ML18100B035
| ML18100B035 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/21/1994 |
| From: | Stone J Office of Nuclear Reactor Regulation |
| To: | Miltenberger S Public Service Enterprise Group |
| References | |
| TAC-M83507, TAC-M83508, NUDOCS 9404280021 | |
| Download: ML18100B035 (14) | |
Text
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- '~_j6~~d fr k UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 Mr. Steven E. Miltenberger Vice President and Chief Nuclear Officer Public Service Electric & Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038
Dear Mr. Miltenberger:
April 21, 1994
SUBJECT:
GENERIC LETTER (GL) 92-01, REVISION 1, "REACTOR VESSEL STRUCTURAL INTEGRITY," PUBLIC SERVICE ELECTRIC AND GAS COMPANY, SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 (TAC NOS. M83507 AND M83508)
By letters dated June 30, 1992, December 30, 1992, August 4, 1993, September 29, 1993 and December 1, 1993, Public Service Electric and Gas Company (PSE&G) provided its response to GL 92-01, Revision 1.
The NRC staff has completed its review of your responses.
Based on its review, the staff has determined that PSE&G has provided the information requested in GL 92-01.
The GL is part of the staff's program to evaluate reactor vessel integrity for Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs).
The information provided in response to GL 92-01, including previously docketed information, is being used to confirm that licensees satisfy the requirements and commitments necessary to ensure reactor vessel integrity for their facilities.
A substantial amount of information was provided in response to GL 92-01, Revision 1. These data have been entered into a computerized data base designated Reactor Vessel Integrity Database (RVID).
The RVID contains the following tables: A pressurized thermal shock (PTS) table for PWRs, a pressure-temperature limit table for BWRs, and an upper-shelf energy (USE) table for PWRs and BWRs. provides the PTS tables, Enclosure 2 provides the USE tables for Salem 1 and 2, and Enclosure 3 provides a key for the nomenclature used in the tables. The tables include the data necessary to perform USE and RTpts evaluations. These data were taken from your responses to GL 92-01 and previously docketed information. References to the specific source of the data are provided in the tables.
For Salem Unit 1, we request that you verify the information you have provided has been accurately entered in the summary data file.
No response is necessary unless an inconsistency is identified. If no comments are received within 30 days from receipt of this letter, the staff will consider your actions related to GL 92-01, Revision l, for Salem Unit 1 to be complete and the staff will use the information in the tables for future NRC assessments of the Salem Unit 1 reactor pressure vessel.
As a result of our GL 92-01 review, the staff has identified one open issue for Salem Unit 2. Additional data is required to confirm that the USE at end lfb\\
1 I
Mr. Steven April 21, 1994 of-life {EOL) is greater than 50 ft-lb because you have provided a generic unirradiated USE value, either a mean value from welds fabricated using the same flux type or a value based on your surveillance material. These types of values are unacceptable because they do not consider heat variability of the unirradiated USE.
When the unirradiated USE for a particular heat of material has not been determined, you can determine the lower tolerance limit with 95 percent confidence that at least 95 percent of the population is greater than the tolerance limit. The tolerance limit should be for all welds fabricated by the reactor vessel vendor unless it can be demonstrated that the welds are separable by flux type or other welding variable{s). The licensee must demonstrate that there is a_physical {metallurgical) difference in the welds and a statistical difference in the data to utilize a generic unirradiated USE for a particular flux type or other welding variable{s).
If the lower tolerance limit results in a projected µsE at EOL of less than 50 ft-lb for Salem Unit 2, then you must demonstrate, in accordance with Appendix G, 10 CFR Part 50, that lower values of USE will provide margins of safety against fracture equivalent to those required by Appendix G of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
We request that you submit within 30 days of receipt of this letter, a schedule for performing these analyses for Salem Unit 2.
Further, we request that you verify that the information you have provided for Salem Unit 2 has been accurately entered in the summary data file. If no comments are made in your response to the last request, the staff will use the information in the tables for future NRC assessments of the Salem Unit 2 reactor pressure vessel.
Once your response is received and your schedule is determined to be satisfactory, the staff will consider your actions related. to GL 92-01, Revision 1, for Salem Unit 2, to be complete.
When your analyses are submitted, they will be reviewed as.a plant-specific licensing action.
The information requested by this letter is within the scope of the overall burden estimated in GL 92-01, Revision 1, "Reactor Vessel Structural Integrity, 10 CFR 50.54{f)~
11 The estimated average number of burden hours is 200 person hours for each addressee's response. This estimate pertains only to the identified response-related matters and does not include the time
Mr. Steven April 21, 1994 required to implement actions required by the regulations. This action is covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.
Enclosures:
- 1.
Pressurized Thermal Shock Tables
- 2.
Upper-Shelf Energy Tables
- 3.
Nomenclature Key cc w/enclosures:
See next page DISTRIBUTION:
Docket File NRC & Local PDRs PDI-2 Reading SVarga JCalvo OFFICE NAME DATE CMil 1 er MO'Brien(2)
Sincerely,
/S/
James C. Stone, Senior Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation EWenzinger, RGN-I JWhite, RGN-I D:PDI-2 CM ILLER(
/L /94 I I I I
Mr. Steven April 21, 1994 required to implement actions required by the regulations. This action is covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.
Enclosures:
- 1.
Pressurized Thermal Shock Tables
- 2.
Upper-Shelf Energy Tables
- 3. Nomenclature Key cc w/enclosures:
See next page Sincerely,
~~e~ect Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
MY.-. Steven E. Mi 1 ten'9ger Public Service Electric & Gas Company cc:
Mark J. Wetterhahn, Esquire Winston & Strawn 1400 L Street NW Washington, DC 20005-3502 Richard Fryling, Jr., Esquire Law Department - Tower SE 80 Park Place Newark, NJ 07101 Mr. Calvin A. Vondra Gen~ral Manager - Salem Operations Salem Generating Station P.O. Box 236 Hancocks Bridge, NJ 08038 Mr; J. Hagan Vice President - Nuclear Operations Nuclear Department P.O. Box 236 Hancocks Bridge, New Jersey 08038 Mr. Charles S. Marschall, Senior Resident Inspector Salem Generating Station U.S. Nuclear Regulatory Commission Drawer I Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 Maryland People's Counsel American Building, 9th Floor 231 East Baltimore Street Baltimore, Maryland 21202
- Mr. J. T. Robb, Director Joint Owners Affairs PECO Energy Company 955 Chesterbrook Blvd., SlA-13 Wayne, PA 19087 Mr. S. LaBruna Vice President - Nuclear Engineering Nuclear Department P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Nucle~enerating Station, Units 1 and 2 Richard Hartung Electric Service Evaluation Board of Regulatory Commissioners 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Lower Alloways Creek Township c/o Mary 0. Henderson, Clerk Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Mr. Frank X. Thomson, Jr., Manager Licensing and Regulation Nuclear Department P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. David Wersan Assistant Consumer Advocate Office of Consumer Advocate 1425 Strawberry Square Harrisburg, PA 17120 Ms. P. J. Curham MGR; Joint Generation Department Atlantic Electric Company P.O. Box 1500 6801 Black Horse Pike Pleasantville, NJ 08232 Carl D. Schaefer External Operations - Nuclear Delmarva Power & Light Company P.O. Box 231 Wilmington, DE 19899 Public Service Cammi ssi on of Maryl and Engineering Division Chief Engineer 6 St. Paul Centre Baltimore, MD 21202-6806
t.NL, LU;)UKt. 1 Summary File for Pressurized Thermal Shock Plant Bel tl ine Heat No.
ID Neut.
IRT0.,.
Method of Chemistry Method of XCu Xiii Name I dent.
I dent.
Fluence at Determin.
Factor Determin.
EOL/EFPY
!RT.-
CF Salem 1 Int. shell C*1354*1 1.38E19 50°F e
Plant 157.82 Calculated 0.24 0.53 82402-1 Soecific EOL:
Int. shell C-1354-2 1.38E19 30°F e
Plant 154.37 Calculated 0.34 0.53 8/13/2016 82402-2 Scecific Int. Shell C-1397-2 1.38E19 18°F e
Plant 115.53 Calculated 0.22 0.51 82402-3 Scecific Lower C-1356-1 1.35E19 23°F II MTEB 5-2 128.8 Table 0.19 0.48 Shell 82403-1 Lower C-1356-2 1.35E19 30°F 9
MTEB 5-2 129.9 Table 0.19 0.49 Shell 82403-2 Lower C-1356-3 1.35E19 38°F 9
MTEB 5-2 128.8 Table 0.19 0.48 Shell 82403-3 Int. Shell 398196 &
1.10E19
-56°F Generic 214 Table 0.18 1.00 Axial 348009 llelds (T) 2-042 Int. to 13253 1.25E19
-56°F Generic 195.8 Table 0.25 0.72 Lower Shell Circ. lleld 9-042 Lower 348009 1.35E19
-56°F Generic 220 Table 0.19 1.00 Shell Axial llelds 3-042 References for Salem 1 Material type and data other than that noted below was provided in the June 30, 1992 letter from S.E. Miltenberger CPSE&G) to USNRC Docl.Jlleflt Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54(f), Salem Generating Station Unit Nos. 1 and 2.
Chemistries for the intermediate shell course plates were provided in the August 4, 1993 letter S. LaBruna to USNRC Docl.Jlleflt Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.5054Cf) Request for Additional Information, Salem Generating Station Unit Nos. 1 and 2.
Initial RT~T and l.IJSE values for the beltline plates were provided in the Septenber 29, 1993 letter from s. La8r111a to USNRC Docl.Jlleflt Control Desk, "Supplemental Response to Generic 92-01, Revision 1 Reactor Vessel Structural lnt~rity, 10CFR50.54Cf) Request for Additional Information, Salem Generating Station Unit No. 1.
Fluence reported in a Decetd>er 15, 1992 letter from s. LaBruna CPSE&G) to USNRC.
8Staff determination of IRTNDT from transverse CVN data.
9Staff determination of IRTN T from longitudinal CVN data by converting longitudinal data to transverse ~ata using the 65% correction factor, in accordance with MTEB 5-2.
8
Summary File for Pressurized Thermal Shock Plant 8eltline Heat No.
ID Neut.
I RT""'
Method of Chemistry Method of XCu XNi Name
!dent.
I dent.
Fluence at Determin.
Factor Determin.
EOL/EFPY I RT""'
CF Salem 2 Int. Shell C-4173-1 1.38E19 5°F 9
MTEB 5-2 89.8 Table 0.13 0.56 84712-1 EOL:
Int. Shell C-4186-2 1.38E19 12°F Plant 97.677 Calculated 0.12 0 *. 62 4/18/2020 84712-2 Soecific Int. Shell C-4194-2 1.38E19 50°F 9
MTEB 5-2 73.7 Table 0.11 0.57 84712-3 Lower C-4182-1 1.41E19 33°F II MTEB 5-2 83 Table 0.12 0.60 Shell 84713-1 Lower C-4182-2 1.41E19 22°F 9
MTEB 5-2 82.4 Table 0.12 0.57 Shell 84713-2 Lower B-8343-1 1.41E19 20°F II MTEB 5-2 82.6 Table 0.12 0.58 Shell 84713-3 Int. Shell 13253 &
1.05E19
-40°f Plant 194.89 Calculated 0.23 0.73 Axial 20291 en Specific Welds 2-442 Lower 21935 &
1.07E19
-56°F Generic 202.7 Table 0.20 0.86 Shell 12008 (T)
Axial Welds 3-442 Int. to 90099 1.38E19
-56°F Generic 95 Table 0.18 0.20 Lower Shell Circ. Weld 9-442 References for Salem 2 Material type, initial RT~r and data other than that noted below was provided in the June 30, 1992 letter S.E.
Miltenberger CPSE&G) to USNRC Docunent Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54Cf), Salem Generating Station Units Nos. 1 and 2.
The chemistry for plate 84712-2 was provided in the August 4, 1993 letter from s. LaBrll\\ll to USNRC Docunent Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54(f) Request for Additional Information, Salem Generating Station Unit Nos. 1 and 2.
Fluence reported in a December 15, 1992 letter from s. LaBrllla (PSE&G) to USNRC.
9Staff determination of IRTN T from longitudinal CVN data by converting longitudinal data to transverse ~ata using the 65% correction factor, in accordance with MTEB 5-2.
9
t.NLLU~UKt. ~
8 Summary File for Upper Shelf Energy Plant Name 8eltline Heat No.
Material 1/4T USE 1/4T Unirrad.
Method of
!dent.
Type at Neutron USE Determin.
EOL/EFPY Fluence at Unirrad.
EOl/EFPY USE Salem 1 Int. shell C-1354-1 A5338-1 70 8.2E18 91 Direct 82402-1 EOL:
Int. shell C-1354-2 A5338-1 83 8.2E18 98 Direct 8/13/2016 82402-2 Int. Shell C-1397-2 A533B-1 85 8.2E18 104 Direct 82402-3 Lower C-1356-1 A533B-1 68 8.0E18 93 65X Shell 82403-1 Lower C-1356-2 A533B-1 61 8.0E18 83 65X Shell 82403-2 Lower C-1356-3 A5338-1 62 8.0E18 85 65X Shell 82403-3 Int. Shell 398196 &
Linde 53 6.6E18 75 NRC Axial 348009 (T) 1092, SAW Generic Welds 2-042 Int. to 13253 Linde 69 8.0E18 111 Surv.
Lower 1092, SAW Weld Shell Circ. Weld 9-042 Lower 348009 Linde 75 8.0E18 112 Sister Shell 1092, SAW Plant Axial Welds 3-042
9 Plant Name Beltline I dent.
Reference for Salem 1 Summary File for Upper Shelf Energy Heat No.
Material Type 1/4T USE at EOL/EFPY 1/4T Neutron Fluence at EOL/EFPY Unirrad.
USE Method of Determin.
Unirrad.
USE Material type and data other than that noted below was provided in the June 30, 1992 letter S.E. Miltenberger (PSE&G) to USNRC Document Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54(f), Salem Generating Station Unit Nos. 1 and 2.
Chemistries for the intermediate shell course plates were provided in the August 4, 1993 Letter from s. LaBruna to USNRC Document Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54(f) Request for Additional Information, Salem Generating Station Unit Nos. 1 and 2.
Initial RT..,, and UUSE values for the beltline plates were provided in the Septeni>er 29, 1993 Letter from s. LaBruna to USNRC Document Control Desk, "Supplemental Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54(f) Request for Additional Information, Salem Generating Station Unit No. 1.
UUSE values for the beltline welds except int. shell axial welds 2-042, were provided in the Deceni>er 4, 1993 Letter from S. LaBruna to USNRC Document Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54(f) Request for Additional Information, Salem Generating Station Unit Nos. 1 and 2.
UUSE value for int. shell axial welds 2-042 is a generic value for welds fabricated by Combustion Engineering using Linde 1092, 0091, 123 and Arcos B-5 fluxes (Ref. Letter from S.
Bloom (USNRC) to T. L. Patterson (OPPO) dated Deceni>er 3, 1993.
Fluence based on data provided in a Deceni>er 15, 1992 Letter from s. LaBr\\.118 (PSE&G) to USNRC.
10 Summary File for Upper Shelf Energy Plant Name Bel tline lieat No.
Material 1/4T USE 1/4T Unirrad.
Method of I dent.
Type at Neutron USE Determin.
EOL/EFPY Fluence at Unirrad.
EOL/EFPY USE Salem 2 Int. Shell C-4173-1 A 5338-1 70 8.2E18 90 65X 84712-1 EOL:
Int. Shell C-4186-2 A 5338-1 66 8.2E18 83 65X 4/18/2020.
84712-2 Int. Shell C-4194-2 A 5338-1 60 8.2E18 75 65X 84712-3 Lower C-4182-1 A 5338-1 66 8.4E18 83 65X Shell 84713-1 Lower C-4182-2 A 5338-1 70 8.4E18 88 65X Shell 84713-2 Lower B-8343-1 A 5338-1 70 8.4E18 88 65X Shell 84713-3
/
Int. Shell 13253 &
Linde 60 a 6.3E18 96 a Generic Axial 20291 en 1092, SAW Welds 2-442 Lower 21935 &
Linde 75 6.4E18 114 Sister Shell 12008 en 1092, SAW Plant Axial Welds 3-442 Int. to 90099 Linde 52 8.2E18 75 NRC Lower 0091, SAW Generic Shell Circ. Weld 9-442 6Licensee utilized a generic UUSE value; additional information required to confirm the value.
11 Summary File for Upper Shelf Energy Plant Name Beltline Ident.
Heat No.
-Material Type 1/4T USE at EOL/EFPY 1/4T Unirrad.
Neutron USE Method of Determin.
Unirrad.
USE Reference for Salem 2 Fluence at EOL/EFPY Material type, Initial RT..,T and data other than that noted below was provided in the June 30, 1992 letter S.E. Miltenberger (PSE&G) to USNRC Document Control Desk, "Response to Generic Letter 92*01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54Cf), Salem Generating Station Unit Nos. 1 and 2.
The chemistry for plate 84712-2 was provided in the August 4, 1993.letter from s. LaBrll'UI to USNRC Document Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity, 10CFR50.54Cf) Request for Additional Information~ Salem Generating Station Unit Nos. 1 and 2.
UUSE value for the beltline welds except for 9-442 were provided in the December 4, 1993 letter from S. LaBruna to USNRC Document Control Desk, "Response to Generic Letter 92-01, Revision 1 Reactor Vessel Structure Integrity, 10CFR50.54(f) Request for Additional Information, Salem Generating Station Unit Nos. 1 and 2.
UUSE values for weld 9-442 is a generic value for welds fabricated by Combustion Engineering using Linde 1092, 0091, 123 and Arcos B-5 fluxes (Ref. letter from s. Bloom CUSNRC) to T.L.
Patterson (OPPD) dated December 3, 1993.
Fluence based on data provided in a De~ember 15, 1992 lette from s. LaBruna CPSE&G) to USNRC.
ENCLOSURE 3 PRESSURIZED THERMAL SHOCK TABLES AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE Pressurized Thermal Shock Table Column 1:
Plant name and date of expiration of license.
Column 2:
Beltline material location identification.
Column 3:
Beltline material heat number; for some welds that a single-wire or tandem-wire process has been reported, (S) indicates -
single wire was used in the SAW process, (T) indicates tandem wire was used in the SAW process.
Column 4:
End-of-life (EOL) neutron fluence at vessel inner wall; cited directly from inner diameter (ID) value or calculated by using Regulatory Guide (RG) 1.99, Revision 2, neutron fluence attenuation methodology from the quarter thickness (T/4) value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Col~mn 5:
Unirradiated reference temperature.
Column 6:
Method of determining unirradiated reference temperature
( IRT).
Plant-Specific This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
MTEB 5-2 This indicates that the unirradiated reference temperature was determined from following MTEB 5~2 guidelines for cases where the IRT was not det_ermined using American Society of*
Mechanical Engineers Boiler and Pressure Vessel Code, Section Ill, NB-2331, methodology.
Generic This indicates that the unirradiated reference temperature was determined from the mean value of tests on material of similar types.
Column 7:
Chemistry factor for irradiated reference temperature evaluation.
Column 8:
Method of determining chemistry factor.
Table This indicates that the chemistry factor was determined from the chemistry factor tables in RG 1.99, Revision 2.
- Calculated This indicates that the chemistry factor was determined from surveillance data via.procedures described in RG 1.99, Revision 2.
ENCLOSURE 3 Page 2 Column 9:
Copper conteht; cited directly from licensee value except when more than one value was reported.
(Staff used the average value in the latter case.)
No Data This indicates that no copper data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.
Column 10: Nickel content; cited directly from licensee value except when more than one value was reported.
(Staff used the average value in the latter case.)
No Data This indicates that no nickel data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.
Upper Shelf Energy Table Column 1:
Plant name.and date of expiration of license.
Column 2:
Beltline material loc~tion identification.
Column 3:
Beltline material heat number; for some welds that a single-wire or tandem-wire process has been feported, {S) indicates single wire was used in the SAW process.
(T) indicates tandem wire was used in the SAW process~
Column 4:
Material type; plate types include A 533B-l, A 302B, A 302B Mod., and forging A 508-2; weld types include SAW welds using Linde 80, 0091, 124, 1092, ARCOS-BS flux, Rotterdam welds using Graw Lo, SMIT 89, LW 320, and SAF 89 flux, and SMAW welds using.no flux.
Column 5:
EOL upper-shelf energy (USE) at T/4; calculated by using the EOL fluence and either the cooper value or the surveillance data.
(Both methods are described in RG 1.99, Revision 2.)
EMA This indicates that the USE issue may be covered by the approved equivalent margins analysis in a topical report.
Column 6:
EOL neutron fluence at T/4 from vessel inner wall; cited directly from T/4 value or calculated by using RG 1.99, Revision 2, neutron fluence attenuation methodology from the ID value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 7:
Unirradiated USE.
EMA ENCLOSURE 3 Page 3 This indicates that the USE issue may be covered by the approved equivalent margins analysis in a topical report.
Column 8:
Method of determining unirradiated USE.
Direct For plates, this indicates that the unirradiated USE was from a transverse specimen.
For welds, this indicates that the unirradiated USE was from test date.
6S%
This indicates that the unirradiated USE was 6S% of the USE from a longitudinal specimen.
Generic This indicates that the unirradiated USE was reported by the licensee from other plants with similar materials to the beltline material.
NRC generic This indicates that the unirradiated USE was derived by the staff from other plants with similar materials to the beltline material.
10, *30, 40. or SO °F This indicates that the unirradiated USE was derived from Charpy test conducted at 10, 30, 40, or SO °F.
Surv. Weld This indicates that the unirradiated USE was from the surveillance weld having the same weld wire heat number.
Equiv. to Surv. Weld This indicates that the unirradiated USE was from the surveillance weld having different weld wire heat number.
Sister Plant This indicates that the unirradiated USE was derived by using the reported value from other plants with the same weld wire heat number.
Blank Indicates that there is insufficient data to determine the unirradiated USE.