ML20209H305
| ML20209H305 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/15/1999 |
| From: | Dugger C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137X504 | List: |
| References | |
| W3F1-99-0087, W3F1-99-87, NUDOCS 9907200184 | |
| Download: ML20209H305 (10) | |
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Enti y Operztions, Inc.
Killona, LA 700G6-07S1 Tel 504 739 6G00 ce Pre d r r ons Waterford 3 W3F1-99-0087 A4.05 PR ATTACHMENT CONTAINS PROPRIETARY INFORMATION July 15,1999 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-218 Extend Pressure Temperature Curve to 20 EFPY Gentlemen:
In accordance with 10CFR50.90, Entergy is hereby proposing to amend Operating License NPF-38 for Waterford 3 by requesting the attached changes to the Technical Specifications (TS). The attached description and safety analysis support the proposed changes to the Waterford 3 TS. The proposed changes modify TS 3/4.4.8.1, Figures 3.4-2 and 3.4-3 to extend the Reactor Coolant System Pressure Temperature Curves to 20 Effective Full Power Years (EFPY). No change to the TS Bases is required with this change.
By letter dated December 14,1993, Entergy proposed TS Change Request NPF 148 for Waterford 3. The proposed change included revision of Reactor Coolant System (RCS) pressure-temperature (PT) curves from 0 to 8 EFPY to 0 to 20 EFPY.
Based on NRC Staff comments, Entergy submitted a letter dated March 3,1995 requesting that the proposed 20 EFPY on TS Figures 3.4-2 and 3.4-3 be modified to 15 EFPY. By letter dated May 18,1995, the NRC Staff issued the TS Change Request as Amendment 106, for 0-15 EFPY.
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Technical Specification Change Request NPF-38-218 Extend Pressure Temperature Curve to 20 EFPY W3F1-99-0087 Page 2 July 15,1999 This proposed change has been evaluated in accordance with 10CFR50.91(a)(1),
using the criteria in 10CFR50.92(c), and it has been determined that this request involves no significant hazards consideration.
Please note that' Attachment D, B&W report 51-1234900-00, Fluence Uncertainty Information For Extending Waterford Unit 3 P-T Limits to 20 EFPY, contains information that is considered proprietary pursuant to 10 CFR 2.790. In this regard, Entergy requests that this Attachment be withheld from public viewing. The j
respective B&W affidavit pursuant to 10 CFR 2.790 is enclosed.
The circumstances surrounding this change do not meet the NRC's criteria for
- exigent or emergency review; however, the current PT curves (0-15 EFPY) are expected to expire in 2003 assuming 100% power operation. If surveillance capsules need to be pulled to extend the PT limits to 20 EFPY, they have to be pulled during Refuel 10 (September 15,2000).- This schedule will allow one year to
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test and evaluate the capsule results and 18 months to obtain NRC Staff approval of the TS change (by Spring 2003). In order that Entergy have adequate time to make preparations for specimen removal during Refueling Outage 10, if required, approval of this TS change request must be received by March 15,2000. Therefore, Entergy respectfully requests that these changes be processed accordingly. Entergy requests the effective date for this TS. change be within 60 days of approval.
Should you have any questions or comments concerning this request, please contact Everett Perkins at (504) 739-6379 or Curt Taylor at (504) 739-6725.
Very truly yours, y
C.M. Dugger Vice President, Operations Waterford 3:
CMD/CWThtk Attachments:
Affidavit NPF-38-218
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Technical Specification Change Request NPF-38-218 Extend Pressure Temperature Curve to 20 EFPY W3F1-99-0087 Page 3 July 15,1999 cc:
E.W. Merschoff, NRC Region IV C.P. Patel, NRC-NRR J. Smith N.S. Reynolds i
NRC Resident inspectors Office.
Administrator Radiation Protection Division (State of Louisiana)
American Nuclear insurers t
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UNITED STATES OF AMERICA i
NUCLEAR REGULATORY COMMISSION in the matter of
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Entergy Operations, Incorporated
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Docket No. 50-382 Waterford 3 Steam Electric Station
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AFFIDAVIT Charles Marshall Dugger, being duly sworn, hereby deposes and says that he is Vice President Operations - Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Technical Specification Change Request NPF-38-218; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
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' Charles Marshall Dugger' OV Vice President Operations - Waterford 3 l
STATE OF LOUISIANA
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Subscribed and sworn to before me, a Npta)ry Public in and for the Parish and State above named this
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DESCRIPTION AND NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION OF PROPOSED CHANGE NPF-38-218 Summary of Proposed Changes The proposed change modifies the Limiting Condition for Operation for Technical Saecification (TS) 3/4.4.8.1 and Figures 3.4-2 and 3.4-3 to extend the Reactor Coolant System Pressure Temperature Curve to 20 Effective Full Power Years. No changes to the TS Bases are required for this change.
Existing Specification See Attachment A Proposed Marked-up Specification See Attachment B Proposed Specification See Attachment C References See Attachment D
- 1. Affidavit Pursuant to 10CFR2.790 Regarding Proprietary Information
- 2. B&W report 51-1234900-00, Fluence Uncertainty Information For Extending Waterford Unit 3 P-T Limits to 20 EFPY
Background
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide TS for the operation of the plant. In particular,10 CFR 50.36(c)(2) j requires that limiting conditions of operation be included in the TS. The Pressure Temperature (PT) limits are among the limiting conditions of operation in the TS for all commercial nuclear plants in the U.S. Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material l
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1 surveillance that must be considered in setting PT limits. An acceptable method for constructing the PT limits is described in Standard Review Plan Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards for conduct of the tests. These tests define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature.
Appendix G also requires the licensee to account for the effects of neutron irradiation on vessel embrittlement and Charpy upper shelf energy (USE). Generic Letter 88-11,"NRC Position On Radiation Embrittlement Of Reactor Vessel Materials And Its impact On Plant Operations," requested that licensees and permitter use the methods in Regulatory Guide (RG) 1.99, Rev. 2, " Radiation Embritti meat of Reactor Vessel Materials," to predict the effect of neutron irradiation on react r 4essel materiais. This guide defines the Adjusted Reference Temperature (ART) as the sum of unirradiated reference temperature, the increase in reference temperature resu! ting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone materials of the reactor beltline.
By letter dated December 14,1993, Entergy proposed TS Change Request NPF 148 for Waterford 3. The proposed change included revision of Reactor Coolant System (RCS) pressure-temperature (PT) curves from 0 to 8 EFPY to 0 to 20 EFPY.
The revised curves were based on ABB Combustion Engineering (CE) Report C-MECH-ER-021, Rev. 00, that was included with the initial request, and Babcock and Wilcox (B&W) Report BAW-2177 (Analysis of Capsule W-97), submitted for NRC Staff review in November of 1992.
The NRC Staff requested additionalinformation by letter dated January 11,1995, concerning the methods for predicting fluence and the uncertainty associated with the measurements and calculations used in the fluence prediction. The neutron fluence results from BAW-2177 were used by ABB CE to develop RCS PT limits for 20 EFPY. The specific concern was the projection of the fluence value and associated PT limits to 20 EFPY, based on the analysis results of one surveillance capsule.
To resolve this issue, Entergy submitted a letter dated March 3,1995 requesting that the proposed 20 EFPY on TS Figures 3.4-2 and 3.4-3 be modified to 15 EFPY. The proposed PT curves (excluding the title blocks) were otherwise unchanged from the 2
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original request. The modified curves were still based on a 0 to 20 EFPY peak surface fluence of 2.29 x 10" neutrons per square centimeter (n/cm2). By letter dated May 18,1995, the NRC Staff issued the TS Change Request as Amendment 106, for 0 to 15 EFPY.
Description and Safety Considerations Waterford 3 proposes to amend its Operating License by modifying the Limiting Condition for Operation for Technical Specification (TS) 3/4.4.8.1 and Figures 3.4-2 and 3.4-3 to extend the Reactor Coolant System (RCS) Pressure Temperature (PT)
Curve from 15 Effective Full Power Years (EFPY) to 20 EFPY.
Changes in the RCS heatup and cooldown limits were evaluated for extension of the PT curves to 20 EFPY during the development of TS Change Request NPF-38-148 and it was determined that no changes were required. Therefore, no changes to the heatup and cooldown limits are being proposed in this request for the extension of the PT curve to 20 EFPY. Additionally, the low temperature overpressure protection (LTOP) enable temperature of 285 degrees F was evaluated at that time for 20 EFPY and was revised to 272 degrees F. Therefore, no changes to the LTOP are required for revision of the PT curve to 20 EFPY.
- Subsequent to the submittals discussed in the " Background" section of this request, Entergy contracted with B&W to perform an uncertainty analysis on the fluence. The analysis is documented in B&W report 51-1234900-00, Fluence Uncertainty Information For Extending Waterford Unit 3 P-T Limits to 20 EFPY and is provided as an Attachment to this request. The report determined that the greater than 1.0 MeV (million electron volts) fluence has a standard deviation of 7.0 % Since the fluence uncertainty is less than 20 %, the fluence predictions in BAW-2177 are valid for embrittlement analysis of the Waterford 3 reactor vessel using the formulations in Regulatory Guide 1.99, Revision 2. Therefore, there are no fluence uncertainty concerns with the Waterford 3 PT limits being valid for 20 EFPY with the projected fluence value of 2.29 x 10" n/cm2. This report addresses the NRC Staff's specific questions asked during their review of TSCR NPF-38-148.
No change to the TS Bases is required for this change. Amendment 106 revised the Bases for the Pressure Temperature Limits 3/4.4.8 to state that the RCS will be protected from pressure transients caused by starting a reactor coolant pump which could exceed 10 CFR 50 Appendix G limits. This was applicable when the RCS cold legs are less than or equal to 285 degrees F. Amendment 106 changed this value to 272 degrees F.
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1 No Significant Hazards Consideration Determination The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
The proposed changes will not increase the probability or consequences of any accident previously evaluated since the proposed changes revise the pressure / temperature limits in accordance with 10CFR50, Appendix G utilizing the latest NRC guidelines in Regulatory Guide 1.99, Revision 2 relative to estimating neutron irradiation damage to the reactor vessel. The proposed changes also maintain the conservative limits with respect to the low temperature overprotection (LTOP) system and heatup and cooldov 1 restrictions.
Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
The proposed changes will not create the possibility of a new or different kind of accident from any previously analyzed since they do not introduce new systems, failure modes, or other plant perturbations. The proposed changes revise the pressure / temperature limits in accordance with 10CFR50, Appendix G utilizing the latest NRC guidelines in Regulatory Guide 1.99, Revision 2 relative to estimating neutron irradiation damage to the reactor vessel.
Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
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Response
The proposed changes will not involve a significant reduction in the margin of safety since equal or more stringent pressure / temperature limitation requirements for reactor operation will be applied. The proposed changes i
were derived in accordance with approved NRC methodology which was developed to assure the reactor coolant system pressure boundary is designed with sufficient margin to withstand any condition during normal operation including anticipated operational occurrences and system inservice leak and hydrostatic tests.
These requirements were revised in accordance with 10CFR50, Appendix G utilizing the latest NRC guidance in Regulatory Guide 1.99, Revision 2 relative to estimating neutron irradiation damage to the reactor vessel. The LTOP system limits were also reanalyzed for the proposed changes.
Therefore, the proposed change will not involve a significant reduction in a margin of safety.
Safety and Significant Hazards Determination Based on the above No Significant Hazards Evaluation, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92: and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC final environmental statement.
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ATTACHMENT A EXISTING SPECIFICATIONS