BSEP-96-0344, Application for Amends to Licenses DPR-62 & DPR-71,revising P-T Limits Curves Currently Located in Units 1 & 2 TSs & Deleting Current 8,10 & 12 Effective Full Power Yr Hydrostatic Test pressure-temp Limit Curves to Add New Ones
| ML20133F353 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/07/1997 |
| From: | Campbell W CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20133F357 | List: |
| References | |
| BSEP-96-0344, BSEP-96-344, NUDOCS 9701140155 | |
| Download: ML20133F353 (8) | |
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C._P_.&_L Carolina Power & Light Company William R. Campbell PO Box 10429 Vice President Southport NC 28461-0429 Brunswick Nuclear Plant January 7,1997 1
SERIAL: BSEP 96-0344 10 CFR 50.90 TSC 95TSB06 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk i
Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS PRESSURE-TEMPERATURE LIMITS CURVES Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power & Light Company hereby requests a revision to the Technical Specifications for the Brunswick Steam Electric Plant (BSEP), Unit Nos.1 and 2. The purpose of this request is to:
1.
Exchange the pressure-temperature limits curves currently located in the Unit 1 and Unit 2 Technical Specifications. In Licensee Event Report 1-94-05, CP&L reported that the Unit 1 and Unit 2 pressure-temperature (P-T) limits curves had been inadvertently transposed.
This request is an administrative change to relocate the pressure-temperature limits curves to Technical Specifications of the unit to which they correctly correspond.
2.
Delete the current 8,10 and 12 effective full power year (EFPY) hydrostatic test pressure-temperature limits curves and incorporate new 14 and 16 effective full power year hydrostatic test P-T limits curves for the Brunswick Unit 1 and 2 reactor pressure vessels.
The current reactor vessel pressure-temperature limits curves contained in the Technical Specifications for hydrostatic pressure tests are suitable for up to 12 effective full power years (EFPY) of reactor operation. It is anticipated that both units will surpass this threshold during 1997. Based on this expectation, new hydrostatic test pressure-temperature limits curves for 14 and 16 EFPY were developed.
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The proposed revisions have been incorporated in CP&L's license amendment application for the I
Improved Technical Specifications; therefore, CP&L requests that the revisions provided herein be issued prior to or concurrent with approval of the Brunswick Improved Technical
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Specifications (ITS). The Company anticipates that both Brunswick Unit 1 and Unit 2 will
/M' I surpass the 12 EFPY threshold during 1997; thus, the hydrostatic test pressure-temperature limit
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curves contained in the Technical Specifications will no longer be valid for use. Based on this expectation CP&L requests that these license amendment requests be approved no later than the start of Unit 2 Refueling Outage 12 (currently scheduled to begin September 13,1997) to 9701140155 970107 PDR ADOCK 05000324 P
PDR Tel 910 457-2496 Fox 910 457 2803
Document Control Desk BSEP 96-0344 / Page 2 assure the availability of a valid hydrostatic pressure test limit curve for Unit 2. Exigent review and approval by the NRC may be necessary if an unplanned mid-cycle outage and hydrostatic pressure test should be required. In order to support orderly revision of plant procedures and training, CP&L requests the proposed license amendments be effective on the date of issuance and implementation required within 30 days following issuance.
Carolina Power & Light Company is providing, in accordance with 10 CFR 50.91(b), Mr. Dayne H.
Brown of the State of North Carolina with a copy of the proposed license amendments.
Please refer any questions regarding this submittal to Mr. Mark A. Turkat at (910) 457-3066.
Sincerely, William R. Campbell WRM/wrm
Enclosures:
1.
Basis for Change Request 2.
10 CFR 50.92 Evaluation
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3.
Environmental Considerations 4.
Page Change Instructions 5.
Typed Technical Specification and Bases Pages - Unit 1 6.
Typed Technical Specification and Bases Pages - Unit 2 7.
Marked-up Technical Specification and Bases Pages - Unit 1 j
8.
Marked-up Technical Specification and Bases Pages - Unit 2 William R. Campbell, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Carolina Power & Light Company.
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Notary (Seal)
My commission expires:
,My Commission Expires August 21, 1999
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i, Document Control Desk j
BSEP 96-0344 / Page 3 j
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U. S. Nuclear Regulatory Commission ATTN.: Mr. Luis A. Reyes, Regional Administrator 101 Marietta Street, N.W., Suite 2900 Atlanta, GA 30323-0199 Mr. C. A. Patterson NRC Senior Resident inspector - Brunswick Units 1 and 2:
U.S. Nuclear Regulatory Commission L
ATTN.: Mr. David C. Trimble, Jr. (Mail Stop OWFN 14H22) 11555 Rockville Pike i
Rockville, MD 20852-2738 The Honorable R. Hunt I
Chairman (Acting)- North Carolina Utilities Commission j
P.O. Box 29510
,j Raleigh, NC 27626-0510 l
Mr. Dayne H. Brown Director - Division of Radiation Protection North Carolina Department of Environment, Health, and Natural Resources P.O. Box 27687 i
Raleigh, NC 27611-7687 i
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ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AND 2 NRC DOCKETS 50-325 AND 50-324 OPERATING LICENSES DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS PRESSURE-TEMPERATURE LIMITS CURVES' BASIS FOR CHANGES CURRENT REQUIREMENT Technical Specification 3/4.4.6 describes the limiting conditions for operation and surveillance requirements for the reactor coolant system pressure and temperature limits. Technical Specification 3.4.6.1 requires that the reactor coolant system temperature and pressure be limited:
As shown in Technical Specification Figure 3.4.6.1-1 for heat-up by non-nuclear means, cool down following a nuclear shutdown, and low power physics testing.
As shown in Technical Specification Figure 3.4.6.1-2 for operations with a critical core other than low power physics tests or when the reactor vessel is vented.
As shown in Technical Specification Figures 3.4.6.1-3a, 3.4.6.1-3b, or 3.4.6.1-3c, as applicable, for hydrostatic or leak testing.
PROPOSED CHANGES The following changes are proposed for Technical Specification 3/4.4.6.1:
1.
Exchange Technical Specification Figures 3 4.6.1-1 and 3.4.6.1-2 between the Brunswick Unit 1 and Brunswick Unit 2 Technical Specifications.
2.
Delete the current Figures (Figures 3.4.6.1-3a, 3.4.6.1-3b, and 3.4.6.1-3c) for 8,10 and 12 effective full power year (EFPY) hydrostatic test pressure-temperature limits curves and incorporate new 14 and 16 EFPY hydrostatic test pressure-temperature limits curves for the Brunswick Unit i and 2 reactors (proposed Figures 3.4.6.1-3a and 3.4.6.1-3b, respectively). Commensurate changes to the references in Technical Specification 3/4.4.6.1 and Bases 3/4.6 are also proposed to reflect the deletion of current Technical Specification Figure 3.4.6.1-3c.
3.
Reformat Technical Specification Figures 3.4.6.1-1, 3.4.6.1-2, 3.4.6.1-3a, and 3.4.6.1-3b.
The changes associated with reformatting the figures are administrative in nature.
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BASIS FOR PROPOSED CHANGE Pressure-Temoerature Limits Curves Exchanae:
In Licensee Event Report 1-94-05 dated March 22,1994 (Reference 1), Carolina Power & Light 4
Company (CP&L) reported to the U.S. Nuclear Regulatory Commission (NRC) that the Unit 1 and j
Unit 2 pressure-temperature (P-T) limits curves had been inadvertently transposed. The Company reported that this condition had resulted in the potential use of less-conservative i
curves for Unit 1 during previous heatup and cooldown evolutions. Supplemental submittals to
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Licensee Event Report 1-94-05 were provided by letters dated April 29,1994 (Reference 2) and September 23,1994 (Reference 3). The supplemental Licensee Event Reports confirmed that the Unit 1 and Unit 2 pressure-temperature limits curves were indeed transposed, provided a j
detailed explanation of the circumstances that led to the transposition of the Unit 1 and Unit 2 pressure-temperature limits curves, and an evaluation of Unit 1 operations that were outside the g
more conservative Unit 2 pressure-temperature limits curves.
Unit 2 was the first reactor constructed at the Brunswick Plant site. General Electric (GE) had i
designated equipment for this unit as Carolina 1. The Unit 1 equipment was designated as Carolina ll. These GE " Carolina" designsWns were employed for designating fabrication / shipping sequences for materials / components at the GE Brunswick project.
Two vessels were constructed for the Brunswick project by Chicago Bridge & Iron (CB&l);
Carolina I (CB&l Contract #68-2471) and Carolina !! (CB&l Contract #68-2472). The Carolina I vessel was originally intended to be installed in Unit 2, and the Carolina ll vessel installed in Unit 1. CB&l also fabricated surveillance specimens for each of the vessels. The surveillance specimens were installed into surveillance baskets and designated as G1, G2, and G3 for Carolina I (#68-2471) vessel and G4, G5, and G6 for the Carolina !! (#68-2472) vessel. The specimens were marked with a binary code representing "38" for Carolina I and "39" for Carolina 11.
In 1971, a decision was made to set the Carolina II (#68-2472) vessel in Unit 2 and Carolina I I
(#68-2471) vesselin Unit 1. In order to accommodate this change, a Field Disposition Instruction (FDI) was issued, directing site personnel to install the surveillance baskets G1, G2, and G3 into vessel #68-2471 and G4, G5, and G6 baskets into vessel #68-2472, consistent with fabrication records.
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The capsule removed from Unit 1 in August 1993 was G1, with a binary code "38", representing i
the Carolina I vessel (#68-2471). However, the GE NEDO documents (NEDO-24161 dated November 1978 and NEDO-24157, Revision 1 dated December 1978) used as the basis of the current Technical Specification pressure-temperature limits curves incorrectly indicated that vessel #68-2471 was in Unit 2 and that vessel #68-2472 was in Unit 1; therefore, GE believed that the G1 capsule, which was associated with vessel #68-2471, should have come from Unit 2.
Review of this issue indicated that the cause of this event was inadequate review of the design documentation used in the development of the GE NEDO documents.
On April 15,1994, CP&L verified the nameplate data of the Unit 2 reactor pressure vessel as CB&l Contract No. 68-2472. This verification confirmed that the materials information contained in the NEDO documents used in development of the Technical Specification pressure-temperature limits curves were reversed. Therefore, the Unit 1 and Unit 2 Technical Specification pressure-temperature limits curves were also reversed.
The proposed amendments do not alter the pressure-temperature limits curves currently found in the Unit 1 and Unit 2 Technical Specifications. Instead, the proposed amendments correct an E1-2
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error that was inadvertently introduced into the Technical Specifications of both Brunswick units.
Reolacement of Hydrostatic Test Pressure-Temoerature Limits Curves:
A review of the current reactor vessel pressure-temperature limits curves contained in the Technical Specifications for hydrostatic pressure tests revealed that the curves are suitable for up to 12 EFPY of reactor operation. The Company anticipates that both Brunswick Unit 1 and Unit 2 will surpass this threshold during 1997. Therefore, new 14 and 16 EFPY hydrostatic test pressure-temperature limits curves for the Brunswick Unit 1 and 2 reactor vessels are provided in l
Enclosures 5 and 6 (proposed Technical Specification Figures 3.4.6.1-3a and 3.4.6.1-3b, respectively) and the 8,10 and 12 EFPY hydrostatic test pressure-temperature limits curves (current Technical Specification Figures 3.4.6.1-3a,3.4.6.1-3b, and 3.4.6.1-3c) are being deleted.
In order to assure adequate safety margins for the structuralintegrity of the reactor coolant pressure boundary during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, operational limits are imposed on the pressure boundary. Appendices G and H to 10 CFR 50 specify the fracture toughness requirements for the ferritic materials included in the reactor coolant pressure boundary and beltline regions.
Appendix G to 10 CFR 50 specifies the pressure-temperature limits and minimum temperature requirements for the reactor vessel defined by operating condition, vessel pressure, whether fuel is in the vessel, and whether the core is critical. Specifically, this appendix requires that the pressure-temperature limits provide safety margins at least as great as those recommended in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI. Appendix H of 10 CFR 50 specifies the requirements for monitoring changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. Regulatory Guide 1.99, Revision 2 provides the calculation procedure for determining the effects of neutron radiation on reactor vessel materials (adjusted reference temperature).
The 14 and 16 EFPY hydrostatic pressure test curves included in this submittal were developed from these criteria. The inclusion of two separate pressure-temperature limits curves (14 and 16 EFPY for hydrostatic and leak tests) minimizes the time necessary to reach the test temperature by taking advantage of the lower adjusted reference temperature (ART) applicable to the vessel limiting material as time progresses. This approach allows for conducting hydrostatic and leak tests in an efficient manner while ensuring reactor coolant system pressures and temperatures are maintained within safe limits. Additionally, the previous CP&L commitment (Reference 4) to use the GE method or other NRC approved methods for Reference Temperature (RTuor) determination (
Reference:
GENE Procedure NEDC-32399-P) has been incorporated. The use of the GE method for RTnor determination has resulted in a change in the initial RTuor for the Unit i vessel flange from 10 F to 16"F. The Unit 2 vessel flange was not affected by the use of the GE method for RTnor determination. Reference Temperature (RTuor) adjustment to the vessel flange to account for neutron irradiation is not required due to the low neutron fluence accumulation in that region. For hydrostatic pressure and leak tests, with fuel in j
the vessel, and vessel pressure less than or equal to 20 percent of vessel pre-service hydrostatic test pressure, the change in the Unit i vessel flange region RT r value had no impact on the uo new pressure-temperature curves. The minimum specified temperature of 70 F is well above the ASME Code,Section XI Appendix G limit (RTsar + Irradiation effects) and minimum temperature requirement from Appendix G to 10 CFR 50 (RTno7). For pressures greater than 20 percent, a limit of RTsar + 90'F was used for the vessel flange region up to the intersection point for the pressure-temperature curve for the limiting component. The limiting components in the beltline region are the N16A/B instrument nozzles. The RT and fluence sar values used for curve construction were taken from Enclosure 3 of the CP&L response to NRC E1-3
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Generic Letter 92-01, Revision 1, Supplement 1 (Reference 5). The limiting material adjusted reference temperature was calculated in accordance with Regulatory Guide 1.99, Revision 2.
j This information, along with the pressure to stress intensity factor relationship previously developed by finite element analysis for the N16 nozzles, was utilized in the development of the j
hydrostatic pressure test curves for 14 and 16 EFPY.
Reformat Fiaures:
Each of the Technical Specification Figures containing the pressure-temperature limits curves is proposed to be reformatted. The changes associated with the reformatting of proposed 1
Technical Specification Figures 3.4.6.1-1, 3.4.6.1-2, 3.4.6.1-3a, and 3.4.6.1-3b reflect presentation preferences which do not result in technical changes (either actual or interpretational) to the requirements of the pressure-temperature limits curves. Therefore, the changes associated with reformatting the Technical Specification Figures containing the j
pressure-temperature limits curves are considered to be administrative in nature.
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REFERENCES
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1.
Letter from J. Cowan (CP&L) to NRC Document Control Desk dated March 22,1994 (Serial: BSEP 94-0115), Licensee Event Report 1-94-005."
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2.
Letter from J. Cowan (CP&L) to NRC Document Control Desk dated April 29,1994 (Serial.
B6EP 94-0165), " Licensee Event Report 1-94-005, Supplement 1."
3.
Letter from J. Cowan (CP&L) to NRC Document Control Desk dated September 23,1994 j
(Serial: BSEP 94-0353), " Supplemental Licensee Event Report 1-94-005."
i 4.
Letter from Roy A. Anderson (CP&L) to NRC Document Control Desk dated May 13,1994 (Serial: BSEP 94-0179), " Response To Request For Additional information Regarding i
Generic Letter 92-01, ' Reactor Vessel Structural Integrity,10 CFR 50.54(f)' - Revision 1,"
(see Commitment 1 in Enclosure 2}.
' 5.
Letter from William R. Campbell (CP&L) to NRC Document Control Desk dated November 16,1995 (Serial: BSEP 95-0572), " Response To NRC Generic Letter 92-01, Revision 1, Supplement 1, Reactor Pressure Vessel Integrity."
6.
Structural Integrity Associates, Inc. Calculation Package CPL-420-302, Revision 1.
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ATTACHMENT 1 l
TO ENCLOSURE 1 1
l BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DESIGN STRUCTURAL INTEGRITY CALCULATION PACKAGE CPL-420-302, REVISION 1 i
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