ML053550091
| ML053550091 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/03/2006 |
| From: | Stephen Monarque NRC/NRR/ADRO/DORL/LPLC |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| Monarque S | |
| References | |
| TAC MC5509, TAC MC5510 | |
| Download: ML053550091 (15) | |
Text
January 3, 2006 Mr. David A. Christian Sr. Vice President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, Virginia 23060-6711
SUBJECT:
SURRY POWER STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENTS ON REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS (TAC NOS. MC5509 AND MC5510)
Dear Mr. Christian:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-32 and Amendment No. 244 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments change the Technical Specifications in response to your application dated December 17, 2004.
These amendments revise the reactor coolant pressure and temperature limits, low-temperature overpressure protection system (LTOPS) setpoint values, and LTOPS enable temperatures for up to 47.6 effective full-power years (EFPY) and 48.1 EFPY of operation at Surry Power Station, Unit Nos. 1 and 2, respectively.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Stephen Monarque, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281
Enclosures:
- 1. Amendment No. 245 to DPR-32
- 2. Amendment No. 244 to DPR-37
- 3. Safety Evaluation cc w/encls: See next page
ML053550091 NRR-058 OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/CVIB/B C
NRR/SPWB/BC OGC NRR/LPL2-1/BC NAME SMonarque MOBrien MMitchell JNakoski MBupp EMarinos DATE 12/21/05 12/21/05 11/23/2005 11/23/2005 12/14/2005 12/21/05
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 245 Renewed License No. DPR-32 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated December 17, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 Commission's 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 Commission's 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 3.B of Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:
(B)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 245, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 180 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 3, 2006
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 244 Renewed License No. DPR-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated December 17, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 Commission's 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 Commission's 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 3.B of Renewed Facility Operating License No. DPR-37 is hereby amended to read as follows:
(B)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 244, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 180 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 3, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 245 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 LICENSE AMENDMENT NO. 244 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 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 Pages Insert Pages 3.1-9 3.1-9 3.1-10 3.1-10 3.1-11 3.1-11 3.1-23a 3.1-23a Figure 3.1-1 Figure 3.1-1 Figure 3.1-2 Figure 3.1-2
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 245 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 244 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281
1.0 INTRODUCTION
By letter dated December 17, 2004, Virginia Electric and Power Company (the licensee) submitted license amendments to change the Technical Specifications (TSs) for Surry Power Station, Unit Nos. 1 and 2 (Surry 1 and 2). The proposed amendments would provide new reactor coolant system (RCS) pressure-temperature (P-T) limit curves, low-temperature overpressure protection system (LTOPS) setpoints, and LTOPS enable temperatures (Tenable).
The licensee submitted these proposed TSs changes to replace the current P-T limits that are valid for 28.8 effective full-power years (EFPY) and 29.4 EFPY of operation at Surry 1 and 2, respectively, with 47.6 EFPY and 48.1 EFPY for Surry 1 and 2, respectively. The calculations for the revised P-T limit curves are delineated in WCAP-15130, Revision 1, Surry Units 1 and 2 WOG [Westinghouse Owners Group] Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Curves for Normal Operation, April 2001.
2.0 REGULATORY EVALUATION
The Nuclear Regulatory Commission (NRC) staff evaluated the acceptability of the licensees proposed P-T limits based on the following regulations and guidance:
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 50.60(a) states:
Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the certifications required under
§50.82(a)(1) have been submitted, must meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in appendices G and H to this part.
Appendix H to 10 CFR Part 50, Reactor Vessel Material Surveillance Program Requirements, establishes requirements related to facility reactor pressure vessel (RPV) material surveillance programs.
Regulatory Guide (RG) 1.190, March 2001, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence.
RG 1.99, Revision 2, May 1988, Radiation Embrittlement of Reactor Vessel Materials, contains methodologies for determining the increase in transition temperature resulting from neutron radiation.
Appendix G to 10 CFR Part 50, Fracture Toughness Requirements, requires that facility P-T limit curves for the RPV be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society for Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code). The most recent version of Appendix G to Section XI of the ASME Code, which has been endorsed in 10 CFR 50.55a, and therefore by reference in 10 CFR Part 50, Appendix G, is the 2001 edition through the 2003 addenda of the ASME Code. This edition of Appendix G to Section XI incorporated the provisions of ASME Code Case N-641. Code Case N-641 presents alternative procedures for calculating P-T relationships and LTOPS system effective temperatures and allowable pressures. This procedure takes into account alternative fracture toughness properties, circumferential and axial reference flaws, and plant-specific LTOPS Tenable calculations. Additionally, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20 percent of the preservice hydrostatic test pressure.
Generic Letter (GL) 92-01, Revision 1, dated March 6, 1992, Reactor Vessel Structural Integrity, 10 CFR 50.54(f), requested that licensees submit the RPV data for their plants to the NRC staff for review, and GL 92-01, Revision 1, Supplement 1, dated May 19, 1995, requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations.
NUREG-0800, Standard Review Plan (SRP), Section 5.3.2, Pressure Temperature Limits, provides guidance on using these regulations and documents in the NRC staffs review.
Additionally, Section 5.3.2 provides guidance to the NRC staff in performing check calculations of the licensees submittal.
3.0 TECHNICAL EVALUATION
3.1 Licensees Evaluation In its submittal dated December 17, 2004, the licensee submitted adjusted reference temperature (ART) values and P-T limit curves valid for up to 47.6 EFPY and 48.1 EFPY of facility operation at Surry 1 and 2, respectively. Furthermore, the licensee identified the following limiting material for the RPVs at Surry 1 and 2.
Summary of Limiting ART Values for Surry 1 and 2 EFPY 1/4T Limiting ART (0 F) 3/4T Limiting ART (0 F) 47.6 (Unit 1) 238.2 183.9 48.1 (Unit 2) 216.5 183.7 The Surry 1 lower shell longitudinal weld L2, SA-1526/299L44 is the limiting material located at the 1/4T location. The intermediate-to-lower shell circumferential welds, SA-1585/72245 and SA-1650/72445 at Surry 1 are the limiting materials at the 3/4T location. The licensee used the limiting ART values of 238.2 0F at 1/4T and 183.9 0F at 3/4T to generate the bounding P-T limits for both Surry 1 and 2.
WCAP-15130, Revision 1, documented detailed thermal and fracture mechanics evaluations to establish the proposed P-T limits at Surry 1 and 2. The RPV temperatures at the 1/4T and 3/4T locations for 60 0F/hour heatup and 100 0F/hour cooldown can be found in Appendix E of the licensees application dated December 17, 2004. Based on the temperature distribution across the RPV wall, the licensee derived the thermal stresses and the applied thermal stress intensity factors (KIt), for the tip of the postulated flaw at the 1/4T location, for the 100 EF cooldown transient and, for the 1/4T and 3/4T locations, for the 60 EF heatup transient. Based on these applied KIt and the reference stress intensity factors (KIC) at the crack tips, the licensee calculated the corresponding applied pressure stress intensity factors (KIp) at the tip of the postulated flaw at the 1/4T and 3/4T locations, and the pressure itself. The licensee stated that the proposed P-T limit methodology, as applied to RPV beltline materials, is in accordance with Appendix G of Section XI of the ASME Code.
By letter dated December 17, 2004, the licensee proposed the following TSs changes:
(1) Replace TS Figures 3.1-1 and 3.1-2, Reactor Coolant System Heatup and Cooldown Limitations, and (2) Revise TS 3.1.G.1.c (4) to change the PORV lift setting from 390 psig to 395 psig.
3.2
NRC Staff Evaluation
3.2.1 Neutron Fluence The licensee has a plant-specific vessel fluence methodology as shown in Topical Report VEP-NAF-3A, Reactor Vessel Fluence Analysis Methodology, November 1997, which was approved by the NRC on April 13, 1999. This methodology was approved before the issuance of RG 1.190. At that time, Draft Guide (DG) 1035 was acceptable in lieu of the RG because it contained the most important points of the future RG. By letter dated July 1, 2004, as supplemented by letters dated October 28, and November 16, 2004, the licensee requested a license amendment in order to revise the P-T, LTOPS, and LTOPS Tenable values for North Anna Power Station, Unit Nos. 1 and 2 (North Anna 1 and 2). During the NRC staffs review of this license amendment application for North Anna 1 and 2, the NRC staff found that the licensees methodology followed the guidance of DG 1035, the technical contents of which did not change upon issuance of RG 1.190. Although the October 28, 2004, letter was a response for a request for additional information pertaining to North Anna 1 and 2, the NRC staff found that the vessel fluence methodology also applied to Surry 1 and 2, and this methodology was benchmarked on dosimetry measurements from the North Anna and Surry units. The NRC staff approved of this license amendment for North Anna 1 and 2 on July 8, 2004.
By letter dated May 29, 2001, as supplemented October 15, 2002, the licensee applied for a license renewal at Surry 1 and 2 and North Anna 1 and 2. In its submittals, the licensee stated that the Surry 1 longitudinal weld L2 (SA-1526), which contains an end of renewed license 1/4T RTNDT value of 238.2 EF, was the limiting material for Surry 1 and 2. In its letter dated October 15, 2002, the licensee informed the NRC staff that the inner surface fast neutron (E > 1.0 MeV) is 0.79x1019 n/cm2 and the 1/4T value is 0.482x1019 n/cm2. The P-T limit curves, the LTOPS setpoint and the LTOPS Tenable were calculated to be consistent with the 1/4T RTNDT value of 238.2 EF. These values include the pressure overshoot for the mass addition transient analysis, which is the same as the original analysis. None of the parameters affecting overshoot pressure is affected by the changes proposed here, thus, they remain the same. The NRC staff approved of the renewal application for Surry 1 and 2 on November 5, 2002.
In order to implement these changes, effective to the end of the extended license for Surry 1 and 2, the licensee changed the TS applicability limits to 47.6 and 48.1 EFPYs for Surry 1 and 2, respectively. Similarly, the LTOPS setpoint and the LTOPS Tenable TSs were modified to reflect the new applicability limits. The licensee changed the P-T limit curves to comply with the new higher RTNDT value and these changes are reflected in the TSs.
The NRC staff has determined that the licensee developed the P-T limit curves, the LTOPS setpoint and the LTOPS Tenable using fluence values that were already approved by the NRC staff on November 5, 2002 and July 8, 2004. As such, the NRC staff finds the licensees proposed changes to be acceptable.
3.2.2 LTOPS Setpoint and Tenable Values - Measurement Uncertainties The LTOPS ensures that the reactor coolant system (RCS) material integrity limits are not exceeded during design-basis accidents. The LTOPS setpoints are the pressures at which the power-operated relief valves (PORVS) will lift when the LTOPS is enabled to limit the RCS pressure within the acceptable limits during a pressurization transient. The licensee stated that the LTOPS Tenable value was determined using ASME Code,Section XI and ASME Code Case
-641.
The licensee, in its submittal dated December 17, 2004, calculated the channel statistical accuracy (CSA) for the RCS pressure measurement uncertainty to be 2.05 percent of a 0 psig to 1000 psig instrument span. This value is to be used in establishing the LTOPS setpoint. For conservatism, the licensee used a value of 25 psi for the development of the LTOPS setpoint.
This uncertainty reflects the narrow range RCS pressure uncertainty for the actuation of the PORV bistables. Narrow range RCS pressure is measured in the RCS hot leg.
The licensee derived the RCS pressure measurement uncertainty CSA for the P-T limits from the wide-range instrument, which includes the indication uncertainty for the 0 psig to 3000 psig. This uncertainty was determined to be 2.213 percent or 67 psi. The wide-range pressure measurement uncertainty is accounted for in the RCS hot leg.
The licensee calculated the CSA for the wide-range RCS temperature measurement uncertainty to be 2 percent of a 0 EF to 700 EF instrument span. For conservatism, the licensee used a value of 20 EF in the development of the LTOPS Tenable. The wide-range RCS temperature is measured in the RCS cold leg. The wide-range RCS temperature is used for confirming RCS temperature during normal operation heatup and cooldown, and as input for the LTOPS Tenable.
The licensee determined the pressure difference between the point of measurement (narrow range RCS pressure) and the point of interest (reactor vessel beltline) to be 57 psi. The licensee applied this value as a bias to measure RCS pressure in order to simulate pressure measurement at the reactor vessel beltline.
The NRC staff finds the licensees values to be reasonable, within the expected range of values, and in compliance with the requirements of Appendix G to ASME Code,Section XI, as modified by Code Case -641. As such, the NRC staff finds these changes acceptable.
3.2.3 PORV Lift Setpoint In its application, the licensee has proposed to revise the PORV lift setting value of 390 psig to 395 psig. The setpoint was derived using the method described in Topical Report WCAP-14040P-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoint and RCS Heatup and Cooldown Limit Curves, Revision 4, May 2004. This method has been reviewed and approved by the NRC staff. The setpoint value is based on the results of the heat and mass addition transients. The calculated value corresponds to 327.9 EF. The TS values are 395 psig setpoint and 350 EF for the Tenable. Because the licensee used an NRC staff-approved methodology and these proposed changes are bounded by this methodology, the NRC staff finds these values conservative and this proposed TS change acceptable.
3.2.4 ART Value and P-T Limit Curves To assess the validity of the licensees proposed P-T curves, the NRC staff performed an independent calculation of the ART values for the limiting materials of the RPVs at Surry 1 and 2 using the methodology in RG 1.99, Revision 2. Based on these calculations, the NRC staff verified that the limiting materials for the RPVs are the Surry 1 lower shell longitudinal weld L2, SA-1526/299L44 at the 1/4T location and Surry 1 intermediate-to-lower shell circumferential welds, SA-1585/72245 and SA-1650/72445 at the 3/4T location. The NRC staffs ART values for the limiting materials at the 1/4T and 3/4T locations were calculated using materials information for Surry 1 and 2 contained in the NRC reactor vessel integrity database. The ART values calculated by the NRC staff are in agreement with the licensees calculated values.
The NRC staff then evaluated the licensees P-T limit curves for acceptability by performing independent calculations using the methodology referenced in Appendix G to Section XI of the ASME Code (as indicated by SRP Section 5.3.2) based on information submitted by the licensee.
The licensee stated that the proposed P-T limits were based on the use of the ASME KIC curve instead of the reference stress intensity curve (KIa) for the RPV materials in the P-T limit calculations. Use of the KIC stress intensity formulation is allowed by 10 CFR Part 50, Appendix G. The NRC staff verified that the licensees proposed P-T limit methodology is in accordance with Appendix G to Section XI of the ASME Code and the licensees proposed P-T limits satisfy the requirements of Appendix G to 10 CFR Part 50.
In addition to beltline materials, Appendix G to 10 CFR Part 50 also imposes minimum temperature requirements for the closure head flange based on the most-limiting reference temperature for the flange material.Section IV.A.2 of 10 CFR Part 50, Appendix G, states that when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the temperature of the closure flange regions that are highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120 EF for normal operation and by 90 EF for hydrostatic pressure tests and leak tests. WCAP-15130, Revision 1, reported that the most-limiting reference temperature for the flange material is 10 EF. So the minimum allowable temperature of this region is 130 0F at pressures greater than 621 psig.
These minimum temperature limits are reflected appropriately in the proposed heatup and cooldown and hydrostatic/leak test curves. Therefore, the licensees proposed P-T limit curves are acceptable for operation of the Surry 1 and 2 RPVs through the end of the renewed license.
The NRC staff has reviewed the licensees application regarding the proposed P-T limit curves.
Based on the above evaluations, the NRC staff concludes that the proposed P-T limit curves for the pressure test, core not critical, and core critical conditions satisfy the requirements of Appendix G to 10 CFR Part 50 and the requirements of Appendix G to the ASME Code,Section XI, as modified by Code Case -641. Hence, the proposed P-T limits are approved for incorporation into the TSs for Surry 1 and 2. In addition, these P-T limits are valid for 47.6 and 48.1 EFPY of facility operation for Surry 1 and 2, respectively.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Virginia State official was notified of the proposed issuance of the amendments. The State official had no comment.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (70 FR 9999). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22©)(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 amendments.
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 Commission's regulations, and (3) the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
- 1.
Letter from L.N. Hartz, Virginia Electric and Power Company to USNRC, Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed Technical Specifications Change Request for Reactor Coolant System Pressure/Temperature Limits, LTOPS Setpoint, and LTOPS Enable Temperature With Exemption Request for Alternate Material Properties Basis Per 10 CFR 50.60(b), dated December 17, 2004.
2.
Topical Report VEP-NAF-3A, Reactor Vessel Fluence Analysis Methodology, November 1997.
3.
Letter from L.N. Hartz, Virginia Electric and Power Company to USNRC, Virginia Electric and Power Company (Dominion) North Anna Power Stations Units 1 and 2 Request for Additional Information Proposed Technical Specifications Change Request Reactor Coolant System Pressure/Temperature Limits LTOPS Setpoint and LTOPS Enable Temperature, dated October 28, 2004.
4.
Letter D.A. Christian, Virginia Electric and Power Company to USNRC, Virginia Electric and Power Company, Surry and North Anna Power Stations Units 1 and 2, License Renewal Application, dated May 29, 2001.
5.
Letter from L.N. Hartz, Virginia Electric and Power Company to USNRC, Virginia Electric and Power Company (Dominion), Surry and North Anna Power Stations Units 1 and 2, Response to Request for Supplemental Information License Renewal Applications, dated October 15, 2002.
6.
Letter from USNRC to D.A. Christian, License Renewal Safety Evaluation Report for North Anna, Units 1 and 2, and Surry, Units 1 and 2," dated November 5, 2002.
7.
Letter from USNRC to D.A. Christian, North Anna Power Station, Units 1 and 2 -
Issuance of Amendments on Reactor Coolant System Pressure and Temperature Limits (TAC Nos. MC3705 and MC3706), dated July 8, 2005.
Principal Contributors: L. Lois N. Ray Date: January 3, 2006
Surry Power Station, Units 1 & 2 cc:
Ms. Lillian M. Cuoco, Esq.
Senior Counsel Dominion Resources Services, Inc.
Building 475, 5th Floor Rope Ferry Road Waterford, Connecticut 06385 Mr. Donald E. Jernigan Site Vice President Surry Power Station Virginia Electric and Power Company 5570 Hog Island Road Surry, Virginia 23883-0315 Senior Resident Inspector Surry Power Station U. S. Nuclear Regulatory Commission 5850 Hog Island Road Surry, Virginia 23883 Chairman Board of Supervisors of Surry County Surry County Courthouse Surry, Virginia 23683 Dr. W. T. Lough Virginia State Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23218 Dr. Robert B. Stroube, MD, MPH State Health Commissioner Office of the Commissioner Virginia Department of Health Post Office Box 2448 Richmond, Virginia 23218 Office of the Attorney General Commonwealth of Virginia 900 East Main Street Richmond, Virginia 23219 Mr. Chris L. Funderburk, Director Nuclear Licensing & Operations Support Dominion Resources Services, Inc.
Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, Virginia 23060-6711