ML20113C603

Forwards Final Response to GL 92-01,Rev 1,Suppl 1, Rv Structural Integrity

ML20113C603
Person / Time
Site:	Wolf Creek
Issue date:	06/25/1996
From:	Muench R WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ET-96-0044, ET-96-44, GL-92-01, GL-92-1, NUDOCS 9607010371
Download: ML20113C603 (8)

Text

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i W@) LF NUCLEAR CREEKOPERATING Richard A. Muench W:e President Engineering June 25, 1996 ET 96-0044 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, D. C. 20555

Reference:

1) NRC Generic Letter 92-01, Revision 1, Supplement 1:

l Reactor Vessel Structural Integrity, dated May 19, 1996 l 2) Letter ET 95-0084 dated August 17, 1995, from l R. C. Hagan, WCNOC, to USNRC

3) Letter dated May 11, 1994, from W. D. Reckley,

! USNRC, to N. S. Carns, WCNOC

4) Letter WM 94-0108, dated August 25, 1994, from N. S. Carns, WCNOC, to USNRC

Subject:

Docket No. 50-482: Final Response to Generic Letter 92-01, Revision 1, Supplement 1 1

! r l Gentlemen:

I i i Reference 1 requested all Licensees of operating plants to provide additional {

information with respect to the analysis of the structural integrity of their reactor pressure vessels. Wolf Creek Nuclear Operating Corporation (WCNOC) provided an interim response (Reference 2) to that request. Reference 2  !

indicated that WCNOC would be . working with the Westinghouse Owner's Group to ,

obtain some of the requested information, and would provide a final response i by July 1, 1996. The attachment provides WCNOC's final response to Reference 1.

If you have any questions concerning the above issues, please contact me at (316) 364-8831, extension 4034, or Mr. Terry S. Morrill at extension 8707.

Very truly yours, 9607010371 960625 cy,gpt f

PDR ADOCK 05000482

, p PDR Richar A. Muench RKi/jra Atcachment  ;

cc: L. J. Callan (NRC), w/a 4

W. D. Johnson (NRC), w/a (

J. F. Ringwald (NRC), w/a f

J. C. Stone (NRC), w/a P.O. Box 411/ Burhngton. KS 66839 / Phone. (316) 364-8831 An Equal Opportunity Employer M F'HC/ VET 1

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4 STATE OF. KANSAS )

) SS COUNTY OF.COFFEY )

Richard A. Muench,'of lawful age, being first duly sworn upon oath says that }

he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the. content thereof; that he , has executed that same for and on behalf of said Corporation with full.

power and authority to do so; and that the facts therein stated are true and correct to tce best of his knowledge, information and belief.

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By Richard A. Muench Vice President ] '

Engineering M day of MMe.

SUBSCRIBED and sworn to before me this , 1996.

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- w in>} Yt uxl-ANGELA E. WESSEL Notary public '

f Notary PttS State of Kansas My Aest. Exefres M/0J/9 9  ;  ;

Expiration Date d'/!d f!9 7 I l

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AttjAhment to ET 96-0044 Page 1 of 6 -

Wolf Creek Nuclear Operatino Corporation (WCNOC) Review i of Generic Letter 92-01, Revision 1, Succlement 1 Reference 1 required Licensees to provide information for four areas. Below is WCNOC's final response to the four required actions:

t Required Action l' i

"(Provide) a description'of those. actions taken or planned to locate all data l

relevant to the determination of RPV integrity, or an explanation of why the l existing database is considered' complete as previously submitted;" i WCNOC's Response to Required Action 1 ,

WCNOC provided preliminary information addressing Required Action' 1 in '

References 2 and 4. '

Those preliminary responses are updated and presented below. -

- NUREG-1511 Review WCNOC stated in Reference 2 that there was a discrepancy between NUREG-1511, j

" Reactor Vessel Status Report," and WCNOC databases concerning the '

identification of the limiting beltline material for the Wolf Creek Generating j Station (WCGS) reactor vessel, and- that we would provide the correct j information in our final response. WCNOC his since determined that the i discrepancy was caused by applying different regulatory positions on the shell.

materials in the reactor vessel beltline region. I Appendix A to WCAP-10015 outlines the selection _ criteria for the WCGS reactor 'i I

- pressure vessel beltline material. Based ' on the initial RTm, chemical composition (copper and phosphorus) and the end-of-license (EOL) neutron fluence, vessel lower shell plate R2508-3 and the intermediate to lower shell girth weld (E3.16) are predicted to have the highest EOL RTm using the methods of Regulatory Guide 1.99, Revision 1. These two materials were, therefore, considered to be the limiting vessel beltline region materials and have been used in the reactor vessel surveillance program.

However, the beltline material surveillance capsule results were reviewed again using the definitions outlined in 10CFR50. 61 (b) (2 ) . and (b) (3) (i) 'for Pressurized Thermal Shock ' (PTS) , 10CFR50, Appendix G. IV. A.1 for Upper Shelf Energy (USE), and Regulatory Guide 1.99, Revision 2. Based on this review,  !

lower shell plate R2508-1, instead of R2508-3, represents the limiting beltline material for both the PTS and USE data. These surveillance capsule results were used in predicting transition temperature shift and USE changes j due to irradiation. As a result, vessel lower shell plate R2508-3 is now i

. predicted to have a less limiting value for EOL RTm and USE than lower shell plate R2508-1. Therefore, the WCGS limiting beltline region material has been changed from lower shell plate R2508-3 to lower shell plate R2508-1. i i

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• Attadhment to ET 96-0044 Page 2 of 6 1

Combustion Encineerina Materials Certification Reports (CEMCRs) Review

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In Reference 2 WCNOC indicated that a preliminary review had been performed of 1 the CEMCRs'of the chemical composition of the WCGS reactor vessel plate and '

weld material. WCNOC has since completed the review of these CEMCRs. This review determined that the proper values for the unirradiated USE were those obtained through the initial work of Combustion Engineering (CE) done in 1975 through 1977 in accordance with ASTM SA-370.

Reference 3 requested WCNOC's evaluation of the WCGS data associated with [

Generic Letter 01, Revision 1. WCNOC responded to this request in Reference 4. However, based on recent reviews of this evaluation WCNOC has determined that we initially misinterpreted the WCGS raw data and inadvertently, in Reference 4, provided incorrect USE data.

WCNOC has identified the following three misinterpretations made in evaluating the raw data reported on in Reference 3:

1. The first misinterpretation appears to be a typographical error for the CE unirradiated USE value for R2508-3. The CE unirradiated USE value for R2508-3 of 89 ft-lbs should have been ,

86 ft-lbs as was noted in Reference 3. See Table 1 below.

2. The second misinterpretation was .the inclusion of six Charpy .

values in the minimum average of the raw data instead of only the  :

initial three Charpy data sets specified by ASTM SA-370 (1992),

Section 26.1, when the percent shear was 100%. This occurred for only one of the raw data sets (G2.06) and resulted in the initiating value differing by 1 ft-lb from the correct value (149 .

vs. 150) as is shown below in Table 1.

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3. The third misinterpretation for the E3.16 data is attributed to a minor rounding error associated with the data. The correct initial Charpy value differs by only 1 ft-lb from the value l reported in Reference 4. The average for the data shown (96.67 '

ft-lbs) was rounded to 97 ft-lbs. This is also shown in Table 1 ,

below. l i

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l Attachment to ET 96-0044 Page 3 of 6 )

Table 1 Combustion Engineering Raw Data I Beltline Reference 2 CE Raw Data Percent CE Unirradiated CE Unirradiated Identification Data for Unitradiated Shear USE Data with USE Data with

! Unirradiated USE  % Average of Six Correct I

USE ft-lbs ft-lbs Average ofinitial ft-Ibs Three I ft-lbs l R2508-3 89 93 100 85 86 ]

79 100 l 86 100 l 85 100 I 80 100 I j- 87 100 l

Axial Weld 149 151 100 149 150

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l G2.06 150 100

! 148 100 148 100 155 100

j. 145 100 ,

l Cire. Weld 98 99 100 97* 97 l E3.16 96 100

!. 95 100

• Only three (3) 100 % shear data sets were attained.

Sister Plants for WCNOC In Reference 2 WCNOC stated that we would attempt to ' identify any ~ sister  ;

plants that have the same material as WCGS reactor vessel limiting weld or plate. WCNOC's review has concluded that no other plants meet this criterion.

Updated WCNOC Databases i

WCNOC has updated the WCGS Summary Files for RTns and USE (Tables 2 and 3 at ,

l - the end of this attachment), based on'a higher EOL neutron fluence projection

' that accounts for the effects of the power uprate implemented early in 1994. f Note that the EOL RTns values'for plates, forgings and axial weld material are rauch smaller than the 270'F screening criterion specified by the PTS rule ,

(10CFR50.61), with the closest value being 109 . 6'F .. for lower shell plate  !

t R2508-1. The EOL RTns is likewise even smaller ( 31. 4*F) than the 3 0 0'F i I

screening criterion specified by the PTS rule (10CFR50.61) for circumferential weld material. Note also that the lowest 1/4 T Charpy USE (68'F) for lower

(. shell plate R2508-1 remains 2 50 ft-lb throughout the projected life of the i vessel (EOL). The notes for Tables 2 and 3 also provide other information that helps explain WCNOC's methodologies.

l Required Action 2 5

"(Provide) an assessment of any change in best-estimate chemistry based on '

I consideration of all relevant data:"

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• Attachment to ET 95-0044 Page 4 of 6 WCNOC's Response to Required Action 2 All relevant data have been considered and no changes in best estimate chemistry have been identified.

Required Action 3

"(Provide) a determination of the need for use of the ratio procedure in accordance with the established Position 2.1 .of Regulatory Guide 1.99, Revision 2, for those licensees that use surveillance data to provide a basis for.the RPV integrity evaluation;"

WCNOC's Response to Required Action 3 WCNOC utilized the ratio procedure of the subject Regulatory Guide as a basis for the Reactor Pressure Vessel integrity evaluation.

Required Action 4

• (Submit) a written report providing any newly acquired data as specified above and (1) the results of any necessary revisions to the evaluation of RPV integrity in accordance with the requirements of 10 CFR 50.60, _10 CFR 50.61, Appendices G and H to 10 CFR Part 50, and any potential impact on the LTOP or P-T limits in the technical specifications or (2) a certification that previously submitted evaluations remain valid. Revised evaluations and certifications should include consideration of Position 2.1 of Regulatory Guide 1.99, Revision 2, as applicable, and any new data."

WCNOC's Response to Required Action 4 This attachment provides relatively minor corrections to previously reported data and provides some new information, as specified above. With the exception of the revised data, however, the previously submitted evaluations remain valid.

This completes WCNOC's evaluations for and final response to Generic Letter 92-01, Revision 1, Supplement 1.

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i *Att'ac'hm:nt to ET 96-0044 l Page 5 of 6 TABLE 2

, Susunary File for Reference Temperature Pressurized Thermal Shock ( RTor. ) based on a "Proiected" 3.16 x 10" n / cm* EOL Fluence, Reaulatory Guide 1.99, i

Revision 2 and 10CFR50.61.

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Plant Name 13ellime lleat No. ID RI Nor.c Method of Chemistry Method of I Identification ident. Neutron Iktermining Factor Detennining  % Cu ' % Ni' RTm Fluence at IRTm (CF) CF 1 EOL '

2

( "F )

(n/cm )

j Wolf Intermediate NR 61 3.16 E 19 20*F Plant 26 ~1 able - 2 0.04 0.66 47.9 i Creek Shell R20051 836-1 specific EOL:

3/11/2025 intermediate NR61 3.16 E 19 -20*F Plant 26 'I able - 2 0.04 0.64 47.9 I Shell R2005-2 783-1 specific intermediate NR61 3.16 E 19 20*F Plant 31 l able - 2 0.05 0.63 54.4 j Shell R2005-3 799-1 specific

, Lower shcIl 118759-2 3.16 E l9 0*F Plant 58 Table . 2 0.09 0.67 109.6 g R2508-1 specific Lower shcli C4840-2 3.16 E 19 10*F Plant 37 lable - 2 0.06 0.64 92.2

.; R2508-2 specific j Lower shell C4935-2 3.16 E 19 40*F Plant 36.6 0.07 0.62 104.7 i 2 R2508-3 specific Calculated

int. and lower 11 4 3.16 E 19 -50*F Plant 27.8 l'able 1 0.04 0.04 42.2

, shell axial 90146 specific Linear welds G2.06 Interpolation I

int. to lower 11 4 3.16 E 19 50*F Plant 41.0 0.05 0.05 31.4 shell cire. 90146 specific Calculated *  ;

weld E3.16 ,

References

1. The copper and nickel contents and IRTsms of the beltline materials are from Table fl 1 of  !

! WCAP-13365, Rev.1.

l 2. Calculated values as described in 10CFR50.61, Equation 5. effective January 1996. l I

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' Attachment to ET 96-0044 PLge 6 of 6 TABLE 3 gjagnary File for Charov Uocer Shelf Enerav Based on a "Proiected" 3.16 x 10" n / cm* EOL Fluence, a 1/4 T value of 1.88 x 10" n / cm* and Reaulatory Guide 1.99, Revision 2. ,

l Plant Name 15cittme fleat No. Material 'lype 1/4 l' USE t/4 l' Neutron Unirradiated Method of -

i Identification ( note 4 ) at EOL Fluence at EOL USE Determining - ,

( ft Ib. ) (n/cm') ( fi-Ib. ) Unirradiated  ;

USE  ;

Wolf Intermediate NR bl 836-1 A 533B-1 1.88 E 19 127 Direct l Creek Shell R2005-1 99 ' ,

EOL:  !

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l 3/11/2025 Intermediate NR 617831 A 533151 1.88 E 19 127 Direct l 8 Shell R2005-2 99 l

l Intermediate NR 61799-1 A 53311-l 1.88 E 19 135 Direct f Shell R2005-3 105' I i

Lower shell B8759-2 A 53311-1 1.88 E 19 87 Direct ,

R2508-1 68 '

Lower shell C4840-2 A 5338-1 1.88 E 19 100 Direct R2508-2 78 '

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! Lower shell C4935-2 A 533B-1 1.88 E 19 86 Direct 1 2

R2508-3 83  !

l Intermediate and B4 90146 Linde 0091 1.88 E 19 150 Direct 8

lower shell axial SAW 117 welds G2.06 Intermediate to B4 90146 Linde 124 1.88 E 19 Surveil:ance 8 4 lower shcIl SAW 85 97 Weld circumferential weld E3.16  :

References i

f Material type for plates is from Table A-3 of WCAP-il553. Unirradiated USE data are from Table A-3 of WCAP-ll553 except as noted.

Material type for welds is from Table A-5 of WCAP-10015.  !

Notes ,

I. Based on a 22% reduction per Regulatory Position 1.2 and Figure 2 of Reg. Guide 1.99 Rev. 2.

2. Based on a 3% reduction per Regulatory Position 2.2 and Figure 2 of Reg. Guide 1.99 Rev. 2.

3. Based on a 12% reduction per Regulatory Position 2.2 and Figure 2 of Reg. Guide 1.99 Rev. 2.

4. Combustion Engineering Materials Certification Reports (CEMCRs),(Original Unirradiated USE data used due to apparent rounding in the WCAP. All Unirradiated USE values now agree with the original CEMCRs.).

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