ML18011A431
| ML18011A431 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 05/13/1994 |
| From: | Le N Office of Nuclear Reactor Regulation |
| To: | Robinson W CAROLINA POWER & LIGHT CO. |
| References | |
| GL-92-01, GL-92-1, TAC-M83468, NUDOCS 9405180371 | |
| Download: ML18011A431 (18) | |
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Docket No. 50-400 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Hay 13, 1994 Hr.
W.
R. Robinson Vice President - Harris Plant Carolina Power
& Light Company Shearon Harris Nuclear Power Plant Post Office Box 165 - Hail Code:
Zone 1
New Hill, North Carolina 27562-0165
Dear Hr. Robinson:
SUBJECT:
RESPONSE
TO GENERIC LETTER (GL) 92-01, REVISION 1, "REACTOR VESSEL STRUCTURAL INTEGRITY," SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 (TAC NO. H83468)
By letter dated July 6, 1992, Carolina Power
& Light Company (CP&L) provided its response to GL 92-01, Revision 1, For the Shearon Harris Nuclear Power Plant (SHNPP).
The NRC staff has completed its review of your response.
Based on their review, the staff has determined that CP&L 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 the Reactor Vessel Integrity Database (RVID).
The RVID contains the following tables:
A pressurized thermal shock (PTS) table for PWRs, a
pressure-temperature limits table for BWRs and an upper-shelf energy (USE) table for PWRs and BWRs.
Enclosure 1 provides the PTS table, Enclosure 2
provides the USE table for your SHNPP, and Enclosure 3 provides a key for the nomenclature used in the tables.
The tables include the data necessary to perform USE and RT evaluations.
These data were taken from your response to GL 92-01 and previously docketed information.
References to the specific source of the data are provided in the tables.
We request that you verify that the information that you have provided for the SHNPP has been accurately entered in the summary data files.
No response is necessary unless an inconsistency is identified. If no comments are received within 30 days from the day of this letter, the staff will consider your actions related to GL 92-01, Revision 1, to be complete and the staff will use the information in the tables for future NRC assessments of your reactor pressure vessel.
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May 13, 1994 Hr.
W. R. Robinson 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)."
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 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.
Sincerely, Original Signed. by:
Ngoc B. Le, Project Manager Project Directorate II-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Pressurized Thermal Shock or Pressure-Temperature Limit Table 2.
Upper-Shelf Energy Table 3.
Nomenclature Key cc w/enclosures:
See next page DISTRIBUTION Docket File, NRC
& Local PDRs PDI II-1 Reading S. Varga G. Lainas W. Bateman P. Anderson N.
Le D. HcDonald E. Hackett J. Strosnider OGC ACRS (10)
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W.
R. Robinson 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)."
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 required to implement actions required by the regulations.
This action is covered by the Office of Hanagement and Budget Clearance Number 3150-0011, which expires June 30, 1994.
Sincerely',
pcs 8 Ngoc B. Le, Project Hanager Project Directorate II-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Pressurized Thermal Shock or Pressure-Temperature Limit Table 2.
Upper-Shelf Energy Table 3.
Nomenclature Key cc w/enclosures:
See next page
I I
Mr. W.
R. Robinson Carolina Power
& Light Company Shearon Harris Nuclear Power Plant Unit 1
CC:
Mr. H.
Ray Starling Manager - Legal Department Carolina Power
& Light Company Post Office Box 1551
- Raleigh, North Carolina 27602 Resident Inspector/Harris NPS c/o U.S. Nuclear Regulatory Commission 5421 Shearon Harris Road New Hill, North Carolina 27562-9998 Karen E.
Long Assistant Attorney General State of North Carolina Post Office Box 629
- Raleigh, North Carolina 27602 Public Service Commission State of South Carolina Post Office Drawer 11649
- Columbia, South Carolina 29211 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta St.,
N.W. Suite 2900 Atlanta, Georgia 30323 Mr. W.
R. Robinson Vice President - Harris Plant Carolina Power
& Light Company Shearon Harris Nuclear Power Plant Post Office Box 165, MC: Zone 1
New Hill, North Carolina 27562-0165 Mr. Dayne H. Brown, Director Division of Radiation Protection N.C. Department of Environmental Commerce
& Natural Resources Post Office Box 27687
- Raleigh, North Carolina 27611-7687 Mr. H.
W. Habermeyer, Jr.
Vice President Nuclear Services Department Carolina Power
& Light Company Post Office Box 1551
- Raleigh, North Carolina 27602 Admiral Kinnaird R.
McKee 214 South Morris Street Oxford, Maryland 21654 Mr. Robert D. Martin 3382 Sean Way Lawrenceville, Georgia 30244
ENCLOSURE 1
Summary File for Pressurized Thermal Shock Plant Name Shearon Harris Belt line Ident.
Int. Shell Heat No.
Ident.
A9153-1 ID Neut.
Fluence at EOL/EFPY 3.42E19 600F Method of Determin.
IRT Plant S cific Chemistry Factor 58 Method of Determin.
CF Table XCu 0.09 EOL:
10/24/
2026 Int. Shell 84197-2 3.42E19 910F Plant Specific 38 Calculated 0.10 Lower Shell Lower Shell Int. and Lower Axial Maids Circ.
'LIQld C-9924-1 C.9924-2 484784 5P6771 3.42E19 3 '2E19 1.33E19 3.42E19 54'F 57'F
-200F
-200F Plant S cific Plant Specific Plant Specific Plant S
cific 51 51 82 54 Table Table Table Table 0.08 0.08 0.06 0.04 REFERENCES FOR SHEARON HARRIS:
- Fluence, IRT~ and chemistry data from April 2, 1992 surveiLlance capsule report, "Analysis of Capsule V."
ENCLOSURE 2
Summary File for Upper Shelf Energy Plant Name Beltline Ident.
Heat No.
Material Type 1/4T USE at EOL/EFPY 1/4T Neutron Fluence at EOL/EFPY Unirrad.
USE Method of Determin.
Unirrad.
USE Shearon Harris Int. Shell A9153-1 A 5338-1 2.148E19 83 Direct EOL:
10/24/2026 Int. Shell 84197-2 A 5338-1 2.148E19 71 Direct Lower Shell Lower Shell Int. and lower Axial Welds Circ.
Weld C-9924-1 C-9924-2 5P6771 A 5338-1 Linde 124, SAW Linde 124, SAW Linde 124, SAW 76 76 2.148E19 2.148E19 0.835E19 2.148E19 98 94 Direct Direct Direct Surv.
Weld REFERENCES FOR SHEARON HARRIS:
- Fluence, UUSE data from April 2, 1992 surveillance capsule report, "Analysis of Capsule V.w
NOMENCLATURE AND TABLES ENCLOSURE "3 PRESSURIZED THERMAL SMOCK AND USE TABLES FOR ALL PWR PLANTS N
MENCLATURE Pressurized Thermal Shock Table Column Column Column Column Column Column 2:
3
~
4
~
5
~
6:
Plant name and date of expiration of license.
Beltline material location identification.
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.
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'atest submittal (GL 92-01, PTS, or P/T limits
.submittals).
Unirradiated reference temperature.
Method of determining unirradiated reference temperature (IRT).
Ppl-t iR This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
Column Column 7
~
8:
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 determined using American Society of Mechanical Engineers Boiler and Pressure Vessel
- Code,Section III, NB-2331, methodology.
Generic This indicates that the unirradiated reference temperature was determined from the mean value of tests on material of similar types.
Chemistry factor for irradiated reference temperature evaluation.
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.
Column 9:
Calculated This indicates that the chemistry factor was determined from surveillance data via procedures described in RG 1.99, Revision 2.
Copper 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 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 Column Column Column Column Column Column I
~
2:
3
~
4
~
5
~
6:
7
~
Plant name and date of expiration of license.
Beltline material location identification.
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.
Material type; plate types include A 533B-I, A 302B, A 302B Mod.,
and forging A 508-2; weld types include SAW welds using Linde 80,
- 0091, 124,
- 1092, ARCOS-B5 flux, Rotterdam welds using Graw Lo, SMIT 89, LW 320, and SAF 89 flux, and SMAW welds using no flux.
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 either owners group or plant-sp'ecific equivalent margins analyses.
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).
Unirradiated USE.
EMA This indicates that the USE issue may be covered by either owners group or plant-specific equivalent margins analyses.
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.
65'his indicates that the unirradiated USE was 65X 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.
II 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 50 'F This indicates that the unirradiated USE was derived from Charpy test conducted at 10, 30, 40, or 50 'F.
Surv.
Weld This indicates that the unirradiated USE was from the surveillance weld having the same weld wire heat number.
E uiv. 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.
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