ML20236P099
| ML20236P099 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/13/1998 |
| From: | Rainsberry J SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-MA0569, TAC-MA0570, TAC-MA569, TAC-MA570, NUDOCS 9807160150 | |
| Download: ML20236P099 (7) | |
Text
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- l. L Rainsberry Manager, Plant Licensing An LD/ SOY /V7tRNAT/Ot(L= Gnupany e
July 13, 1998 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555-0001
Subject:
NRC Request for Additional Information Regarding Reactor Pressure Vessel Integrity at San Onofre Nuclear Generating Station Units 2 and 3 (TAC Nos. MA0569 and MA0570)
References:
1.
Letter from J. W. Clifford (NRC) to H. B. Ray (SCE), dated April 8, 1998,
Subject:
Request for Additional Information Regarding Reactor Pressure Vessel Integrity at San Onofre Nuclear Generating Station Units 2 and 3 (TAC Nos. MA0569 and MA0570) 2.
ABB CE Owners Group (CE0G) Draft Report, CE NPSD-1119, dated June 1998, titled " Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content, CE0G Task 1054" 3.
ABB CE Topical Report, CE NPSD-1039, Rev. 2, dated June 1997, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds" 4.
Letter from W. C. Marsh (SCE) to U.S. Nuclear Regulatory Commission (Document Control Desk), dated August 16, 1995,
Subject:
Response to Generic Letter 92-01, Revision 1, Supplement 1, " Reactor Vessel Structural Integrity,10 CFR 50.54(f)"
This letter provides the information requested in the April 8,1998 NRC letter from J. W. Clifford (NRC) to H. B. Ray (SCE) (Reference 1).
The following information concerning San Onofre Units 2 and 3 is from the ASEA Brown Boveri (ABB) Combustion Engineering Owners Group (CE0G) Draft Report, CE I, O NPSD-1119 (Reference 2) which addresses NRC requests to Utilities on this
- topic, y
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9807160150 900713 PDR ADOCK 05000361 P
PDR San Onolre Nuclear Generating Station P.O.Ikn 128 San Clemente. CA 926744)128 714- % 8-7420
4 Document Control Desk 4 NRC Request: Section 1.0 - Assessment of Best-Estimate Chemistry The staff recently received additional information that may affect the determination of the best-estimate chemistry composition for your RPV welds or
'your surveillance weld material. This information was provided to the NRC by the Combustion Engineering Owners' Group in report CE NPSD-1039, Revision 02, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," dated June 1997.
Based on this information, in accordance with the provisions of Generic Letter 92-01, Revision 1, Supplement 1, the NRC requests an evaluation of the information in the reference above and an assessment of its applicability to the determination of the best-estimate chemistry for all of your RPV beltline welds. Based upon this reevaluation, supply the information necessary to completely fill out the data requested in Table 1 for each RPV beltline weld material. Also provide a discussion for the copper and nickel values chosen for each weld wire heat noting what heat-specific data were included and excluded from the analysis, and the analysis method chosen for determining the best-estimate.
If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the 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.
Southern California Edison (SCE) Response:
CE NPSD-1119 (Reference 2) summarizes the results from Topical Report CE NPSD-1039, Revision 02 (Reference 3) on determining best estimate values for I
copper / nickel content of reactor vessel beltline welds. CE NPSD-1119, prepared for the Combustion Engineering Owners Group on best estimate copper and nickel content for Combustion Engineering fabricated reactor vessel welds, provides the following:
1
- 1. Updated values for copper and nickel which are more consistent with 10 CFR 50.61 in response to NRC staff comments regarding some of the weld wire heats reported in CE NPSD-1039, Revision 02.
- 2. Responses to comments made by the NRC staff with respect to the l
evaluation conducted in CE NPSD-1039, Revision 02.
- 3. A list identifying the weld wire heats used in reactor vessel surveillance program welds fabricated by Combustion Engineering.
This list was generated to identify source data as part of the effort reported in CE NPSD-1039, Revision 02.
Document Control Desk
- B 4.Recommendedbestestimatecopper/nickelvaluesforusewithreactor vessel welds fabricated by Combustion Engineering:
1 6eneric Nickel in Mn-Mo Wire
- 0.13% Ni Generic Copper in Copper Coated Wire
- 0.27% Cu Generic Copper in Non-Coated Wire
- 0.05% Cu Best Estimate Nickel in N1200 Welds
- 1.038% Ni SCE has reviewed Draft Report CE NPSD-1119 for weld wire heat numbers of welds identical to weld wire heat numbers for reactor vessel beltline welds provided in our Generic Letter (GL) 92-01 response (Reference 4) and compared the copper and nickel content. The welds with identical weld wire heat numbers including copper and nickel content are shown in Tables A and B below.
Table A - San Onofre, Unit 2
-GL 92-01 Submittal CE NPSD-1119 RV Belt 11ne'
' Copper /NickelContent.
BestEstimateCopper/ Nickel Welds
-(Reference 1)
Content Seam No.
Heat i
. Copper.
Nickel Heat i Copper Nickel 2-203A BOLA 0.03 0.90 BOLA 0.027 0.913 2-203B BOLA 0.03 0.91 BOLA 0.027 0.913 2-203C BOLA 0.03 0.95 BOLA 0.027 0.913 3-203A 83637 0.05 0.12 83637 0.048 0.066 3-203B 83637 0.04 0.06 83637 0.048 0.066 3-203C 83637 0.06 0.11 83637 0.048 0.066 9-203 90130 0.07 0.29 90130 0.044 0.133 Surveillance 90130 0.03*
0.12*
90130 0.044 0.133 Surveillance 90130 0.03**
0.15**
90130 0.044 0.133 Measured when surveillance program was developed.
Measured when surveillance tests were performed for Capsule 97.
Document Control Desk.
Table B - San Onofre, Unit 3 GL 92-01 Submittal CE NPSD-1119 RV Beltline Copper /NickelContent 8estEstimateCopper/ Nickel Welds'
'(Reference 1)
Content
-Seam No.~
Heat'f
-Copper Nickel--
Heat i Copper Nickel 2-203A 83650 0.04 0.17 83650 0.045 0.087 2-2038 83650 0.05 0.21 83650 0.045 0.087 2-203C 83650 0.04 0.08 83650 0.045 0.087 3-203A 88114 0.04 0.21 88114 0.043 0.189 3-2038 88114 0.04 0.19 88114 0.043 0.189 3-203C 88114 0.04 0.21 88114 0.043 0.189 l
9-203 90144 0.05 0.04 90144 0.042 0.075 l
9-203 90069 0.06 0.04 90069 0.04 0.076 Surveillance 90069 0.03*
0.08*
90069 0.04 0.076 Surveillance 90069 0.03**
0.11**
90069 0.04 0.076 l
Surveillance 90069 0.03 "
0.09 "
90069 0.04 0.076 Measured when surveillance program was developed.
- Measured when surveillance tests were performed for Capsule 97.
SCE's review of CE NPSD-1119 concluded that the SCE response to Generic Letter 92-01, Revision 1, Supplement 1 (Reference 4) remains valid for the following reasons:
- 1. Copper coated electrodes were not used in the fabrication of the San Onofre Units 2 and 3 Reactor Pressure Vessel (RPV) beltline region.
Document Control Desk -
- 2. Consistent with the NRC requested Table 1 information, previously provided in Reference 4 for the beltline plate and weld materials, the beltline plate material remains the limiting material in the derivation of the Pressurized Thermal Shock (PTS) screening criteria and Reference Temperature for nil ductility transition as adjusted for irradiation effects (RTm).
- 3. Although other plants may share material heat numbers with the RPV for San Onofre Units 2 and 3, slight variations in copper content as I
shown in Tables A and B (i.e., differences in chemistry results between different laboratories) will not impact the San Onofre i
Units 2 and 3 PTS, Pressure Temperature (PT), or Low Temperature Overpressure Protection (LTOP) limits.
As such, based on the above, the best estimate copper and nickel values developed and recommended in CE NPSD-1119 for the CE0G do not impact our response to GL 92-01, Revision 1, Supplement 1 (reference 4).
NRC Request: Section 2.0 - Evaluation and Use of Surveillance Data The chemical composition report referenced in Section 1.0 includes updated chemistry estimates for heats of weld metal. These reports not only provide a suggested best estimate value, but also include the source data used in estimating the chemical composition of the heat of material.
This permits the determination of the best estimate chemical composition for the various sources of data including surveillance welds.
Since the evaluation of surveillance data rely on both the best estimate chemical composition of the RPV weld and the surveillance weld, the information in these reports may result in the need to revise previous evaluations of RPV integrity (including LTOP setpoints and PT limits) per the requirements of 10 CFR 50.60, 10 CFR 50.61, and Appendices G and H to 10 CFR Part 50.
Based on this information and consistent with the provisions of Generic Letter l
92-01, Revision 1, Supplement 1, the NRC requests 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 /PT limits evaluation is not a weld, include the information requested in the tables for the limiting material (if surveillance data are available for this material).
L-- - =-------------- -
Document Control Desk.
i L
SCE Response:
l CE NPSD-1119 also included weld wire heat numbers for reactor vessel surveillance program welds fabricated by CE. These weld wire heat numbers are included in Tables A and B above. As concluded in response to Section 1.0, l
.the beltline plate material for both San Onofre Units 2 and 3 remains the L
limiting material for the vessel beltline. As such, the evaluation on surveillance data provided in Reference 4 remains valid and no changes will be l
required to existing reactor vessel PT limits or the LTOP setpoint. The information requested for Table 2, Table 3, and the chemistry factor from. the surveillance data for each heat of material for which surveillance weld data were available was provided in Reference 4.
NRC Requests. Section 3.0 - Pressurized Thermal Shock / Pressure Temperature (PTS /PT) Limit Evaluation 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 L
evaluations, provide the revised Reference Temperature for pressurized thermal shock (RT,rs) value.for the. limiting material in accordance with 10 CFR 50.61.
In addition, if the adjusted RT,or value increased, provide a schedule for
. revising the PT and_LTOP limits. The schedule should ensure that compliance with 10 CFR 50 Appendix G is maintained.
l SCE Response:-
CE NPSD-1119 did not. change the San Onofre Units 2 and 3 beltline limiting
. material:for determining the RT,rs or RT,or.
The limiting material remains the
' beltline plate material' for San Onofre Units 2 and 3, as provided in Reference 4..
The existing adjusted RTns and RT,or values, identified in Reference 4, remain valid for San Onofre Units 2 and 3, and no changes are required to Pressure Temperature (PT) and Low Temperature Overpressure Protection (LTOP) limits.
Schedule for CE0G Eeport CE NPSD-1119
-The final CE0G report CE NPSD-1119 is forecasted to be completed in August 1998. The final report will include further discussion on the issue of using Mn-Mo electrode and a N1200 cold wire feed, which is not applicable to San Onofre. The chemical compositions of the copper and nickel content shown in l
Tables A'and B above are not expected to change in the final report. Should
- the final CE0G report CE NPSD-1119 impact this SCE response to the NRC request for additional information, a revised response will be provided.
L
Document Control Desk ~
If you have any questions or would like additional information on this subject, please let me know.
Ve y truly yours, D
b
)
cc:
E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 J. W. Clifford, NRC Project Manager, San Onofre Units 2 and 3 I'
L.