ML060620205
| ML060620205 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/03/2006 |
| From: | Nerses V NRC/NRR/ADRO/DORL/LPLB |
| To: | Christian D Dominion Nuclear Connecticut |
| Nerses V, NRR//DLPM, 415-1484 | |
| References | |
| TAC MC7593 | |
| Download: ML060620205 (13) | |
Text
May 3, 2006 Mr. David A. Christian Sr. Vice President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: REVISION TO TECHNICAL SPECIFICATIONS PERTAINING TO THE REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITS (TAC NO.
Dear Mr. Christian:
The Commission has issued the enclosed Amendment No. 292 to Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2, in response to your application dated July 14, 2005, as supplemented by letter dated January 11, 2006.
The amendment revises the reactor coolant system heatup and cooldown limits Technical Specification 3.4.9.1, Reactor Coolant System, pressure-temperature limit curves to extend their validity to 54 effective full power years of operation.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA by GMiller for/
Victor Nerses, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosures:
- 1. Amendment No. 292 to DPR-65
- 2. Safety Evaluation cc w/encls: See next page
Mr. David A. Christian Sr. Vice President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: REVISION TO TECHNICAL SPECIFICATIONS PERTAINING TO THE REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITS (TAC NO.
Dear Mr. Christian:
The Commission has issued the enclosed Amendment No. 292 to Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2, in response to your application dated July 14, 2005, as supplemented by letter dated January 11, 2006.
The amendment revises the reactor coolant system heatup and cooldown limits Technical Specification 3.4.9.1, Reactor Coolant System, pressure-temperature limit curves to extend their validity to 54 effective full power years of operation.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA by GMiller for/
Victor Nerses, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosures:
- 1. Amendment No. 292 to DPR-65
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
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LLois RidsAcrsAcnwMailCenter RidsNrrDciCvib RHardies ADAMS Accession Number: ML060620205 TS(s): ML Package:
OFFICE LPL1-2/PM LPL1-2/LA SRXB/SC CVIB/C IROB/SC OGC LPL1-2/BC NAME VNerses CRaynor JNakoski MMitchell TBoyce SHamrick DRoberts DATE 3/22/06 5/01/06 08/26/05 02/28/06 4/12/06 4/21/06 5/02/06 OFFICIAL RECORD COPY
Millstone Power Station, Unit No. 2 cc:
Lillian M. Cuoco, Esquire Senior Counsel Dominion Resources Services, Inc.
Building 475, 5th Floor Rope Ferry Road Waterford, CT 06385 Edward L. Wilds, Jr., Ph.D.
Director, Division of Radiation Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 First Selectmen Town of Waterford 15 Rope Ferry Road Waterford, CT 06385 Charles Brinkman, Director Washington Operations Nuclear Services Westinghouse Electric Company 12300 Twinbrook Pkwy, Suite 330 Rockville, MD 20852 Senior Resident Inspector Millstone Power Station c/o U.S. Nuclear Regulatory Commission P.O. Box 513 Niantic, CT 06357 Mr. J. Alan Price Site Vice President Dominion Nuclear Connecticut, Inc.
Building 475, 5th Floor Rope Ferry Road Waterford, CT 06385 Mr. John Markowicz Co-Chair Nuclear Energy Advisory Council 9 Susan Terrace Waterford, CT 06385 Mr. Evan W. Woollacott Co-Chair Nuclear Energy Advisory Council 128 Terrys Plain Road Simsbury, CT 06070 Ms. Nancy Burton 147 Cross Highway Redding Ridge, CT 00870 Mr. Chris L. Funderburk Director, Nuclear Licensing and Operations Support Dominion Resources Services, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 Mr. David W. Dodson Licensing Supervisor Dominion Nuclear Connecticut, Inc.
Building 475, 5th Floor Rope Ferry Road Waterford, CT 06385
DOMINION NUCLEAR CONNECTICUT, INC.
DOCKET NO. 50-336 MILLSTONE POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 292 License No. DPR-65 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Dominion Nuclear Connecticut, Inc., the licensee, dated July 14, 2005, as supplemented by letter dated January 11, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-65 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 292, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Darrell J. Roberts, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 3, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 292 FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert XII XII 3/4 4-19 3/4 4-19 3/4 4-19a 3/4 4-19a 3/4 4-19b 3/4 4-19b
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 292 TO FACILITY OPERATING LICENSE NO. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336
1.0 INTRODUCTION
By application dated July 14, 2005, as supplemented by letter dated January 11, 2006, Dominion Nuclear Connecticut, Inc. (DNC or the licensee) requested Nuclear Regulatory Commission (NRC or the Commission) approval of changes to the Millstone Power Station, Unit No. 2 (MPS2) Technical Specifications (TSs). The changes would modify the reactor coolant system (RCS) heatup and cooldown TSs pressure-temperature (PT) limit curves to extend their validity to 54 effective full power years (EFPYs) of operation.
The current PT limits are valid to 20 EFPYs. However, the proposed PT limits incorporate changes related to materials and Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix G requirements. Specifically, the proposed changes revise TS 3.4.9.1 Reactor Coolant System heatup and cooldown limits. The neutron fluence portion of the proposed changes are based on the results of surveillance capsule W-83 that was removed, tested, and reported in WCAP-16012, February 2003 (Reference 1).
The supplement dated January 11, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 30, 2005 (70 FR 51379).
2.0 REGULATORY EVALUATION
The NRC has established requirements in 10 CFR Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The NRC staff evaluates the PT limit curves based on the following NRC regulations and guidance: 10 CFR Part 50, Appendix G; Generic Letter (GL) 88-11; GL 92-01, Revision 1; GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) 1.99, Revision 2 (Revision 2); and Standard Review Plan (SRP)
Section 5.3.2. Appendix G to 10 CFR Part 50 requires that PT limit curves be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).
Appendix G to 10 CFR Part 50 also provides minimum temperature requirements that must be considered in the development of the PT limit curves. GL 88-11 advised licensees that the NRC staff would use RG 1.99, Revision 2 to review PT limit curves. RG 1.99, Revision 2 contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. GL 92-01, Revision 1 requested that licensees submit their reactor pressure vessel (RPV) materials property data for their plants to the NRC staff for review. GL 92-01, Revision 1, Supplement 1 requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations. This data is used by the NRC staff as the basis for the review of PT limit curves.
SRP Section 5.3.2 provides an acceptable method of determining the PT limit curves for ferritic materials in the beltline of the RPV based on the linear elastic fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME Code. The basic parameter of this methodology is the stress intensity factor KI, which is a function of the stress state and flaw configuration. ASME Code,Section XI, Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 on these stress intensities for hydrostatic testing curves. The flaw postulated in the ASME Code,Section XI, Appendix G has a depth that is equal to 1/4 of the RPV beltline thickness and a length equal to 1.5 times the RPV beltline thickness. The critical locations in the RPV beltline region for calculating heatup and cooldown PT limit curves are the 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively.
The methodology in Appendix G to Section XI of the ASME Code requires that licensees determine the adjusted reference temperature (ART or adjusted RTNDT) by evaluating material property changes due to neutron radiation. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT), the mean value of the adjustment in reference temperature caused by irradiation (RTNDT) and a margin term. The RTNDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Revision 2 or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Revision 2 or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the fluence and the calculational procedures. RG 1.99, Revision 2 describes the methodology to be used in calculating the margin term.
The NRC staff issued RG 1.190 Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (Reference 2). The RG outlines methods, associated approximations, and benchmarking acceptable to the NRC staff for the calculation of pressure vessel fluence. The proposed changes are based on the results of MPS2 surveillance capsule W-83. In Reference 1, the calculations, methods and practices follow the guidance in RG 1.190 and, therefore, are acceptable.
3.0 TECHNICAL EVALUATION
3.1 Licensees Evaluation The licensee stated in its submittal that Appendix G to Section XI of the 2001 Edition with 2002 Addenda of the ASME Code was used in generating the PT limit curves.
The changes submitted by the July 14, 2005, letter and as amended January 11, 2006, include:
Modified estimates of pressure and temperature measurement uncertainty, A fluence estimate that is applicable to 54 EFPYs, and A calculational methodology change from the method required by Appendix G of the 1989 Edition (no Addenda) of the ASME Code to the method currently required by Appendix G of the 2002 Edition of the ASME Code.
The modified estimates of pressure and temperature measurement uncertainty reflect a decrease in the uncertainty associated with monitoring RCS pressure from 86.6 pounds per square inch (psi) to 80.2 psi (for zero reactor coolant pumps) and an increase in the uncertainty associated with monitoring RCS temperature from 10.5 to 13 EF. The licensee indicates these changes will have no impact on plant safety.
The fluence estimate was revised based on analysis of surveillance capsule W-83 and an associated fluence analysis which considered the actual core history and core loading patterns projected through 54 EFPYs.
The licensee stated that the development of the beltline P-T limits was established using ASME Code Section Xl, Appendix G, 2002 Addenda. This edition of the ASME Code, approved for use in the 2004 Edition of 10 CFR Part 50, provides a reference fracture toughness curve (KIC) for establishment of the beltline P-T limits.
The licensee stated that a revision to the existing low temperature overpressure protection (LTOP) evaluation determined the existing LTOP administrative limits described in the TSs are still acceptable to protect the new proposed P-T limits and associated rates. Therefore, the licensee did not request a change to LTOP administrative limits.
ASME Code Section Xl, Appendix G requirements have changed. Earlier versions of ASME Code Section Xl, Appendix G did not provide guidance for determination of the LTOP enable temperature. Therefore, the licensee used Branch Technical Position RSB 5-2 of NUREG-0800 for guidance in earlier LTOP calculations. The currently-approved version of ASME Code Section Xl, Appendix G now provides methodology and requirements for calculation of the LTOP enable temperature. Therefore, the licensee stated the supporting analysis for this submittal was performed in accordance with the currently-approved version of ASME Code Section Xl, Appendix G, and reference to Branch Technical Position RSB 5-2 was no longer required.
3.2 NRC Staffs Evaluation 3.2.1 Instrument Uncertainties The NRC staff concurs with the licensees assessment that the small changes in instrument uncertainty will have no adverse affect on plant safety.
3.2.2 ART Value and PT Limit Curves To assess the validity of the licensees proposed curves, the NRC staff performed an independent assessment of the licensees submittal. The NRC staff first performed an independent calculation of the ART values for the limiting material using the methodology in RG 1.99, Revision 2. Based on these calculations, the NRC staff verified that the licensees limiting material is plate C-506-1. The NRC staffs calculated ART values of 175 EF at 1/4 T and 144 EF for the 3/4T location for the limiting material, using information in the NRC Reactor Vessel Integrity Database, validated the licensees calculated ART values.
The NRC staff then evaluated the licensees PT limit curves for acceptability by performing independent calculations using the methodologies of Appendix G of Section XI of the ASME Code and 10 CFR Part 50, Appendix G. The NRC staff concluded that the PT limit curves were calculated correctly. The NRC staff also found that the minimum temperature requirements of Table 1 of Appendix G to 10 CFR Part 50 were properly implemented in the PT limit curves.
Therefore, the NRC staff verified that the licensees proposed PT limit methodology is in accordance with Appendix G to Section XI of the ASME Code and the proposed PT limits satisfy the requirements of Appendix G to 10 CFR Part 50.
Issues related to RPV neutron fluence calculation were discussed in a NRC staff safety evaluation dated August 26, 2005. The NRC staff determined that the fluence methodology adheres to the guidance of RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001. The RPV neutron fluence is discussed in further detail in Section 3.2.4.
3.2.3 LTOP The NRC staff reviewed the licensees application of the requirements of Appendix G to Section XI of the ASME Code to evaluate the licensees LTOP setpoints. The NRC staff concluded that the licensees application of the requirements of Appendix G to Section XI of the ASME Code was appropriate.
3.2.4 Surveillance Capsule Surveillance capsule W-83 was removed at 15.3 EFPYs of operation and received an average exposure of 1.74x1019 n/cm2. The measured dosimetry values (M) and the corresponding calculated values (C) were analyzed in the form of M/C, i.e., statistically calculated a best estimate (BE) value and compared M/C and M/BE to the performance of surveillance capsules W-97 and W-104 that were removed, measured, and analyzed previously. The dosimeter complement included a Cadmium-covered U-238 dosimeter.
The measured value after correction for a U-235 impurity and U-238(, f) was judged inconsistent with comparable measured data and was rejected from the measurement package.
This does not diminish the performance or the adequacy of the remaining dosimeters because the major threshold dosimeters (i.e., Fe and Ni) yielded excellent results. The maximum deviation of the Fe and Ni M/C values are within 3%. The Cd covered Co dosimeter showed deviation from the calculated value but it was judged to be acceptable. The remaining dosimeters show reasonable agreement of measured-to-calculated values. The plant-specific measurements provided excellent benchmarking, attesting to the viability of the calculation.
The calculated peak vessel inner-radius 54 EFPYs fluence value is 2.40x1019 n/cm2.
In summary, the licensee proposed to change the PT limit curves using the projected fluence value from the results and calculations performed during the analyses of surveillance capsule W-83. The NRC staff review finds that the calculations were performed following the guidance in RG 1.190, and are in good agreement with the plant-specific measurements. Therefore, the proposed PT limit curves are acceptable.
3.2.5 Summary The NRC staff considers that the proposed PT limits curves for MPS2 satisfy the requirements in Appendix G to 10 CFR Part 50 and Appendix G to Section XI of the ASME Code. The proposed PT limit curves also satisfy GL 88-11, because the methodology in RG 1.99, Revision 2 was used to calculate the ART. Furthermore, the fluence methodology adheres to the guidance of RG 1.190; therefore, the fluence value that the licensee used is acceptable for the purpose of developing PT limit curves. Therefore, the proposed PT limit curves may be incorporated into the MPS2 TSs and are valid through 54 EFPYs.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The Connecticut State official agreed with the NRC staffs conclusion as stated in Section 6 of this Safety Evaluation.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes selective administrative controls within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (70 FR 51379). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
1 WCAP-16012, Revision 0, Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program, by J. H. Ledger, et al. Westinghouse Electric Company LLC, February 2003.
2 Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, March 2001.
Principal Contributors: R. Hardies L. Lois Date: May 3, 2006