ML20058G388
| ML20058G388 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 12/03/1993 |
| From: | Hebert J Maine Yankee |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, JRH-93-248, MN-93-111, NUDOCS 9312090206 | |
| Download: ML20058G388 (7) | |
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1 iiEll ABLLEL ECTRICITY SINCE 1972 EDISOfJ DR!vE
- AUGUSTA, MA!NE 04330 + (207) 622-4568 December 3, 1993 MN-93-111 JRH-93-248 UNITED STATES NUCLEAR REGULATORY COMMISSION l
Attention: Document Control Desk Washington, D.C.
20555 I
References:
(a)
License No. DPR-36 (Docket No. 50-309)
(b) USHRC Letter Dated.0ctober 1,1993 - Requast for Additional Information--Maine Yankee Response to Generic letter 92-01, Reactor Vessel Structural-Integrity, Rev.1, (TAC No. M83479).
l (c) USNRC Letter Dated March 6,1992 - Reactor Vessel Structural Integrity,10CFR 50.54(f) (Ge:eric letter 92-01, Revision 1).
(d) MY letter to USNRC dated July 2,1992 (MN-92-65) Response to Generic Letter 92-01, Revision 1 (Reactor Vessel Structural Integrity).
Subject:
Response to Generic Letter 92-01, Revision 1 (Reactor Vessel Structural Integrity)
Gentlemen:
This letter responds to your request for additional information, Reference (b), related to Reactor Vessel Structural Integriiy.
Our response to your request is provided in Enclosure 1.
The response is provided in a question / answer format. Where applicable, our response provides specific references to the sources of previously docketed information, including references, attachments and page numbers.
We trust this information is satisfactory.
Very truly yours,
[
070108 Ja es R. Hebert, Manager Licensing & Engineering Support Department Enclosures c: Mr. Thomas T. Martin Mr. J. T. Yerokun Mr. E. H. Trottier Mr. Patrick J. Dostie Mr. William Olsen Mr. Clough Toppan 9322090206 931203 T
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ENCLOSURE 1 Maine Yankee l
l Response to USNRC Request for Additional Information Related to Reactor Vessel Structural Integrity (Generic Letter 92-01, Revision 1) l
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INTRODUCTION l
Maine Yankee's response to your request for additional information related to i
Reactor Vessel Structural Integrity, ReFcrence (b), is provided in this enclosure. The response is provided in a question / answer format and is intended l
to supplement information provided in Reference (d).
Where applicable, our response provides specific references to the sources of previously docketed information, including references, attachments and page number.
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION USNRC Staff Question No. 1 Your response to Question 2a does not specify unirradiated upper shelf energy l
(USE; values for axial welds 2-203 and 3-303.
Please provide either the unirradiated Charpy USE for each forging, or provide the unirradiated Charpy USE and analysis from forgings and welds that were fabricated using the same vendor, fabrication time frame, fabrication process, and material specification. This information is required to demonstrate that the forgings and welds will meet the USE requirements of Appendix G to 10 CFR Part 50.
1 If this information can not be provided, please submit an analysis demonstrating that lower values of USE will provide margins of safety against fracture that are equivalent to those required by Appendix G to the ASME Boiler and Pressure Vessel l
Code.
l MYAPCo Response to Question 1 Maine Yankee's response to Question 2a of Reference (c), did not specify unirradiated Upper Shelf Energy (USE) values for Longitudinal Welds 2-203 and 3-203 (note correction to typographical error). This is because we do not predict USE to be less than 50 ft-lbs at end of the operating license using the guidance in paragraphs C.2.'2 in Regulatory Guide 1.99, Revision 2.
i However, we are providing the information in the following sections to assist i
your staff in completing its review.
'onoitudinal Weld 2-203 Longitudinal Welds 2-203 A, B and C join three (3) plates in the intermediate shell of the Reactor Pressure Vessel (RPV) using weld wire heat _ Mil-B4 51989 and Linde 124 flux lot 3687.
The unly charpy data available for this combination wer from qualification tests conducted at a temperature of +10'F.
The results of inree (3) tests (50, 60 and 72 ft-lbs) are reported in Table C6-2 of Reference (e) which was docketed by Reference (f). We are not aware of any other charpy data for weld wire heat Mil-B4 51989.
ABB-Combustion Engineering (ABB-CE) performed an evaluation to establish a best estimate initial USE and uncertainty for Weldments using Linde 124. The results of ABB-CE's evaluation yield a best estimate initial USE of 102.3 ft-lbs _and a standard deviation of 9.4 ft-lbs.
We conclude, based upon the results of the three (3) cnarpy tests results at a temperature of +10 F and the results of ABB-CE's evaluation, that the initial USE of Weld 2-203 is ;t 83.5 ft-lbs.
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Lont'tudinal Weld 3-203 I
Longitudinal Welds 3-203A, B and C join three (3) plates in the lower shell of i
the RPV using weld wire heats Mil B-4 Mod.13253 and 12008 in tandem.
The process used Linde 1092 flux lot 3833. The only charpy data available for this
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combination were from qualification tests conducted at a temperature of +10'F.
The results of six (6) tests (62, 47, 62, 79, 79 and 82 ft-lbs) are reported in Table C6-2, of Reference (e) which was docketed by Reference (f).
Although insufficient data exists for this combination to establish an initial USE, the i
constituents were used extensively in similarly constructed vessels and i
associated surveillance programs.
Yankee Atomic Electric Company (YAEC) performed an evaluation to establish an I
initial USE for Maine Yankee Welds 3-203A, B and C.
Their evaluation used available data from similarly constructed vessels which used the same constituents.
YAEC concluded the following based upon their evaluation The probable initial USE is greater than 107 ft-lbs, the initial USE of Maine Yankee's surveillance weld.
The projected USE at end of the operating license is bounded by the USE of the surveillance weld, 53 ft-lbs.
This projection is based upon the use of conservative va Nes of initial USE (85 ft-lbs) and copper content-(0.24w/o)for Welds 3-203A, B, and C.
Maine Yankee participated in a Combustion Engineering Owner's Group (CEOG) task to perform a generic bounding analyses for low USE.
Combustion Engineering's analyses, which were transmitted to the NRC in Reference (m), concluded that actual material properties for CE0G vessels will not fall below the 50 ft-lb criterion of 10 CFR 50 Appendix G.
Furthermore, Combustion Engineering concluded q
that beltline materials with Charpy USE levels significantly below 50 ft-lbs meet j
the acceptance criteria of ASME Code r - N-512.
j Although Maine Yankee's position is that the.3E for the Maine Yankee vessel will not fall below 50 ft-lbs at end of the operating license, the generic bounding analyses provide additional technical evidence that there are no safety concerns with low USE.
USNRC Staff Question No. 2 a
- a. Your response to Question 2b appears to contain a discrepancy in the chemical composition of circumferential weld 9-203. As shown in the following table, a
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weld data in WCAP-12819 differs from that presented in Table 5.1 of the G. D.
Whittier (Maine Yankee) letter to A. C. Thadani (USNRC). The Whittier letter was in response to the pressurized thermal shock (PTS) rule (10 CFR 50.61).
Circumferential Weld 9-203 j
4 Chemical Element WCAP-12819 PTS Response Copper 0.36%
0.31%
Please resolve this apparent discrepancy by providing the correct copper content, and the basis for the content, to characterize this weld.
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USNRC Staff Ouestion No. 2 (CONTINUED)
- b. In addition, your response to GL 92-01 did not provide chemical composition values (phosphorous and sulfur) for axial welds 2-203 and 3-203.
Please provide the phosphorous and sulfur content values for these welds.
j MYAPCo Response to Question 2a.
The staff's value for copper content extracted for WCAP-12819 (0.36 w/o) appears j
in Table 4-1 of Reference (g). This value originates from baseline testing of Maine Yankee's Surveillance Weld specimens performed by ABB-CE and reported in Reference (h). This test result is only one of numerous chemical analysis. tests now available for the weld wire heat and flux lot (IP3571 and Linde 1092 flux lot 3958) comprising the surveillance weld.
The staff's value for copper content for the PTS Response (0.31 w/o) appears to be extracted from Table 5.1 of Reference (i) for Weld 9-203.
This value is a best estimate derived from the mean of seven (7) tests [See footnote 8 of Table 5.1, Reference (i)].
The staff reviewed the chemical compositions reported in Reference (i), including the Copper content of 0.31 w/o for Weld 9-203, and focnd them acceptable in Reference (j). The staff concluded that "the Maine Yankee pressure vessel meets the. material toughness requirements of 10 CFR 50.61 to the end of the current operating license on October 21, 2008."
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Maine Yankee has updated the PTS response twice since Reference (i) was reviewed and approved in Reference (j).
The first update provided in Reference (k),
4 incorporated the following changes i
Revised fluence projections based upon Westinghouse Electric Company (W) re-evaluation of fluence.
j Regulatory Guide 1.99, Revision 2 (information only).
The second update provided in Reference (1) incorporated the following changes:
Revised fluence projections which reflect the results of Wall Capsule 253 analysis, Reference (g), and re-evaluation of other capsule and cavity 4
measurements.
4 Revised PTS Rule (10 CFR 50.61) incorporating Regulatory Guide 1.99, Rev. 2.
Revised the value for nickel content in Weld 9-203 from 0.74 to 0.76 w/o as a result of Wall Capsule 253 Analysis, Reference (g), and re-evaluation of RPV and surveillance capsule weld chemistry [See footnote 8 of -Table 5.1, i
Reference (1)].
The revised nickel content identified above was.the result of an increase in the number of chemical analysis results available for Weld Wire Heat and flux lot (IP3571 and Linde 1092 flux lot 3958) comprising Weld 9-203. Although the best
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estimate nickel content increased slightly (0.74 to 0.76 w/o), the best estimate copper content remained the same (0.31 w/o).
See Tables B-2 of References (i) j and (1) for historical information related to chemical content for Weld 9-203.
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MYAPCo Response to Question 2b.
The chemical composition for Longitudinal Welds 2-203 A, B and C and 3-203 A, B and C are provided in Table 1.
Table 1 Chemical Composition Welds 2-203 A, B and C and 3-203 A, B and C Weld Cu Ni P
S 2-203 A, B and C 0.17' O.17' O.012 0.010" l
6 3-203 A, B and C 0.22" 0.84*
0.016' O.011*
a Value reported in Table 5.1 of References (i) and (1).
b Fort Calhoun head weldment sample:
Identical wire (Heat $1989)/ flux lot (3687).
c Value obtained from same method used to determined Cu and Ni content in Tables B-1 of References (i) and-(l).
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REFERENCES j
(a)
License No. DPR-36 (Docket No. 50-309)
(b)
USNRC Letter Dated October 1,1993 - Request for Additional Information--
Maine Yankee Response to Generic Letter 92-01, Reactor Vessel Structural i
Integrity, Revision 1 (TAC No. M83479).
(c)
USNRC Letter Dated March 6,1992 - Reactor Vessel Structural Integrity, 10CFR 50.54(f) (Generic Letter 92-01, Revision 1).
(d)- MY letter to USNRC dated July 2, 1992 (MN-92-65) Response to Generic Letter 92-01, Revision 1 (Reactor Vessel Structural Integrity).
(e)
" Evaluation of Pressurized Thermal Shock Effects due to Small Break LOCA's with Loss of Feedwater for Combustion Engineering NSSS", C-E Topical Report CEN-189, December 1981.
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(f)
MY Letter to USNRC dated December 31, 1981 (FMY-81-189) Response to Item II.k.2.13 of NUREG-0737, " Thermal Mechanical Report".
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l (g)
E. Terek, et. al., " Analysis of the Maine Yankee Reactor Vessel Second Wall Capsule located at 253*", Westinghouse Report WCAP-12819, March 199],
(h)
" Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Maine Yankee Reactor Vessel Materials", CE Report CENPD-37, December 30, 1971.
(i)
MY Letter to USNRC dated January 21, 1986 (MN-86-14) Response to Requirement of 10 CFR 50.61 (Pressurized Thermal Shock Rule).
(j)
USNRC Letter to My dated November 20,1986-Material Properties and Fast Neutron Fluence for Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (10 CFR 50.61).
(k)
MY Letter to USNRC dated December 6,1988 (MN-88-ll8) Update of Assessment i
10 CFR 50.61 Fracture Toughness for Protection Against Pressurized Thermal Shock.
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(1)
MY Letter to USNRC dated October 28, 1991 (MN-91-151)
Update of PTS Assessment to Address the PTS Rule (10 CFR 50.61).
(m)
CE0G letter to USNRC dated September 27, 1993 (CE0G-93-479) Final Evaluation of Low Upper Shelf Energy, a
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