ML061500044
| ML061500044 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/31/2006 |
| From: | Stephen Monarque Plant Licensing Branch III-2 |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| Monarque, S R, NRR/DORL, 415-1544 | |
| Shared Package | |
| ML061500048 | List: |
| References | |
| TAC MC7971, TAC MC7972 | |
| Download: ML061500044 (16) | |
Text
May 31, 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, VA 23060-6701
SUBJECT:
SURRY POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS ON REVISION OF ACCIDENT MONITORING INSTRUMENTATION LISTING, ALLOWED OUTAGE TIMES, REQUIREMENTS, AND SURVEILLANCES (TAC NOS. MC7971 AND MC7972)
Dear Mr. Christian:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 247 to Renewed Facility Operating License No. DPR-32 and Amendment No. 246 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 (TSs) in response to your application dated July 21, 2005..
These amendments revise the accident monitoring instrumentation listing, allowed outage times, requirements, and surveillances to be consistent with the requirements of the Improved Technical Specifications for post-accident monitoring instrumentation. In addition, editorial changes were made in the Bases for TSs 3.7 and 4.1, pages 3.7-7 and 4.1-3.
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. to DPR-32
- 2. Amendment No. to DPR-37
- 3. Safety Evaluation cc w/encls: See next page
May 31, 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, VA 23060-6701
SUBJECT:
SURRY POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS ON REVISION OF ACCIDENT MONITORING INSTRUMENTATION LISTING, ALLOWED OUTAGE TIMES, REQUIREMENTS, AND SURVEILLANCES (TAC NOS. MC7971 AND MC7972)
Dear Mr. Christian:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 247 to Renewed Facility Operating License No. DPR-32 and Amendment No. 246 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 (TSs) in response to your application dated July 21, 2005..
These amendments revise the accident monitoring instrumentation listing, allowed outage times, requirements, and surveillances to be consistent with the requirements of the Improved Technical Specifications for post-accident monitoring instrumentation. In addition, editorial changes were made in the Bases for TSs 3.7 and 4.1, pages 3.7-7 and 4.1-3.
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. 247 to DPR-32
- 2. Amendment No. 246 to DPR-37
- 3. Safety Evaluation cc w/encls: See next page DISTRIBUTION: Public LPL2-1 R/F RidsAcrsAcnwMailCenter BSingal, DORL DPR RidsNrrDorlLplc(EMarinos)
GHill, OIS(4 hard copies) RidsRgn2MailCenter(KLandis)
RidsNrrPMSMonarque(hard copy)
RidsNrrDirsItsb(TKobetz) RidsOgcRp RidsNrrLAMOBrien(hard copy)
RidsNrrdeeicb(BMarcus)
Package No: ML061500048 Amendment No: ML061500044 Tech Spec No: ML061630453 OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/EICB/BC NRR/ITSB/BC OGC NRR/LPL2-1/BC NAME SMonarque MOBrien AHowe TKobetz H
EMarinos DATE 5/2/06 5/11/06 04/27/2006 04/20/2006 5/22/06 5/26/06 OFFICIAL RECORD COPY
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 247 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 July 21, 2005, 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. 247, 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 60 days.
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: May 31, 2006
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 246 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 July 21, 2005, 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. 246, 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 60 days.
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: May 31, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 LICENSE AMENDMENT NO. 246 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 3.7-2 3.7-2 3.7-7 3.7-7 3.7-8 3.7-8 3.7-9 3.7-9 3.7-29 3.7-29 4.1-3 4.1-3 4.1-9a 4.1-9a
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 246 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 application dated July 21, 2005, Virginia Electric and Power Company (the licensee) requested changes to the Technical Specifications (TSs) for the Surry Power Station, Unit Nos.
1 and 2. The proposed changes would revise the accident monitoring instrumentation listing, allowed outage times (AOTs), requirements, and surveillances to be consistent with the requirements of the Improved Technical Specifications for post-accident monitoring instrumentation. TS 3.7.E, TS Table 3.7-6, and TS Table 4.1-2 are affected by this change.
Likewise, the TS 3.7 Basis is being revised to address the proposed revision of TS 3.7.E and TS Table 3.7-6. In addition, editorial changes were made in the Bases for TSs 3.7 and 4.1, pages 3.7-7 and 4.1-3.
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commissions regulatory requirements related to the content of the TSs are contained in Title 10 to the Code of Federal Regulations (10 CFR), Part 50, Section 50.36. Section 50.36(c)(2)(ii)(C) of 10 CFR requires that TS limiting conditions for operation (LCOs) of a nuclear reactor be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Section 50.36(c)(2)(ii)(D) of 10 CFR requires that TS LCOs of a nuclear reactor be established for a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
NUREG-1431, Revision 2, contains the Improved Standard TSs (ISTSs) for Westinghouse plants and documents the positions of the U.S. Nuclear Regulatory Commission (NRC) based on the Westinghouse Owners Groups proposed standard TSs. The ISTSs are used as the basis for developing improved plant-specific TSs and support the review of requests made in accordance with this regulatory guidance.
Regulatory Guide (RG) 1.97, Revision 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, May 1983.
3.0 TECHNICAL EVALUATION
3.1 Proposed Change to TS 3.7.E TS 3.7.E currently states:
E.
The accident monitoring instrumentation listed in Table 3.7-6 shall be OPERABLE in accordance with the following:
- 1. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.7-6, items 1 through 9, either restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2 With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum OPERABLE channels requirements of Table 3.7-6, items 1 through 9, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The licensee proposes to replace TS 3.7.E with the following information.
E. Prior to the Reactor Coolant System temperature and pressure exceeding 350 EF and 450 psig, respectively, the accident monitoring instrumentation listed in Table 3.7-6 shall be OPERABLE in accordance with the following:
- 1. With one required channel inoperable either restore the inoperable channel to OPERABLE status within 30 days or submit a report to the NRC within the next 14 days. The report shall outline the cause of inoperability and the plans and schedule for restoring the inoperable channel to OPERABLE status.
- 2. With two required channels inoperable either:
- a. Restore an inoperable channel(s) to OPERABLE status within 7 days or initiate the preplanned alternate method of monitoring the appropriate function and submit a report to the NRC within the next 14 days. The report shall outline the preplanned alternate method of monitoring the function, the cause of inoperability, and the plans and schedule for restoring an inoperable channel to OPERABLE status.
- b. If no preplanned alternate method of monitoring the function is available, restore an inoperable channel(s) to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be less than 350 EF and 450 psig within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The licensees proposed 30-day AOT in TS 3.7.E.1. applies when one (or more) function(s) in TS Table 3.7-6 has one required channel that is inoperable. The proposed 30-day AOT to restore one inoperable required channel to operable status takes into account operating experience, the remaining operable channels, and the passive nature of the instrument (i.e., no automatic action is assumed to occur from this instrumentation). The licensees proposed requirement to submit a report within 14 days of the expiration of the 30-day restoration AOT incorporates the fact that the probability of an event requiring the accident monitoring instrumentation during this interval is low, and alternate means of obtaining the required information are available. As such, the NRC staff finds this proposed TS change consistent with NUREG-1431, and, therefore, acceptable.
Regarding TS 3.7.E.2., the licensees proposed 7-day AOT applies when one (or more) function(s) in TS Table 3.7-6 has two required channels that are inoperable. In its submittal dated July 21, 2005, the licensee stated that the long-term operation with two required channels inoperable in a function and with an alternate indication is not acceptable because the alternate indication may not fully meet the performance qualification requirements applied to the accident monitoring instrumentation. The licensee further stated that requiring restoration of one of the two inoperable channels limits the risk that the accident monitoring instrumentation function could be in a degraded condition should an accident occur. If there is no preplanned alternate, the 7-day AOT is followed by a requirement to be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be less than 350 °F and 450 psig within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the 7-day AOT to restore an inoperable channel to operable status is exceeded and either a redundant channel or a preplanned alternate method of monitoring is operable, a report to the NRC within the next 14 days is required. The 7-day AOT to restore one of the two inoperable required channels to operable status is appropriate based on providing a reasonable time for the repair and the probability of an event requiring accident monitoring instrument operation is low. As such, the NRC staff finds this proposed TS change consistent with NUREG-1431, and, therefore, acceptable.
3.2 Proposed Change to TS Basis page 3.7-7.
The TS Basis page 3.7-7. on accident monitoring information currently states:
The operability of the accident monitoring instrumentation in Table 3.7-6 ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. On the pressurizer PORVs, the pertinent channels consist of redundant limit switch indication. The pressurizer safety valves utilize an acoustic monitor channel and a downstream high temperature indication channel. This capability is consistent with the recommendations of Regulatory Guide 1.97, Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident, December 1975, and NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations. Potential gaseous effluent release paths are equipped with radiation monitors to detect and measure concentrations of noble gas fission products in plant gaseous effluents during and following an accident. The gaseous effluent release paths monitored are the process vent stack, ventilation vent stack, main steam safety valve and atmospheric dump valve discharge and the AFW pump turbine exhaust. The potential liquid effluent release paths via the service water discharge from the recirculation spray heat exchangers are equipped with radiation monitors to detect leakage of recirculated containment sump fluid. These radiation monitors and the associated sample pumps are required to operate during the recirculation heat removal phase following a loss of coolant accident in order to detect a passive failure of a recirculation spray heat exchanger tube. These monitors meet the requirements of NUREG-0737.
The licensee proposes to replace the paragraph above with the following information in the TS 3.7 Basis.
The primary purpose of accident monitoring instrumentation is to display unit parameters that provide information required by the control room operators during and following accident conditions. In response to NUREG-0737 and Regulatory Guide (RG) 1.97, Revision 3, a programmatic approach was developed in defining the RG 1.97-required equipment for Surry. The Surry RG 1.97 program review examined existing instrumentation with respect to the RG 1.97 design and qualification requirements. The operability of RG 1.97 instrumentation ensures that sufficient information is available on selected unit parameters to monitor and assess unit status and response during and following an accident. The availability of accident monitoring instrumentation is important so that the consequences of corrective actions can be observed and the need for and magnitude of further actions can be determined.
RG 1.97 applied a graded approach to post-accident indication by using a matrix of variable types versus variable categories. RG 1.97 delineates design and qualification criteria for the instrumentation used to measure five variable types (Types A, B, C, D, and E). These criteria are divided into three separate categories (Categories 1, 2, and 3), providing a graded approach that depended on the importance to safety of the measurement of a specific variable. Category 1 variables, listed in Table 3.7-6, are defined as follows:
Category 1 - are the key variables deemed risk significant because they are needed to:
Determine whether other systems important to safety are performing their intended functions, Provide information to the operators that will enable them to determine the likelihood of a gross breach of the barriers to radioactivity release, and Provide information regarding the release of radioactive materials to allow early indication of the need to initiate action necessary to protect the public and to estimate the magnitude of any impending threat.
The RG 1.97 criteria on redundancy requirements apply to Category 1 variables only and address single-failure criteria and supporting features, including power sources. Failures of the instrumentation, its supporting features, and/or its power source resulting in less than the required number of channels necessitate entry into the required actions.
The 30 day allowed outage time applies when one (or more) function(s) in Table 3.7-6 has one required channel that is inoperable. The 30 day allowed outage time to restore one inoperable required channel to OPERABLE status is appropriate considering the remaining channel is OPERABLE, the passive nature of the instrument (i.e., no automatic action is assumed to occur from this instrumentation), and the low probability of an event requiring accident monitoring instrumentation during this interval. The 7 day allowed outage time applies when one (or more) function(s) in Table 3.7-6 has two required channels that are inoperable. The 7 day allowed outage time to restore one of the two inoperable required channels to OPERABLE status is appropriate based on providing a reasonable time for the repair and the low probability of an event requiring accident monitoring instrument operation. Long-term with two channels inoperable in a function and with an alternate indication is not acceptable because the alternate indication may not fully meet the performance qualification requirements applied to the accident monitoring instrumentation. Requiring restoration of one of the two inoperable channels limits the risk that the accident monitoring instrumentation function could be in a degraded condition should an accident occur. If there is no preplanned alternate, the 7 day allowed outage time is followed by a requirement to be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be less than 350°F and 450 psig within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the 30 day allowed outage time or 7 day allowed outage time to restore an inoperable channel to OPERABLE status is exceeded and either a redundant channel or a preplanned alternate method of monitoring is OPERABLE, a report to the NRC within the next 14 days is required. The report to the NRC in lieu of a shutdown is appropriate because the instrument functional capability has not been lost and given the low likelihood of unit conditions that would require the information provided by the accident monitoring instrumentation.
Note that the Categories 2 and 3 RG 1.97 variables are addressed in a licensee controlled document and are defined as follows:
Category 2 - provides less stringent requirements and generally applies to instrumentation designated for indicating system operating status.
Category 3 - is the least stringent and is applied to backup and diagnostic instrumentation.
The licensee proposed these bases changes in order to incorporate the criteria specified in RG 1.97. This included the purpose of RG 1.97, a description of categories 1, 2, and 3, and a description of the proposed TS changes. The NRC staff finds that the proposed change is consistent with the guidance contained in RG 1.97. In addition, the NRC staff also finds this change to be consistent with NUREG-1431, and, therefore, acceptable.
In addition, in TS 3.7 Basis, TS page 3.7-7, the title Explosive Gas Monitoring is added above the paragraph that discusses explosive gas monitoring for clarity. The NRC staff finds this editorial change to be acceptable.
3.3 Proposed Change to TS Table 3.7-6 The licensee proposes to revise Table 3.7-6 by adding a note stating that separate entry into TS 3.7.E is allowed for each function listed in Table 3.7-6, replacing the two columns listing the total number of channels and minimum operable channels with the number of required channels, and replacing the instrument column with a function column. In addition, the licensee proposes to add instrumentation functions that are Category 1 RG 1.97 variables to Table 3.7-6 along with their required number of channels. Category 2 RG 1.97 variables are being deleted from Table 3.7-6 and will be transferred to the licensees controlled document. The licensee also proposed to delete Note 1 that provides operator actions in the event the number of operable channels is less than the number of minimum operable channels. The licensee indicated that it requested this change in order to establish consistency with NUREG-1431.
Accordingly, the NRC staff has evaluated the licensees proposed change and has determined that the licensee has followed the guidance of RG 1.97 and its proposed change is also consistent with NUREG-1431. As such, the NRC staff finds this proposed change to be acceptable.
3.4 Proposed Change to TS Bases page 4.1-3.
The licensee proposes to add a heading tilted Testing above the paragraph that discusses the operability of the reactor trip and the engineered safety features actuation systems. The NRC staff has determined that this proposed change is an editorial change, and, therefore, acceptable.
3.5 Proposed Change to TS Table 4.1-2.
The licensee has proposed to revise the TS Table 4.1-2 listing of accident monitoring instrumentation to match the functions shown on TS Table 3.7-6. The licensee proposes to add instrumentation functions that are Category 1 RG 1.97 variables to TS Table 4.1-2 and to delete Category 2 RG 1.97 variables from this table. In addition, the licensee has proposed to eliminate the Channel Function Test column shown on TS Table 4.1-2. Because the quarterly channel functional test is performed only on accident monitoring instrumentation that has been identified as Category 2 RG 1.97, and TS Table 4.1-2 will only reflect Category 1 RG 1.97 variables, the channel function test and the quarterly surveillance requirement are no longer needed. The licensee has proposed to add three notes to TS Table 4.1-2 that requires the performance of the channel check for each required instrumentation channel that is normally energized, allows for an operational test for the penetration flow path containment isolation valve position, and excludes the power and source range neutron detectors from channel calibration. The note pertaining to the exclusion of the neutron detectors from channel calibration has been copied from the same note presented in TS Table 4.1-1, Remarks 4).
TS Table 4.1-1, Remarks 4) states that neutron detectors may be excluded from the channel calibration requirements. TS Table 4.1-1, Remarks 4) is applied to the 18-month frequency for nuclear power and source range channels. The licensee proposes to copy this same note and apply it to the 18-month frequency for performing channel calibrations for the power and source range neutron detectors in TS Table 4.1-2. As such, no TS requirements have been reduced or eliminated. Accordingly, the NRC staff has evaluated the licensees proposed change and has determined that the licensee has followed the guidance of RG 1.97 and its proposed change is also consistent with NUREG-1431. As such, the NRC staff finds this proposed change to be acceptable.
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 and changes surveillance requirements. 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 (71 FR 155). Accordingly, the amendments meet 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 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.
Principal Contributor: S. Monarque Date of Issuance: May 31, 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