ML20153G257

Forwards Response to NRC 980710 RAI Re Reactor Vessel Integrity,Including Test Results for Capsule.Three WCAPs from W Which Provide Rept of Capsule V Test Results Also Encl

ML20153G257
Person / Time
Site:	Wolf Creek
Issue date:	09/25/1998
From:	Maynard O WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20153G260	List: ML20153G257 ML20153G269 ML20153G277 ML20153G285
References
GL-92-01, GL-92-1, TAC-MA0584, TAC-MA584, WM-98-0101, WM-98-101, NUDOCS 9809300016
Download: ML20153G257 (11)

Text

. . . . --

tr [

A W$LF CREEK NUCLEAR OPERATING CORPORATION Otto L Maynard' President and Chief Executive Offg:er SEP 2 51998 WM 98-0101 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:

Reactor Vessel Structural Integrity, dated May 19, 1995  ;

2) Letter ET 95-0084 dated August 17, 1995, from R. C. Hagan, WCNOC, to USNRC

3) Letter ET 96-0044 dated June 25, 1996, from R. A. Muer.ch, WCNOC, to USNRC

4) NRC Request for Additional Information Regarding Generic Letter 92-01, Revision 1, Supplement 1, dated March 26, 1998

5) Letter WM 98-0056 dated July 10, 1998, from O. L. Maynard, WCNOC to OSNRC

Subject:

Docket No. 50-43S: Re.sponse to Request for Additional Information Regarding Beactor Pressure Vessel Integrity (Generic Letter 92-01, Revision 1, supplement 1) and Submittal of Reactor lessel Surveillance Capsule Test Results per 10 CFR 50, Appendix H (TAC NO. MA0584)

Gent lemen:

Reference 1 requested all Licensees of operating plants to identify, collect, and report any new data pertinent to the analysis of the structural integrity

[

of their reactor pressure vessels (RPVs). Reference 1 also requested all Licensees to assess the impact of those data on their RPV integrity analyses relative to the requirements of Section 50.60 of Title 10 of the Code of Federal Regulations (10 CFR Part 50.60), 10 CFR Part 50.61, Appendices G and H to 10 CFR Part 50 (which encompass pressurized thermal shock [ PTS) and upper r shelf energy (USE] evaluations), and any potential impact on low temperature over pressure (LTOP) limits or pressure-temperature (P-T) limits. Wolf Creek i Nuclear Operating Corporation (WCNOC) provided responses to that request in References 2 and 3.

In July, 1997, the Combustion Engineering Owners Group (CEOG) issueo a report, CE NPSD-1039, . Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated' Reactor Vessel Welds," that provides additional RPV weld chemistry data for vessel welds fabricated by Combustion Engineering. On March 26, 1998, the NRC issued Reference 4, requesting that WCNOC consider the impact of the revised best-estimate chemistries for vessel beltline welds from the CEOG 9809300016 980925 '

PDR ADOCK 05000482 P PDR g'on, KS 66839 / Phone: (316) 364-8831 An Equal Opportunity Employer M/F/HCNET M o43ooono

~~

g<> i s k ik; < YWM[98-0101-

,. =t

) Page

2 ;of f 2'. .

< s -

)

3.., . ,

.u, . -

I I' , report, l.and'.TtoLconsider~'the need ~ ton revise previous ' evaluations. of RPV integrity for~the'WCGS: reactor pressure vessel.

I 1

2In' Reference 15 WCNOC' informed' the NRC' that we had' shipped .to Westinghouse a i

< vesse10 capsule ~. specimen: (Capsule . V) ' removedf in November, 1997, during our {

ninth: refueling:; outage, andithaththeftest_results.for the capsule, alcng with' J: s i

the"informationineeded to' address Reference 4, would be provided to the NRC by.

~

r

^ ' .I iSept ember.l 30,-.1998. ? Attached totthis' letter.'is WCNOC's response'to Reference  !

i s '4.1'Alsoienclosed.are three WCAPs from Westinghouse, which provide ~the report' '!

.of~ Capsule;V test results, iniaccordance with 10 CFR_50, Appendix H, Section

. .III,7 Paragraph:A;

-l 4

Iflyou have anr quertions regarding this response, please contact me at -(316) j e364-8831, . extension'4000, or Mr. Michael ~J. Angus.at extension ~4077. j s -

Very truly yours, s

,l

^

<- _ l s 4 l

4 .j Otto L. Maynard f OLM/rlr -j

' M JAttachments

Enclosures:

^*"

1. WCAP-15078, " Analysis of f Capsule V - from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel j Radiation ' Surveillance Program," - Revision; 1, t dated 1

.i Sept ember, 1998

2. ;WCAP-l'2079, " Wolf Creek : Heatup = and-- Cooldown Limit Curves for Normal Operation," Revision 1, dated September,i1998 ',

3. WCAP-15080/- " Evaluation of Pressurized Thermal Shock for Wolf Creek", Revision 1, dated September, 1998 ]

, I

^

'cc t W. l D.c Johnson (NRC),'w/a, w/e'.Y E. W. Merschoff~(NRC),.w/a, w/o B . - A. Smalldridge-(NRC), . w/a, w/e i

- K.iM. Thomas-(NRC), w/a,fw/e c , ,

)

}

p p.: _

i-t w

x-  ;

4 5 > <

i t ., , . ,

1

. . . . .- -.. .-.- - . ~ . . - . . . . . . - - . . . -. . - - - .

- 1 l:

i i

l l

'i l

[.

STATE OF KANSAS )  ;

) SS COUNTY OF COFFEY )

Otto L. Maynard, of lawful age, being first duly sworn upon oath says that he  :

'is President and Chie.f Executive Officer 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 the best of his knowledge, information and belief.

l l

ny

[-.

l

/l#///w

/

j

\

Otto L. Naynt.rd '

President'and l Chief Executive Officer l-t I

. SUBSCRIBED and sworn to-before me this 2 W day of h 2N8, .

1998.

? .: -0

LINDA DELONG-OHM!E Notary Public State of Kansas Nothry Public' (6  ;

My Appt Expires August 31,2002 g $3 2002 U

i 1

l I

4 I.

j. 4 i

i.

~

1 l

l-l

Attachment I to WM 98-0101 l Page 1 of 7 Response to NRC Request for Additional Information Regarding Reactor Pressure Vesse] Integrity at Wolf Creek l RAI Section 1.0 Assessment of Best-Estimate Chemistry l Based on the information in the CEOG Report CE NPSD-1039, Revision 2, "Best i Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," and in accordance with the provisions of Generic Letter 92-01, Revision 1, Supplement 1, the NRC requests the following:

1. An evaluation of the information in the reference above and an assessment of its applicability to the determination of the best- l estimate chemistry for all of your RPV beltline welds. Based upon this I reevaluation, supply the information necessary to completelv fill out i the data requested in Table 1 for each RPV beltline weld r arial.

i Also provide a discussion for the copper and nickel valut tosen for each weld wire heat r7 ting what heat-specific data were it zuded and excluded from the analysis and the analysis method chosen zor determining the best-estimate. If the limiting material for your vessel's PTS /P-T limits evaluation is not a weld, include the l information requested in Table 1 for the limiting material also.

Furthermore, you should consider the information provided in Section 2.0 of this RAI on the use of surveillance data when responding.

Response to Item 1:

There is one weld heat, Heat Number 90146, in the Wolf Creek Generating Station (WCGS) vessel beltline. The best-estimate copper and nickel chemistries for this weld heat were reevaluated using the mean-of-the-mean i method recommended in the Combustion Engineering Owners Group (CEOG) report.

In addition, there was a chemistry analysis performed on three weld metal Charpy specimens from WCGS Capsule V. Hence, these values were also included in the determinaticn of the revised copper and nickel values (See Table 1).

Contained in the following table is a comparison of the best-estimat? copper and nickel values from the CEOG report and the latest reactor vessel integrity evaluation for the WCGS reactor vessel beltline weld metal.

l Reactor Vessel CEOG Values Current Values Beltline Weld Material Cu wt. % Ni wt. % Cu wt. % Ni wt. %

l Heat # 90146 0.039 0.082 0.043 0.083

, The current best-estin. ate chemistries are included in Table 2. The current l chemistries are slightly higher than the values determined in the CEOG report.

( Updated Adjusted Reference Temperature values are computed for this weld along

! with the information and ca' ulated values for limiting lower shell plate l R2508-1. The RTns values 'nted in Table 2 are documented in WCAP-15080,

" nal Shock for Wolf Creek", Revision 1, dated l " Evaluation of Pressurize

! September, 1998, and have updated as part of the WCGS Capsule V l

evaluation.

RAI Section 2.0 Evaluation and Use of Surveillance Data Since the evaluation ef surveillance data relies on both the best-estimate chemical composition of the RPV weld and the surveillance weld, the information in the CEOG report may result in the need to revise previous evaluations of RPV integrity (including LTOP setpoints and P-T limits) per the i

I

.. - .. _ . -= _. . . . .. . . _ . - . .- _ _ . - -

l Attachment I 'o WM 98-0101 Page 2 of 7 l

requirements of 10 CFR 50.60, 10 CFR 5 ) . 61, and Appendices G and H to 10 CFR Part 50.

Based on this information and consistent with the provisions of Generic Letter 92-01, Revision 1, Supplement 1, the NRC requests the following:

2. that (1) the information listed in Table 2, Table 3, and the chemistry factor from the surveillance data be provided for each heat of material for which surveillance weld data are available and a revision in the RPV integrity analyses (i.e., current licensing basis) is needed or (2) a certification that previously submitted evaluations remain valid.

Separate tables should be used for each heat of material addressed. If the limiting material for your vessel's PTS /P-T limits evaluation is not a weld, include the information requested in the tables for the limiting material (if surveillance data are available fcr this material).

Response to Item 2:

l The beltline weld metal is not the limiting beltline material. Lower shell '

plate R2508-1 is the limiting material for the PTS evaluation and lower shell plate R2508-3 is the limiting material for the pressure-temperature curves evaluation. The data requested by Item 2 is contained in Tables 2, 3, 4 and 5 below. The surveillance data for the WCGS reactor vessel is considered l credible. The credibility evaluation is presented in Appendix D of WCAP-15078, " Analysis of Capsule V from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program,"

Revision 1, dated September, 1998.

Based on the latest Capsule V evaluation (WCAP-15078), the pressurized thermal shock (PTS) evaluation (WCAP-15080), and the pressure-temperature (P-T) curves I (WCAP-15079), have been revised. Charpy curves developed using the CVGRAPH TANH fitting program are given in Appendix C of WCAP-15078.

RAI Section 3.0 PTS /P-T Limit Evaluation

3. If the limiting material for your plant changes or if the adjusted reference temperature for the limiting material increases as a result of the above evaluations, provide the revised RTn3 value for the limiting material in accordance with 10 CFR 50.61. In addition, if the adjusted RTum value increased, provide a schedule for revising the P-T and LTOP limits. The schedule should ensure that compliance with 10 CFR Part 50, Appendix G, is maintained.

Response 3:

The limiting material for WCGS has not changed. The beltline weld metal is not the limiting beltline material. Lower shell plate R2508-1 is the limiting material for the PTS evaluation and lower shell plate R2508-3 is the limiting material for the pressure-temperature curves evaluation. The PTS Evaluation and the P-T Limit Curves have been updated based on the Capsule V evaluation and the current copper and nickel values given in Section 1 of this response.

The revised RTns values for the limiting materials are documented in WCAP-15080 and the revised P-T Limit Curves are provided in WCAP-15079. The LTOP Limits are being reviewed based on the new P-T Limit Curves and WCNOC anticipates submitting any necessary revisions by December 31, 1998. This schedule will not impact 10 CFR Part 50, Appendix G compliance for WCGS.

I

\

l l

Attachment I to WM 98-0101 Page 3 of 7 l

TABLE 1 Facility: Wolf Creek Vessel Manufacturer: Combustion Engineeri_n_g Information RequeJted on RPV Weld and/or Limiting Material RPV Weld Best- Best- EOL ID Assigned Method of Initial a ao Margin ART or Wire Estimate Estimate Fluence Material Determining RTNOT RTers Heat / Plate 0) Copper Nickel (x 104)D) Chemistry CF(2) (RTNDhU)) at EOL Factor (CF)o) 90146 0.04(4) 0.08td) 2.18 28.3 Sury. Data -50 0 14 28 12 R2508-1 0 09(S 0.67(M 2.18 58.0 Table 0 0 17 34 104 R2508-3 0.09(4) 0.58(4) 2.18 35.8 Sury. Data 40 0 8.5 17 100 (1) or the material identification of the limiting material as requested in Section 1.0 (2) determined from tables or from surveillance data (3) documented in the WCGS PTS Evaluation (WCAP-15080)

(4) Best-estimate copper and nickel values are from Tables 6 and 7 and have been rounded to two decimal places (5) Best-estimate copper and nickel values are from Material Certification Discussion of the Analysis Method and Data Used for Each Weld Wire Heat Weld Wire Heat Discussion 90146 weighted mean copper and nickel based on 16 measurements (See Table 6)

TABLE 2: Heat 90146*

Capsule ID Cu'2) Ni '2 3 Irradiation Fluence Measured Data Used in (including Temperature (x10 ARTm Assessing source) (*F) n/cm[) '

(op) Vessel (Y or N)

U 0.04 0.08 553.7 0.3429 27.21 Y Y 0.04 0.08 553.7 1.308 45.09 Y V 0.04 ,0.08 553.7 2.528 46.33 Y NOTES:

1) All data contained in Table 2, with the exception of the i irradiation temperature, was obtained from the Capsule V analysis l (see WCAP-15078)

2) Best-estimate copper and nickel values are from Table 6 and have l been rounded to two decimal places l

l

.- . - - ~. .. . _ - -- -

l Attachment I to WM 98-0101 l

Page 4 of 7 TABLE 3: Heat 90146"3 Capsule ID Cu '2) Ni(2) Irradiation Fluence Measured Adjusted I Predicted (Adjusted -

(including Temperature Factor ARTmy ARTu ARTer Predicted) source) (*F) (*F) (*F) (*F) ARTer

(*F)01 U 0.04 0.08 553.7 0.705 27.21 19.75 27.50 -7.75 Y 0.04 0.08 553.7 1.075 45.09 32.74 41.93 -9.19 V 0.04 0.08 553.7 1.249 46.33 33.64 48.71 -15.07 NOTES:

1) All data contained in Table 3, with the exception of the irradiation temperature, was obtained from the Capsule V analysis (see WCAP-15078) and the PTS Evaluation report (WCAP-15080)

2) Best-estimate copper and nickel values are from Table 6 and have l been rounded to two decimal places i

3) The " Adjusted ARTner" is the " Measured ARTS:7" adjusted by a ratio of 0.726 (CFvw / CFsw = 31.6 / 43.5 = 0.726) l l

l i

l TABLE 4: Plate R2508-3"8 Capsule ID Orient. Cu '2 3 Ni'2' Irradiation Fluence Measured Data Used (including Temperature (x10" ARTurr in source) (

• F) n /cm,' ) ( o p) assessing Vessel (Y or N) l U Long. 0.09 0.58 553.7 0.3429 36.46 Y Y Long. 0.09 0.58 553.7 1.308 16.03 Y V Long. 0.09 0.58 553.7 2.528 52.03 Y U Trans. 0.09 0.58 553.7 0.3429 23.79 Y Y Trans. 0.09 0.58 553.7 1.308 35.39 Y V Trans. 0.09 0.58 553.7 2.528 54.53 Y NOTES:

1) All data contained in Table 4, with the exception of the irradiation temperature, was obtained from the Capsule V analysis (see WCAP-15078)

2) Best-estimate copper and nickel values are from Table 7 and have been rounded to two decimal places

Attachment I to WM 98-0101

> Page 5 of 7 TABLE 5: Plate R2508-3"'  !

1 l

Capsule ID Orient Cu'D Ni c 2' icadiation Fluence Measured Adjusted Predicted (Adjusted - ,

(induding Temperature Factor ARTuor ARTuor ARTwor Predicted) l source) (*F) (*F) (*F) m (.F) ARTwor  !

(*F)

U Long. 0.09 0.58 553.7 0.705 36.46 36.46 25.24 11.22 Y Long. 0.09 0.58 1.G'. ;

553.7 f 16.03 16.03 38.49 -22.46 V Long. 0 09 0.58 553.7 1.249 52.03 52.03 44.71 7.32 U Trans. 0.09 0.58 553.7 0.705 23.79 23 75 25.24 1.45 l

Y Trans. 0.09 0.58 553.7 1.075 35.39 35.39 38.49 -3.1 V Trans. 0.09 0.58 553.7 1.249 54.53 54.53 44.71 9.82 NOTE'J :

1) All data contained in Table 5, with the exception of the irradiation temperature, was obtained from the Capsule V analysis (see WCAP-15078)

2) Best-estimate copper and nickel values are from Table 7 and have been rounded to two decimal places

3) The Adjusted ARTuer is not applicable, since, the ratio procedure only applies to the weld metal

Attachment I to NM 98-0101 Page 6 of 7 i

TABLE 6: Heat 90146 l Best Estimate Cu and Ni Weight Percent Values for the WCGS j Beltline Weld Seams 4

Reference Measured Measured Weighted Weighted

% Cu  % Ni Average Average

% Cu  % Ni CE NPSD-1039, Rev. 2 0.04 0.04 0.04 0.04 CE NPSD-1039, Rev. 2 0.03 0.09 0.03 0.09 CE NPSD-1039, Rev. 2 0.03 0.10 CE NPSD-1039, Rev. 2 0.03 0.08 CE NPSD-1039, Rev. 2 0.03 0.04 0.0325 0.0725 CE NPSD-1039, Rev. 2 0.04 0.11 CE NPSD-1039, Rev. 2 0.03 0.06 CE NPSD-1039, Rev. 2 0.03 0.08 CE NPSD-1039, Rev. 2 0.04 0.09 0.04 0.1467 CE NPSD-1039, Rev. 2 0.04 0.15 l CE NPSD-1039, Rev. 2 0.04 0.20 CE NPSD-1039, Rev. 2 0.05 0.05 0.05 0.05 CE NPSD-1039, Rev. 2 0.04 0.09 0.067 0.097

& WCAP-10015 WCAP-15078 0.078 0.103 WCAP-15078 0.074 0.094 WCAP-15078 0.075 0.101 SUM 0.697 1.478 0.2595 0.4962 Average. 0.044 0.092 0.043 0.083 NOTE: The values used for the evaluations are the mean of the mean values rounded to two decimal places (i.e., %Cu - 0.04 and %Ni = 0.08).

l~ '

Attachment I to WM 98-0101 Page 7 of 7 TABLE 7: Plate R2508-3 Best Estimate Cu and Ni Weight percent Values for the WCGS Lower Shell Plate R2508-3 i

Reference Measured % Cu Measured % Ni l

1 WCAP-10015 0.07 0.62 WCAP-15078 0.104 0.532 SUM 0.174 1.152 Aver'.ge 0.087 0.576 I

NOTE: The values used for the evaluations are the mean of the mean values rounded to two decimal places. (i.e., %Cu - 0.09 and 1Ni = 0.58) l l

i l

' Attachment II to WM 98-0101 Page 1 of 1 LIST OF COMMITMENTS The following table identifies those actions committed to by Wolf Creek Nuclear Operating Corporation (WCNOC) in this document. Any other statements in this submittal are provided :or information purposes and are not considered to be commitments. Please direct questions regarding these commitments to Mr.

Michael J. Angus, Manager Licensing and Corrective Action at Wolf Creek Generating Station, (316) 364-8831, extension 4077 COMMITMENT Due Date/ Event The LTOP Limits are being reviewed based on the December 31, 1998 new P-T Limit Curves and will be submitted by December 31, 1998.