
Download: ML20059C340 (12)

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231 W McNgon. PO Box 2046 MMoukea WI 5320b2046 - (414)221 2345 VPNPD 18 6 NRC 115 L November-1, 1993 1

Document' Control Desk U.S.' NUCLEAR REGULATORY COMMISSION Mail Station P1-137

Washington, DC 20555 Gentlemen, 5 DOCKETS 50-266 AND 50-301 REQUEST FOR ADDITIONAL INFORMATION BEGARDING NRC REVIEW OF THE POINT BEACH 't RESPONSE TO GENERIC LETTER 92-01, REVISION 1 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 ,

Nuclear Regulatory Commission (NRC) Generic Letter (GL) 92-01,

" Reactor Vessel Structural Integrity," dated March 6,-1992,-was ,

issued to obtain.information-from licensees.to enable the NRC to .

assess compliance with regulatory requirements and commitments- .

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regarding reactor vessel integrity. Our response to GL 92-01.was_

provided to the NRC on June 25, 1992,-and supplemented-on. July.30,

~

1992.' On August 26, 1993, the-NRC requested that additional information be provided to assist in its review of. Wisconsin .  ;

Electric's response.- The requested ~information is provided in the attachment to this letter. ,

Our response to Generic Letter 92-01 as supplemented by the attached information demonstrates our continued compliance with 10 CFR 50.60 and conformance to our commitments made in response to Generic Letter 88-11. Please contact us if-you have any questions  ;

or require additional information regarding this response. v sincerely, f-e- / J,

.Boblink Vice President Nuclear Power 010013 i Attachment  !

JRP/dah 9311010115 931101 EL PDR P

ADOCK 05000266 PDR bi gh  ;

cc:

&f NRC Regional Administrator, Region III '

NRC Resident Inspector gI d subsklary of konsirr D>ergy Girporaten t

Attachment to VPNPD 93-November l' 1993 NRC Request for Additional Information, Point Beach Unit 1:

" Question 2a in GL 92-01 The response to GL 92-01 states that the unirradiated USE of circumferential weld SA-1101 is e 70 ft-Ib. In report BA W-1803, which is cited in GL 92-01, the unirradiated value of circumferential weld SA-1101 is stated to be 65 ft-Ib. Resolve this discrepancy. "

Wisconsin Electric Response:

The unirradiated Cv USE for SA-1101 of 65 ft-Ib stated in BAW-1803 was obtained from Turkey Point Unit 3 surveillance test data. The 70 ft-lb value stated in the generic letter response is taken from Table 3-5 of BAW-1803, and is rounded from 69.7 ft-ib. This value was statistically derived from the entire population of Linde 80 welds. Regulatory Guide 1.99, Rev. 2, acknowledges the use of generic values when measured values are .E not available for the reactor vessel. The statistically derived generic value is considered to be the most representative value obtainable since it was determined using all valid and applicable information for the Linde 80 weld material.

"There is no unitradiated USE value provided for nozzle belt forging 122P23VA 1 in the response to GL 92-01. Provide an unitradiated USE value for this forging, orprovide the neutron fluence specific to this forging and demonstrate that it is not the limiting material. "

Wisconsin Electric Response:

The initial CvUSE of forging 122P237VA1 is not available, however, the initial Cv0SE for similar forgings are available and are shown in Table 1. The chemical compositions of forging 122P237VA1 and the similar forgings are provided in Table 2. (The above request for additional information identifies the nozzle belt forging as 122P23VA1; this is understood by Wisconsin Electric to be a typographical error.)

The mean value of initial CvUSE for the forgings listed in Table 1 is 152 ft-lb with a standard deviation of 20 ft-Ib. The specimens were oriented such that the break is in the strong direction. The acceptance criteria of 10CFR50, Appendix G are based on the break ,

being in the weak direction. The NRC staff acknowledges the use of 65% of the strong ,

direction value in lieu of the weak direction value where the latter is not available (NRC Memorandum, C.Z. Serpan to C.Y. Cheng, dated June 25,1990). Sixty-five percent (65%) of 152 ft-lb is 99 ft-lb. This value is well in excess of the minimum of 75 ft-lb that is required by 10CFR50, Appendix G.  ;

The maximum 32 EFPY neutron exposure on the nozzle belt forging is 3.17E+ 18 nicm2 (E> 1 MeV) [WCAP-12794, Rev. 2). This is nearly one order of magnitude less than the peak neutron exposure of the beltline material (2.43E+19). Because the nozzle belt ,

materialis a forging with inherently low copper content and'has significantly less neutron exposure than the beltline plate material, we can clearly conclude that the nozzle belt materialis not the limiting material.

Page 1 of 11

Attachment to VPNPD ,

November 1,1993

'Ouestion 2b in GL 92-01 The initialRTuor values provided in the response to GL 92-01 for certain plates (A-9811 and C-1423) and welds (SA-812, SA-847, and SA-1101) differ from the values provided in the Correction to Pressurized Thermal Shock (PTS) Submittal (original submittal dated January 20, 1986) Point Beach, Units 1 and 2 (dated March 14, 1986), as shown in the table below.

Resolve these discrepancies. "

Description IRTuor from GL 92-01 IRTuor from Correction to PTS...

Plate A-9811 1 -2 Plate C-1423 1 -20 Weld SA-812 -5 0 Weld SA-847 -5 0 Weld SA-1101 -5 0 Wisconsin Electric Response:

In response to the revision to 10CFR50.61 in 1991, Wisconsin Electric submitted an updated PTS evaluation on September 4,1992. The IRTum values provided in this latest PTS submittal and in the GL 92-01 response are compared in Table 3. As shown in Table 3, the values provided in these two submittals are the same for all materials except weld metals SA-775, SA-812, SA-847, and SA-1426. In the 1992 PTS submittal the more -

conservative generic value for IRTum provided in 10CFR50.61 was used for these materials.

The March 14,1986 PTS submittal was made before Wisconsin Electric joined the B&W.

Owners Group. In joining the B&W Owners Group, Wisconsin Electric gained access to additional information on reactor vessel materials and to additional analytical methods.

This information has allowed Wisconsin Electric to refine our beltline material property data and is reflected in the GL 92-01 response and the 1992 PTS submittal.

" Provide the unitradiated reference temperature for all welds (from Point Beach and other reactor vessels) fabricated using SA 1101 or weld wire heat number 71249. Assess the offect of this data on the reported unirradiated reference temperature and end oflicense RT,,s value for the SA 1101 weld metalin the Point Beach 1 reactor vessel. "

Table 4 lists all reactor vessel beltline welds that were f abricated using weld wire heat 71249 and the IRTum for those welds. The IRTum for all SA-1101 welds is + 10 F and for all other welds is -5'F. The + 10'F value for SA-1101 was determined by measurement as reported in EPRI NP-373. The -5 F value for all other 71249 weld materials is the statistically determined value for Linde 80 welds as reported in BAW-1803, Rev.1. The values of IRTux and EOL RTers for SA-1101 were not influenced by the other 71249 weld r materials.

Page 2 of 11

Attachment to VPNPD  ;

November 1,1993

" Question 3a in GL 92-01 i Provide for each fuelcycle, the cold leg temperatures, the neutron fluence at the limiting l weld, and the effective fullpower time. For cycles with coastdown periods, provide the i requested information during the constant power operating period and the coastdown period.

Provide the weighted average cold leg temperature for all cycles of power operation until the end of the current cycle and the value projected for end of license. The averages should be ,

weighted based on both time and irradiation embrittlement.

Indicate the cycles when each surveillance capsule was withdrawn from the reactor vessel ,

and provide the weighted average coldleg temperature to which it was exposed.

Based on the information provided above and the Point Beach surveillance data, assess the  ;

effect ofirradiation temperature on the end oflicense RTng value for the SA 1101 weld metal in the Point Beach reactor vessel ".

s Wisconsin Electric Response:

Table 5 presents the cold leg temperature, neutron fluence at the beltline circumferential weld, and effective full power time for each fuel cycle and coastdown period for Point Beach Unit 1. The table also provides the weighted average cold leg temperature based on both time and neutron embrittlement following each constant power operating period  ;

and coastdown. The table indicates when each surveillance capsule was withdrawn, and provides the weighted average cold leg temperatures to'which the capsules were exposed.

The weighting of the average cold leg temperature based on neutron embrittlement was interpreted as being based on the cumulative neutron fluence at the limiting weld. The ,

information presented in Table 5 was weighted in this manner. l In our July 30,1992 supplemental response to GL 92-01, Wisconsin Electric addressed the five periods of coastdown operation where the cold leg temperature dropped below ,

525"F. As shown in Table 5, one additional coastdown period was identified (Fuel  ;

Cycle 15); in this case the cold leg temperature was as low as 525'F. '

As stated in our response to GL 92-01, based upon analyses of surveillance materials, the I period of reduced power operation (with coincident reduced operating temperature) from  :

approximately December 1,1979 to October 1,1983 does not appear to have affected the irradiation embrittlement characteristics of the vessel materials. The additional periods of reduced power operation associated with coastdowns are very short when compared to  ;

the reduced power operation period of 1979 through 1983 and to the licensed operating life of Point Beach Unit 1. ,

Based on the properties of the materials analyzed from surveillance capsules removed before and after the period of reduced power operation, no effect is expected on the end ,

of license RTns value for the SA 1101 weld metal because of irradiation temperature. We I conclude that the short coastdown periods will not have an appreciable effect on reactor ,

vessel embrittlement. J h

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Attachment to VPNPD ,

November 1,1993 NRC Request for AdditionalInformation, Point Beach Unit 2:

" Question 2a in GL 92-01 There is no unitradiated USE value provided for nozzle belt forging 122P23VA 1 in the response to GL 92-01. Provide an unitradiated USE' value for this forging, or provide the ,

neutron fluence specific to this forging and demonstrate that it is not the limiting material. "

Wisconsin Electric Response: ,

The initial CvuSE of forging 123V352VA1 is 137 ft-lb (Bethlehem Steel Corporation, Report of Test No. 808, June 13,1968). (The above request for additional information identifies the nozzle belt forging as 122P23VA1; this is understood by Wisconsin Electric to be a typographical error.) The specimens were oriented such that the break is in the strong direction. The acceptance criteria of 10CFR50, Appendix G are based on the break being in the weak direction. The NRC staff acknowledges the use of 65% of the strong direction value in lieu of the weak direction value where the latter is not available (NRC Memorandum, C.Z. Serpan to C.Y.Cheng, dated June 25,1990). Sixty-five percent (65%) of 137 ft-Ib is 89 f t-Ib. This value is well in excess of the minimum of 75 ft-lb that is required by 10CFR50, Appendix G.

The maximum 32 EFPY neutron exposure on the nozzle belt forging is 3.70E+ 18 n/cm2 (E> 1 MeV) (WCAP-12795, Rev. 21. This is nearly one order of magnitude less than the peak neutron exposure of the beltline material (2.52E+ 19). Because the nozzle belt materialis a forging with inherently low copper content and has significantly less neutron exposure than the beltline material, we can clearly conclude that the nozzle belt materialis not the limiting material.

" Question 2b in GL 92-01 The response to GL 92-01 does not provide the following information for the nozzle belt / intermediate shell circumferential weld: heat number of wire and flux lot number used in fabrication, and chemicalcomposition in weight percent of phosphorous and sulfur. Please provide this data for the aforementioned weld. "

Wisconsin Electric Response:

The nozzle belt to intermediate shell ccurse weld was fabricated by Combustion Engineering, Inc. (CE). Information on this weld is not currently available, however Wisconsin Electric is participating in the ABBICE Reactor Vessel Group efforts to obtain weld preparation information from CE's quality assurance records.

In lieu of informa'. ion on the PBNP weld, a review of available responses to Generic Letter 92-01, Revision 1 was made for reactor vessels with Combustion Engineering welds. The phosphorus and sulfur contents of the welds were tabulated and a statistical analysis was performed of these values. The results of this statistical analysis are presented in Table 6, This information is provided as the indicator of the expected values for the nozzle belt to intermediate shell weld for Point Beach Unit 2.

Page 4 of 11

f' Attachment to VPNPD ,

November 1,1993 I

The maximum 32 EFPY neutron exposure on the nozzle belt to intermediate shell circumferential weld is 3.70E+ 18 n/cm2 (E> 1 MeV) [WCAP-12795, Rev. 21. This is nearly one order of magnitude less than the peak neutron exposure of the beltline material '

(2.52E + 19). Because the nozzle belt / intermediate shell weld will receive significantly less neutron exposure than the beltline circumferential weld, it is concluded that the nozzle belt / intermediate shell weld is not the limiting material. .

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Attachment to VPNPD .

November 1,1993 TABLE 1. REACTOR VESSEL FORGINGS NSSS Where Forging Unirrad Forging Forging Forging Specification C,USE Number Manufacturer Located (6) (ft-lb) 125P666VA1 (1) Bethlehem R. E. Ginna A508, Class 2 183 Code Case 1332-2 125S255VA1 (1) Bethlehem R. E. Ginna A508, Class 2 140 t Code Case 1332-2 123V500VA1 (2) Bethlehem Point Beach A508, Class 2 180 Unit 2 Code Case 1332-2 122W195VA1 (2) Bethlehem Point Beach A508, Class 2 145 ,

Unit 2 Code Case 1332-2 123P461VA1 (3) Bethlehem Turkey Point A508, Class 2 145 Unit 3 Code Case 1332-2 123S266VA1 (3) Bethlehem Turkey Point A508, Class 2 154 Unit 3 Code Case 1332-2 123P481VA1 (4) Bethlehem Turkey Point A508, Class 2 135 Unit 4 Code Case 1332-2 122S180VA1 (4) Bethlehem Turkey Point A508, Class 2 132 Unit 4 Code Case 1332-2 122P237VA1 (5) Bethlehem Point Beach A508, Class 2 --

Unit 1 Code Case 1332-7 NOTES: (1)

Reference:

WCAP-7254 and WCAP-10086 (2)

Reference:

WCAP-7712 and BAW-2140 (3)

Reference:

WCAP-7656 and SWRI-02-5131 (4)

Reference:

WCAP-7660 and SWRI-02-5380 (5)

Reference:

Bethlehem Steel Corporation, Report of Test No. 2133, October 19,1966 (6) Code Case 1332 allows the use of designated steels for the fabrication of Class A nuclear vessels in accordance with Section lil; this includes A508, Class 2.

Page 6 of 11 i

A_-_-___.__---__---_-_______--______.-_-. - - _ . - . _ . _ . . _ . _ _ - _ _ _ - ._ - . _ . _ . . . _ . . _ . . . . .

Attachment to VPNPD . . ,.

November 1,1993 TABLE 2. CHEMICAL COMPOSITION OF REACTOR VESSEL FORGINGS (1,2)

C Mn P S Si Ni Cr Mo Cu u er 125P666VA1 0.19 0.67 0.010 0.011 0.20 0.69 0.37 0.57 0.05 125S255VA1 0.18 0.66 O. 'O O.007 0.23 0.69 0.33 0.58 0.07 123V500VA1 0.20 0.65 0.009 0.009 0.24 0.71 0.35 0.59 0.09 122W195VA1 0.22 0.59 0.010 0.008 0.23 0.70 0.33 0.60 0.05 123P461VA1 0.20 0.64 0.010 0.010 0.26 0.70 0.40 0.62 0.06 123S266VA1 0.20 0.62 0.010 0.008 0.20 0.67 0.38 0.58 0.08 123P481VA1 0.22 0.67 0.010 0.009 0.20 0.71 0.33 0.56 0.05 122S180VA1 0.21 0.67 0.011 0.009 0.23 0.70. 0.31 0.56 0.06 122P237VA1 0.21 0.65 0.010 0.008 0.22 0.82- 'O.33 0.62- --

NOTES:

(1) Composition in weight percent (2)

Reference:

Same as noted in Table 1.

Page 7 of 11

I Attachment to VPNPD l November 1,1993  ;

j TABLE 3. INITIAL RTuor VALUES  ;

Material GL 92-01 Response 1992 PTS Submittal i identification IRTuor ( F) lRTuor ( F) 122P237VA1 + 50 + 50 A9811-1 + 1 + 1 C1423-1 + 1 + 1 SA-1426 - 5 0 S A-1101 +10 +10 t S A-812 - 5 0 SA-775 - 5 0 ,

SA-847 - 5 0 L

t Page 8 of 11

Attachment to VPNPD November 1,1993 TABLE 4. INITIAL RTuor OF WELD WIRE 71249 NSSS With W eld Material in IRTuor identification Flux Lot RV Beltline ( F)

SA-1101 8445 R. E. Ginna +10 Point Beach-1 +10 Turkey Point-3 +10  ;

Turkey Point-4 +10 SA-1229 8472 Oconee-1 - 5 SA-1769 8738 Crystal River-3 - 5 ,

Zion-1 - 5 Zion-2 - 5 1

F i

n Page 9 of 11

Attachment to VPNPD ,

November 1,1993 ' ,

TABLE 5. POINT BEACH UNIT 1 FUEL CYCLE INFORMATION '

FLUENCE AT AVERAGE CUMULATIVE - AVERAGE CUMULATIVE FUEL COLD LEG BELTLINE WELD COLD LEG TEMPERATURE COLD LEG TEMPERATURE CYCLE TEMPERATURE (*F) (E > 1.0 MeV) (n/cm') EFPY BASED ON FLUENCE (*F) BASED ON TIME ('F) COMMENTS 1 542 1.83E + 18 1.49 542.0 542.0 Capsule V Withdrawn 2 542 1.08E + 18 0.92 542.0 542.0 3 542 1.49E + 18 1.21 542.0 542.0 Capsule S Withdrawn 4 542 9.18E + 17 0.70 542.0 542.0 5* 542 8.66E + 17 0.70 542.0 542.0 5 C/D 517 1.10E + 17 0.09 541.6 541.6 Capsule R Withdrawn /Coastdown 6 542 9.60E + 17 0.81 541.6 541.6 7 542 1.06 E + 18 0.87 541.7 541.7 8 511 6.48E + 17 0.64 539.4 539.0 9 511 5.24E + 17 0.60 537.9 536.9 'i 10 511 5.30E + 17 0.65 536.5 535.0 11 511 6.11 E + 17 0.62 535.0 533.4 Capsule T Withdrawn 12' 542 7.31 E + 17 0.92 535.4 534.2 12 C/D 522 3.58E + 16 0.04 535.4 534.1 Coastdown 13 542 6.76E + 17 0.79 535.8 534.7 14* 542 6.53E + 17 0.79 536.1 535.2 14 C/D 515 4.52E + 16 0.05 536.0 535.1 Coastdown 15' 542 6.79E + 17 0.81 536.3 535.5-15 C/D 525 2.11 E + 16 0.03 536.3 535.5 Coastdown 16* 542 6.57E + 17 0.78 536.6 535.9 16 C/D 509 6.14E + 16 0.07 536.4 535.7 Coastdown 17' 542 5.23E + 17 0.74 536.6 536.0 17 C/D 511 6.60E + 16 0.09 536.5 535.9 Coastdown 18 542 5.50E + 17 0,83 036.7 536.2 19 542 5.57E + 17 0.86 !39.9 536.5 20 542 5.12 E + 17 0.79 !37.1 536.8 21 542 5.19 E + 17 0.80 537.2 537.0 Present Cycle-Projected 22-EOL 542 9.28E + 18 14.31 538.9 539.2 Future Cycles-Projected to -

End of License (EOL)

• Constant power operation period only.

Page .10 Of 11

'l 4 -- - r- . - .w a'w- - w- sem-- w -~p~4-e + iw s' -- e k - a+w-

Attachment to VPNPD ,

November 1,1993 TABLE 6. COMBUSTION ENGINEERING REACTOR VESSEL bet.TLINE WELDS STATISTICAL

SUMMARY

Sample Phosphorous, Phosphorous, Sulfur, Sulfur, Flux Type Population Mean Standard Deviation Mean Standard Deviation None given 21 0.013 0.005 0.013 0.004 ARCOS B5 Mod 2 0.016 0.000 0.017 0.000 Linde 0091 23 0.011 0.005 0.010 0.002 Linde 1092 43 0.016 0.004 0.013 0.004 Linde 124 7 0.013 0.004 0.012 0.002 All listed 96 0.014 0.005 0.012 0.004 NOTE: Composition in weight percent.

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