ML20137M018

Results of Assessment of Reactor Pressure Vessel Beltline Matls Required by 10CFR50.61 (Pressurized Thermal Shock Rule) for Maine Yankee Atomic Power Plant

ML20137M018
Person / Time
Site:	Maine Yankee
Issue date:	01/23/1986
From:	Heather Jones, Plante P Maine Yankee
To:
Shared Package
ML20137L988	List: ML20137L984 ML20137M018
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR 6956L-HFJ, NUDOCS 8601280169
Download: ML20137M018 (26)

Text

. *. MAINE YANKEE ATOMIC POWER COMPANY Results of an Assessment of Reactor Pressure Vessel Beltline Materials Required by 10 CFR 50.61 (Pressurized Thermal Shock Rule) for the Maine Yankee Atomic Power Plant January 23, 1986 Prepared by:

Paul J. Plante Staff Metallurgical Engineer Prepared by: h H . T . Jogdrt , Jr . "

Principal Nuclear Engineer Approved by: ,h.

J. R. Hebert, Manager Plant Engineering Department Approved by:  %

G.-D. Whittier, Manager Nuclear Engineering & Licensing .

8601280169 860121 e 6956L-HFJ ADOCK O hDR

.. *. MAINE VANKEE ATOMIC POWER COMPMdY ABSTRACT This report provides results of an assessment of Reactor Pressure Vessel (RPV) materials to respond to .the requirements of 10 CFR 50.61

" Pressurized Thermal Shock Rule". Included are 'alues of RT PTS for January 23, 1986 and expiration of License on October 21, 2008. The report also specifies the bases for values of RT PTS. These results demonstrate that at expiration of License, a minimum margin of 56oF exists to the Pressurized Thermal Shock (PTS) screening criterion. Furthermore, the results demonstrate that the PTS screening criterion would not be reached until 2054 A.D. If the plant were to continue operation, 46 years beyond expiration of current

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License. The United States Nuclear Regulatory Commission Staff (USNRC) concluded in Reference (1) that values of RTPTS below e PTS screening criterion present an acceptably low risk of vessel failure from PTS events.

Therefore, PTS is not a significant safety concern for the Maine Yankee Atomic Power Plant, 11 6956L-HFJ

-n - L =-

.. *. M AINE Y ANKEE ATOMIC POWER COMPANV TAELE OF CONTENTS -

Page d

ABSTRACT 11 LIST OF TAELES iv LIST OF FIGURES V 1

1.0 INTRODUCTION

1 2.0

SUMMARY

OF RESLLTS 2 3.0 REACTOR PRESSURE VESSEL DRAWING AND WELD MAP 3-4.0 BEST ESTIMATE PRGJECTIONS OF FAST EUTRON FLUENCE .6 4.1 Basis for Projections of Fast Neutron Fluences 6 4.2 Assumptions About Core Loadings for Future Operation 7 5.0 PMTERIALS PROPERTIES 9 5.1 Initial RTtOT Values 9 5.2 Chemistries 9

[ 5.3 RT PTS Values for January 23,1986 and 9-Expiration of License i 5.4 Time to Reach Screening Criterion 10

6.0 CONCLUSION

13 i

7.0 REFERENCES

14 APPEPOICES A.- Excerpts from Maine Yankee Cycle 9 Core Performance A-1 Analysis Report-

8. Results of Chemistry Tests Made on Samples of Welds B-1 i

111 6956L-HFJ

. *. MAINE YANKEE ATOMIC POWER COMPANY LIST OF TABLES Number Title P_aage, 5.1 INITIAL DATA FOR REACTOR VESSEL PLATES Am ELD 11 SEAMS FOR THE MIE YAWEE ATOMIC POWER PLANT 5.2 RT DATA ON JA N 23, 1986 A@ NINIM 12 PTS 0F LICENSE FOR BELTLIE PLATES AND WELDS FOR THE MIE YANKEE ATOMIC POWER PLANT B-1 CALCULATION OF BEST ESTIMTE FROM CHEMISTRY B-2 TESTS MDE ON SAWLES OF WELDS USING MIL B-4 MODIFIED' WELD WIRE (' HEATS 13253 and 12008) AND LIl40E 1092 FLUX (LOTS, 3774, 3791, 3833).

B-2 CALCULATION OF BEST ESTIMTE FROM CHEMISTRY B-3 TESTS MDE ON SAWLES OF WELDS USING MIL B-4 MODIFIED WELD WIRE (HEAT IP 3571)

AND LINOE 1092 FLUX (LOT 3958) iv 6956L-HFJ

.. ** MAINE YANKEE ATOMIC POWER COMPANV LIST OF FIGURES Number Title Py 3.1 Section Drawing Depicting Material and Weld 4 Seam Identification for the Maine Yankee.

Reactor Pressure Vessel 3.2 Maine Yankee Reactor Pressure Vessel Map 5

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4.1 Maine Yankee Atomic Power Plant Maximum Fast 8 Neutron Fluence (E greater than 1.0 MeV.)

Versus Year of Operation v

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. . MAINE YANKEE ATOMIC POWER COMPANY. ,

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1.0 INTRODUCTION

This report provides results of an assessment of RPV materials to respond 4

to the requirements of 10 CFR 50.61(b)(1). This includes projected values of RTPTS (a e 1mer vessel surface) of W W11m matedals for January 23, 1986, and for expiration of License (October 21,2008) as well as the date when the critical weld reaches the PTS screening criterion.

This report also specifies the bases for the projection. including the following:

1 (a) ' Fluence values for materials at specified times during plant lifetime

) [the' values of "f" specified in 10 CFR 50.61(b)(2)(iv)],

3-(b) The bases for the fluences for future operation, including assumptions about core loadings, required by 10 CFR 50.61(b)(1),

(c) Initial reference temperatures (Initial RT WT values) of the

j. unirradiated materials [the value of "I" specified in 10 CFR f 50.61(b)(2 )(1)],

(d) The chemistries or best estimate weight percent copper and nickel in the materials [the values of "Cu" and "Ni" specified in 10 CFR i 50.61(b)(2)(iii)],

(e) The equation specified in 10 CFR 50.61(b)(2) used to determine RT PTS for each. material, and (f) Margin added to cover uncertainties [the values of "M" specified in 10 CFR 50.61(b)(2)(11)].

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• . MAINE Y NKEE ATOMIC POWER COMPANY 2.0

SUMMARY

-0F RESULTS The results of this assessment demonstrate that at expiration of License, a minimum margin of 56 F exists to the PTS screening criterion.  ;

Projection beyond expiration of current License indicate that the PTS screening criterion would not be reached until approximately 2054 A.D.,

assuming continued plant operation. Tnis is the time the circumferential weld seam between the middle and lower shells (9-203) reaches 3000F, Furthermore, the results indicate the next material projected to reach the

, screening criterion is plate D-8407-2 in approximately 2080 A.D.

The USNRC staff concluded in Reference (1) that values of RTPTS I*

- the screening criterion present an acceptably low risk of vessel failure i from PTS events. Therefore, PTS is not a significant safety concern for i

the Maine Yankee Atomic Power Plant.

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. .. MAINE YANKEE ATOMIC POWER COMPANY i-J 3.0 REACTOR PRESSURE VESSEL ORAWING AND WELD MAP A section drawing depicting the material and weld seam identification of the Maine Yankee fFV is provided in Figure 3.1. A vessel weld map is provided in Figure 3.2. Included in the vessel weld map is the outline of the active region of the core which defines the beltline of the RPV.

I There are six plates within the beltline (D-8406-1, 2, and 3; and D-8407-1, 2 and 3). These plates are joined by seven weld seams (2-203 A, B, C; 3-203 A, B, C; and 9-203). Subsequent sections of this report will concentrate on these materials.

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FIGURE 3.1 SECTION DRAWING DEPICTING MATERIAL AND WELD SEAM IDENTIFICATION FOR THE MAINE YANKEE REACTOR PRESSURE VESSEL REACTOR VESSEL BELTLINE MATERIALS NOT SHOWN ,- -

INTERMEDIATE SHELL WELD SEAM No. 2-203C NggN LOWER SHELL -

WELD SEAM No. 3-203B WELD SEAM No. 3-203C e Q Q PLATE No. D8407-1 s s

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INTERMEDIATE SHELL y LONGITUDINAL WELD s ky INTERMEDIATE TO LOWER SHELL GIRTH SEAM SEAM No. 2-203A WELD No. 9-203 INTERMEDIATE SHELL N PLATE No. C8406-1

\ " LOWER SHELL LOWER SHELL PLATE -

No. D8407-2

' PLATE No. D8407-3 LOWER SHELL \

LONGITUDINAL WELD ' h SEAM No. 3-203A

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4.0 BEST ESTIMATE PROJECTIONS'0F FAST PEUTRON FLUENCE Best estimate projections.of fast neutron fluences (E greater than 1.0 MeV.) of the RPV are provided in Figure 4.1. These values correspond to the clad-base metal interface on the inside surface of the vessel where the materials receive the highest fluence. Included are projections for

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both (a) plates and circumferential weld seam 9-203, and (b) longitudinal weld seam 2-203 A, B, C and 3-203 A, B, C. The projections of fluence for i the plates and circumferential weld seam 9-203 correspond to the peak vessel fluence to which a portion of the material is exposed. The projections for the longitudinal weld seams corresponds to the fluence 10 degrees from the peak fluence location (location opposite core flats) to which one of three seams (A, B or C) are exposed. The remaining two seams for each type weld have somewhat lower fluences.

The projection of flue ~nces provided in Figure 4.1 include consideration of the flux reduction program committed to in Reference (2). These measures

] were demonstrated in Reference (3) to be adequate to meet the screening-criterion at expiration of License.

4.1 Basis for Projections of Fast Neutron Fluences The basis for projections of fast neutron fluences used in this assessment is the result.of an analysis of #ast neutron flux (E greater than 1.0 MeV.) levels in the RPV performed by Westinghouse Electric Corporation (W). These results are reported in Reference (4) and were provided to the' USNRC in Reference (2).

The W results were used in previous PTS submittals', most recently Reference (3). Changes include reduction in the normalization factor from 1.28 to 1.23 and use of estimated Flux Reduction Factors (FRFs) achieved in the final designs of Cycles 7 and 8. The justification for the reduced normalization factor is provided in Reference (3). The estimated FRFs achieved in design of Cycle 7 and 8 are provided in Section 4.10 of.

Reference (5). Excerpts from Reference (5) are provided in Appendix A.

The FRFs assumed in Cycles 9 and all subsequent cycles correspond to the target FRFs provided in Reference. (2).

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. MAINE VANKEE ATOMIC POWEQ COMPANY 4.2 Assumptions About Core Loading for Future Operation The projection of fast neutron fluences provided in Figure 4.1 assume the target FRFs provided in Reference (5) are met or exceeded for Cycle 9 and all subsequent cycles. A comparison of estimated FRFs for Cycle 9 to the target FRFs is provided in Section 4.10 of Reference (5). Excerpts from Reference (5) are provided in Appendix A. Results of this comparison indicate that~ the target FRFs are exceeded by a significant margin.

Similar results have been-obtained in fuel management scoping calculations for future cycles. Therefore, the projections adequately envelope future operation. We will continue to assess the effectiveness of flux reduction

. measures in future cycles to assure the target FRFs are achieved on the average.

-6956L-HFJ

FIGURE 4.1

- Maine Yankee Atomic Power Plant Maximum Fast Neutron Fluence (E>1.0 MeV)

Versus 4.0 ' Year of Operation m

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Fluence (Year) = A + B (Year - 1985)

,' Where A and B are provided below.

' Coeffi'cients 1.0 < f Legend:

,' A B f All Plates & Weld 9.203 0.588 0.0388 s

/ - - - Welds 2-203A, B, & C & 0.475 0.0288

' 3-203A, B, & C 0.0 .

1990 2000 2010 2020 2030 2040 2050 2060 A.D.

End-of-Calendar Year

, MAINE YONKEE ATOMIC POWER COMPONY 5.0 MATERIALS PROPERTIES l 5.1 Initial RTmT. Values A summary of initial RTET values for all plates and weld seams in the beltline is presented in Table 5.1. The initial RT mT values for plate materials were determined by testing and are in accordance

. with Reference (6). The generic mean values of -56oF corresponding

to Linde 124 and 1092 fluxes were used for all weldments.- Additional ,

information on the basis for the initial RTmT values is provided in Table 5.1.

5.2 Chemistries A summary of best estimate copper and nickel chemistries for plates and weldments is shown in Table 5.1. Sources for chemistry values are also provided in Table 5.1. The guidelines used for estimating chemistries are in accordance with the order of preference presented in Appendix E of Reference (1). Values which have changed since earlier PTS submittals are for weld seam 1-203 A, B, C; 8-203;

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2-203 A, B, C; 9-203; and 3-203 A, B, C. The values for weld seams l-203A, B, C and 3-203 A, B, C are best estimates derived from eight

. tests on samples from weldments made with identical weld wire heats and similar flux lots. Table B-1 in Appendix B lists values and -

sources of each .of these four tests. The values for weld seams 8-203 and 2-203 A, B, C have been tocained from samples of weldments made with identical weld wire heats and flux lots. The value for 9-203 is the mean of seven tests made on samples from weldments using l identical weld wire heats and flux lots. Table B-2 in Appendix B -

lists the values and sources of each of these seven tests.

5.3 RTPTS Values for January 23, 1986 and Expiration of License A summary of RT values for plates and weld seams, in the high PTS

fluence regions of the RPV beltline ~, is given in Table 5.2. Values are shown for January 23, 1986 and at expiration of License.

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.J '. MAINE YANKEE ATOMIC POWECD COMPAPdY The limiting materials of the vessel with respect to PTS

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considerations are weld seam 9-203 and plate D-8407-2 since they have the least margin to the PTS screening criterion at expiration of License. However, the value of RT PTS for these materials are still 560F below the screening criterion.

5.4 Time to Reach Screening Criterion Calculations of .the time for the critical material to reach the screening criterion have been made and show that circumferential weld seam 9-203 will have an RTPTS of 300oF in the year 2054 A.D.

Furthermore, calculations indicate the next material projected to reach the screening criterion is. plate D-8407-2 in approximately 2080 A.D.

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. M AINE YANKEE QTOMIC POWER COMPANV TABLE 5.1 INITIAL DATA FOR REACTOR VESSEL PLATES AND WELD SEAMS FOR TE MAIE YAWEE ATOMIC POWER PLANT PLATE OR WELD SEAM WELD WIRE / FLUX TYPE INITIAL RTNDT  % Cu  % Ni D-8405-1 N/A 0 F( ' 4) .17( } .51 D-8405-2 N/A 40oF(I' 4) .17(4) .54(4 )

D-8405-3 N/A 200F(I' 4) .17 E#) .58(#)

l-2034,B,C B4-MOD /1092 -56oF(2) .22(10) .84 8-203 B4-MOD /1092 -56oF(2) .21(5) ,74(5)

D-8406-1 N/A -10oF( ) .15(6) .59(6)

D-8406-2 N/A OoF(I' 4) .17(6) .56(0}

D-8406-3 N/A OoF(I' 4) .12(6) .62(6) 2-203A,B,C B-4/124 -560F( } .17( ) .17 9-203 B-4 MOD /1092(') -560F(2) ,31(8) ,74(8)

D-8407-1 N/A -20 F(I' #)- .24(6) .62(6)

D-8407-2 N/A 2 F( ' 4) .23(6) .62(6)

D-8407-3 N/A 00F(I' 4) .13(6) .65(6) 3-203A,B,C B-4 MOD /1092 -56oF(2) .22(10) .84

NOTES

(1) Source: Reference (6).

(2) Generic mean value per 10CFR50.61(b)(2)(1).

(3) Maine Yankee surveillance capsule data: Identical base metal plates.

(4) Source: Reference (7).

(5) Cooper surveillance capsule data: Identical wire (Heat 20291)/ flux (Lot 3833).

(6) Source: Reference (8).

(7) Fort Calhoun head weldment sample: Identical wire (Heat 51989)/ flux (Lot 3687).

(8) Best Estimate: Mean of 3 CE tests, 3 Battelle tests and 1 Westinghouse test of identical wire (Heat IP 3571)/ flux (Lot 3958). -See Appendix B for basis.

(9) A Mil B-4 type weld wire (Heat 33A277) and Linde 0091 type flux (Lot 3922) was used for part of 1 day out of 11 days total for welding. -The weld is conservatively assumed to be entirely Mil B-4 MOD /Linde 1092 weld wire / flux type.

(10) Best estimate mean of eight tests of identical wire (Heats 13253 and 12008)/ flux (Lot 3774, 3791, 3833). See Appendix B for Basis.

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, TABLE 5.2 ~. ,

4 RTpys OATA ON JANUARY 23, 1986 Am EXPIRATION OF' LICENSE FOR BELTLIE PLATES AND WELOS FOR THE M4IE YANKEE ATOMIC POWER PLANT (1)(2)(3) .-

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PLATE OR INITIAL FLUENCE (n/cm2)x1019(6) JAN. 23, 1986 FLUENCE (n/cm2)x1019(6) EOL(7) RTPTS SCREENING WELD SEAM RTNOT.(4) M(5) (JAN. 23, 1986) RTpys EOL(7)- RTPTS CRITERIA (9)

D-8406-1 -100F 48oF .59 ll70F 1.47 1390F 2700F' t

D-8406-2 OT 480F .59 138oF 1.47- 162oF 270 F-D-8406-3 00F 48oF .59 llloF 1.47 1280F 270oF 2-203A,B,C --56*F 590F .48 690F 1.13. 86*F- -270oF

, 9-203 -56 F 59 F .59 190oF 1.47 ~2430F(8) 3000F D-8407-1 -20oF 480F .59 162oF 1.47 200oF 2700F

.~D-8407-2 20F 480F .59 1780F 1.47 214 F 270oF D-8407-3 00F 48 F .59 118*F 1.47 137*F 270 F

' 3-203A,B,C -56 F 590F .48 132 F 1.13. 1670F 2700F NOTES:

.(1) Plates D-8405-1, 2, 3 and welds 1-203A,B,C and 8-203 are not included because fluence in these areas is significantly-

' lower than the vessel beltline and they do not exhibit significant shifts in RT mT *

(2) See Table'l for initial chemistry data. .

0.

(3) Equation 1 [RTPTS = I + M + (-10 + 470 Cu + 350 Cu N1) f .270] from 10CFR50.61(b)(2) is always used to calculate RT PTS *

, (4) See .-Table 1 for source of initial RTET *

. (5) Margin from 10CFR50.61(b)(2)(li).

, (6) Fluence ' values are calculated from the following equations based on a 75% availability factor, 19 2 f (a) Plates and circumferential welds. F(x 10 n/cm ) = 0.588'+ 0.0388 (YEAR - 1985).-

(b) ."Wial (longitudinal) welds - 10o from flats. F(x 10 n/cm 19

) =2 0.475 + 0.0288 (YEAR - 1985).- ,

(7) Expiration of License - 10/21/2008.

(8) Critical weld 203. Screening criterion (300F) is reached in 2054.

, (9) Screening criteria from 10CFR50.61(b)(2).

s . MAINE YANKEE ATOMIC POWER COMPANV

6.0 CONCLUSION

S-

The USNRC Staff concluded in Reference (1) that values of RTPTS IU" the Pressurized Thermal' Shock (PTS) screening criterion present an acceptably low risk of vessel failure from PTS events. Results of the assessments provided in Section 5.0 demonstrates that PTS screening criterion would not be reached until 2054 A.D. if the plant continued operation, 46 years beyond expiration of the current License. Therefore,

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j PTS is not a significant safety concern for the Maine Yankee Atomic Power Plant.

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s *. MAINE YANKEE ATOMIC POWER COMPANV ,

7.0 REFERENCES

1. " Policy Issue for Pressurized Thermal Shock (PTS)", SECY-82-465,~

November 23, 1982.

2. MYAPCo Letter to USNRC dated April 22, 1983 (MN-83-76).

3. MYAPCo Letter to USNRC dated June 1,- 1984 (t94-84-88).

4. S. L. Anderson, " Analyses of Fast Neutron Flux Levels for the Maine Yankee Reactor Pressure Vessel", November,1981.

5. " Maine Yankee Cycle 9 Core Performance Analysis", YAEC-1479, April, 1985.

'6. USNRC Standard Review Plan - Branch Technical Position MTEB 5-2, NUREG-0800, Rev. 1, 1981.

7. " Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA's with Loss of Feedwater for the Combustion Engineering NSSS",

CEN-189,~ December 1981.

8. " Summary Report on Manufacture of Test Specimens-and Assembly of Capsules for Irradiation Surveillance of Maine Yankee Reactor Vessel-Materials", CEtPD-37, December 30, 1971.

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s MAINE YANKEE ATOMIC POWER COMPANY d

APPENDIX A Excerpts From Maine Yankee Cycle 9 Core Performance Analysis Report YAEC-1479 (Reference 5)

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. l 4.10 Pressure Vessel Fluence A program for reduction in pressure vessel fluence has been in place for Maine Yankee since Cycle 7 to limit the potential for Pressurized Thermal Shock (PTS) concerns. The Cycles 7, 8, and 9 core desis 7 have been a progression of lower leakage loading patterns with particular em'hasis p on reduced fluence in the area of the critical longitudinal weld, which is positioned at 10 degrees from a perpendicular line to the core shroud flats.

The core shroud flats'are the core boundary lines defined by assembly numbers-1 and 2 (or 45 and 54) in Figure 3.2. The O to 10 degree region is the high '

fluence area.

The program for fluence reduction has been detailed in (59) and (60),

with target fluence reductions for Cycles 7 and 8 and subsequent cycles relative to the Cycle 6 fluence level as a reference. The Cycle 6 out-in fuel A-2

management provided relative fluences in the 0-10 degree region which were similar to the fluence history accumulated from Cycles 1, 1A, and 2 through 5.

The fluence reductions, expressed as flux reduction factors relative to the Cycles 1 through 6 fluence history, are shown in Table 4.17. The target fluence reductions in (59) for Cycles 7, 8, and 9 are compared to the actual core design fluence reductions obtained by a view-factor weighting technique of the~ average quarter-assembly powers calculated for the cycles. The result is that the cumulative fluence reduction factor target to end-of-Cycle 9 has been achieved for both the 0 and 10 degree azimuthal angles. At the critical longitudinal weld at 10 degrees, the target cumulative fluence reduction is 14% relative to the case in which no fluence reduction measures were instituted. This is achieved by a cycle 9 flux reduction factor in excess of 50%. Similar flux reduction factors are expected for future cycles to meet the targets set forth in (59).

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TABLE 4.17

. MAINE YANYEE CYCLES 6-9 RELATIVE PRESSURE VESSEL FLUENCE COMPARISONS Flux Reduction Factors at Azimuthal Angle from Perpendicular to Core Shroud Flats Total Effective Full-Power ------ 00 -------- ------ 100 --------

Cycles Years to EOC* Target Designed Tartet Designed 1-6 6.51 1.00 1.00 1.00 1.00 7 7.50 1.02 1.05 1.28 1.21 8 8.55 1.35 1.42 1.51 1.43 9 9.72 1.35 1.55 1.51 1.56 Future Cycles -- 1.35 --

1.51 --

Cycle 1-9 Average 9.72 1.08 1.12 1.14 1.14

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s - M AINE YANKEE ATOMIC POWER COMPANY APPENDIX B Results of Chemistry Tests Made on Samples of Welds B-1 69%L-HFJ

MAINE YANKEE ATOMIC POWER COMPANY 1

i TAELE B-1 CALCULATI'ON OF BEST ESTI MTE FROM CHEMISTRY TESTS

MADE ON SA W LES OF WELDS USING MIL B-4 MODIFIED WELD WIRE (HEAT 13253 and 12008) AND LINDE 1092 FLUX (LOTS 3774, 3791, 3833) i Test Source (1) Wire / Flux  % Cu  % Ni

1' DC Cook 1, SC 13253/3833- .27 .74

2 Salem 2, SC ' 13253/(3) .23 .71 3 Fort Calhoun 13253/3791~ .14 .73 4 Test Sample i

Mean .21 .73 4 CE Weld Qual (2) 12008/(3) .23 .95 Overall Mean .22 .84 i

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! (1) SC - Surveillance Capsule (2) Mean of five weld deposit records of tandem arc welds in which second weld wire heat copper content is known.

(3) Flux type - Linde 1092.

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. MAINE YANKEE ATOMIC POWER COMPANV i ,

l 4 TABLE B-2 j . CALCULATION OF BEST ESTImTE FROM CFEMISTRY TESTS MDE ON SAWLES OF WELDS USING MIL'B-4 MWIFIED WELO WIRE (HEAT IP 3571) AND LIFOE 1092 FLUX (LOT 3958)

' Test Source ( Conducted By  % Cu  % Ni 1 Weld Qual. Combustion Eng. 0.40 0.82 2 Weld Qual. Combustion Eng. 0.37 0.75 3 Plant A, SC Westinghouse 0.20 0.77 I 4 W , SC Combustion Eng. 0.36 0.78 5 W , SC Battelle Labs 0.25 0.66 i 6' W , SC Batte.Ile Labs 0.25 0.70 i 7 W, SC (2) Battell,e Labs 0.33 0.71 Mean 0.31 0.74

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