ML100670110
| ML100670110 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/09/2010 |
| From: | V Sreenivas Plant Licensing Branch II |
| To: | Heacock D Virginia Electric & Power Co (VEPCO) |
| Sreenivas V.NRR/DORL/LPL2-1 415-2597 | |
| References | |
| TAC ME0264, TAC ME0265 | |
| Download: ML100670110 (15) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 9, 2010 Mr. David A. Heacock President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
CORRECTION TO AMENDMENT NOS. 258 AND 259 FOR NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS TO ADOPT TSTF-490, REVISION 0, REGARDING DELETION OF E BAR DEFINITION AND REVISION TO REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NOS. ME0264 AND ME0265)
Dear Mr. Heacock:
On March 3, 2010, the U.S. Nuclear Regulatory Commission (NRC) issued Amendment No. 258 to Renewed Facility Operating License No. NPF-4 and Amendment No. 259 to Renewed Facility Operating License No. NPF-7 for the North Anna Power Station, Unit Nos. 1 and 2, respectively.
These amendments were in response to your application dated December 17, 2008, as supplemented by letters dated January 26, May 26, and November 23,2009.
These amendments revised the license and Technical Specifications (TSs) to rel1ect changes in the adoption of Technical Specification Task Force (TSTF)-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS [reactor coolant system] Specific Activity Technical Specification,"
for pressurized water reactor Standard Technical Specifications (STS). The NRC staff has corrected the Unit 2 Amendment to reflect the correct Amendment No. which should be 239.
D. Heacock
- 2 Enclosed are the corrected amendment page, license page, errata sheet, TS pages 1.1-2, 1.1-3, 1.1-4, 1.1-5, 1.1-6, 3.4.16-1, 3.4.16-2, and page 1 of the Safety Evaluation (SE). The NRC regrets any inconvenience this may have caused. The revisions are identified by lines in the margins.
Sincerely,
~&(t::J Dr. V. Sreenivas, Project f:::ager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339
Enclosures:
- 1. Amendment No. 239 to NPF-7
- 2. License page
- 3. Errata Sheet
- 4. TS pages
- 5. Page 1 of SE cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 239 Renewed License No. NPF-7
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company et aI., (the licensee) dated December 17, 2008, as supplemented by letters dated January 26, May 26, and November 23, 2009, 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.
Corrected by letter dated March 9, 2010
- 2
- 2.
Accordingly, the license is amended by changes to paragraph 2.C.(1) of Renewed Facility Operating License No. NPF-7, as indicated in the attachment to this license amendment, and is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 239, 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 of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. NPF-7 and the Technical Specifications Date of Issuance: March 3, 2010 Corrected by letter dated MaTch q, ZnlO
ATTACHIVlEI\\JT TO LICENSE AMENDMENT NO. 258 RENEWED FACILITY OPERATING LICENSE NO. I\\IPF-4 DOCKET NO. 50-338 AND TO LICENSE AMENDMENT NO. 239 RENEWED FACILITY OPERATING LICENSE NO. I\\IPF-7 DOCKET NO. 50-339 Replace the following pages of the Licenses and the Appendix "A" Technical Specifications (TSs) with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Pages Insert Pages Licenses Licenses License No. NPF-4, page 3 License No. NPF-4, page 3 License No. NPF-7, page 3 License No. NPF-7, page 3 1.1-2 1.1-2 1.1-3 1.1-3 1.1-4 1.1-4 1.1-5 1.1-5 1.1-6 1.1-6 3.4.16-1 3.4.16-1 3.4.16-2 3.4.16-2 3.4.16-3 Corrected by letter dated lVIarch 9, 2010
-3 (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCa to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30,40, and 70, VEPCa to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material, without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30,40, and 70, VEPCa to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations as set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level VEPca is authorized to operate the facility at steady state reactor core power levels not in excess of 2940 megawatts (thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment N0239,are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the condition or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission:
- a. If VEPCa plans to remove or to make significant changes in the normal operation of equipment that controls the amount of radioactivity in effluents from the North Anna Power Station, the Renewed License No. NPF-7 NORTH ANNA - UNIT 2 flrrendrent r-b. 239 Corrected by letter dated March 9, 2010
1.1 Definitions 1.1 Definitions CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)
CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)
DOSE EQUIVALENT 1-131 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.
CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using the thyroid dose conversion factors listed in Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.
North Anna Units 1 and 2 1.1-2 Amendments 258 and 239 Corrected by letter dated March q. 2010
1.1 Definitions 1.1 Definitions DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12,1993, "External Exposure to Radionucl ides in Air, Water, and Soil."
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
LEAKAGE shall be:
- a. Identifi ed LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (continued)
North Anna Units 1 and 2 1.1-3 Amendments 258 and 239 Corrected by letter dated ~1arch Q. 2010
Definitions 1.1 1.1 Definitions LEAKAGE (continued)
MASTER RELAY TEST MODE OPERABLE-OPERABILITY
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
- 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
- b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
A MODE shall correspond to anyone inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, (continued)
North Anna Units 1 and 2 1.1-4 Amendments 258 and 239 Corrected by letter dated~~rch Q,
~nln
Definitions 1.1 1.1 Definitions OPERABLE-OPERABILITY (continued)
PHYSICS TESTS QUADRANT POWER TILT RATIO (QPTR)
RATED THERMAL POWER (RTP)
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME SHUTDOWN MARGIN (SDM) component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
- a. Described in Chapter 14, Initial Tests and Operation, of the UFSAR;
- b. Authorized under the provisions of 10 CFR 50.59; or
- c. Otherwise approved by the Nuclear Regulatory Commission.
QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2940 MWt.
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS tri p setpoi nt at the channel sensor unti 1 loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequenti al, overl appi ng, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
(continued)
North Anna Units 1 and 2 1.1-5 Amendments 258 and 239 I Corrected by letter dated t~r~h 0, ?nln
Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM)
(continued)
SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)
- a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA Bf highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequent i a1, overl app"i ng, or total steps.
A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy.
The TADOT may be performed by means of any series of sequent i ai, overl appi ng, or total channel steps.
North Anna Units 1 and 2 1.1-6 Amendments 258 and 239 Corrected by letter dated MRrch q, 7.nln
Res Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.
APPLICABILITY:
MODES 1, 2, 3, and 4 ACTIONS A.
CONDITION DOSE EQUIVALENT 1-131 not within limit.
REQUI RED ACTION
NOTE-----------
LCO 3.0.4.c is applicable.
I COMPLETION TIME B.
DOSE EQUIVALENT XE-133 not within limit.
A.l Verify DOSE EQUIVALENT 1-131 ~ 60 ~Ci/gm.
AND A.2 Restore DOSE EQUIVALENT 1-131 to within limit.
NOTE-----------
LCO 3.0.4.c is applicable.
Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 48 hours C.
Required Action and associated Completion Time of Condition A or B not met.
OR DOSE EQUIVALENT 1-131
> 60 flCi/9m.
B.1 C.1 AND C.2 Restore DOSE EQUIVALENT XE-133 within limit.
Be in MODE 3.
Be in MODE 5.
to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 16 hours 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> North Anna Units 1 and 2 3.4.16-1 Amendments 258 and 239 Corrected by 1etter dateo
~'<1rclt O. ?,(111)
RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant DOSE EQUIVALENT XE 133 specific activity ~ 197 ~Ci/gm.
7 days SR 3.4.16.2 Verify reactor coolant DOSE EQUIVALENT I-I31 specific activity ~ 1.0 ~Ci/gm.
14 days AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of 2 15% RTP withi n a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> peri od North Anna Units 1 and 2 3.4.16-2 Amendments 258 and 239 Corrected by letter dated March Q, 7010
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 258 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-4 AND AMENDMENT NO. 239 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION. UNIT NOS. 1 AND 2 DOCKET NOS. 50-338 AND 50-339
1.0 INTRODUCTION
By letter dated December 17, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML083530982). Virginia Electric and Power Company (the licensee) submitted a request for changes to the North Anna Power Station, Unit Nos. 1 and 2 (NAPS 1 and 2), Technical Specifications (TSs). The requested changes are the adoption of Technical Specification Task Force (TSTF)-490, Revision 0, "Deletion of E-Bar Definition and Revision to Reactor Coolant System (RCS) Specific Activity Technical Specification," to the pressurized water reactor (PWR) Standard Technical Specifications (STSs). Subsequently, the licensee's original license amendment request (LAR) was supplemented by letters dated January 26 (ADAMS Accession No. ML090260528), May 26 (ADAMS Accession No. ML091460589), and November 23, 2009 (ADAMS Accession No. ML093280238). The supplements dated January 26, May 26, and November 23, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the original proposed no significant hazards consideration determination.
By letter dated September 13, 2005 (ADAMS Accession No. ML052630462), the TSTF submitted TSTF-490 for Nuclear Regulatory Commission (NRC) staff review. The TSTF with proposed changes for incorporation into the STS as TSTF-490, Revision 0, was referenced in the Federal Register Notice (FRN) (71 FR 67170) of November 20,2006, for public comments. By FRN (72 FR 12217) dated March 15,2007 (ADAMS Accession No. ML070250176), the NRC published a "Notice of Availability Model Application Concerning Technical Specification Improvement Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process." This TSTF involves changes to NUREG-1430, NUREG-1431, and NUREG-1432 STS Section 3.4.16.
Corrected by letter dated March 9, 2010
D. Heacock
- 2 Enclosed are the corrected amendment page, license page, errata sheet, TS pages 1.1-2, 1.1-3, 1.1-4, 1.1-5, 1.1-6,3.4.16-1,3.4.16-2, and page 1 of the Safety Evaluation (SE). The NRC regrets any inconvenience this may have caused. The revisions are identified by lines in the margins.
Sincerely, IRA!
Dr. V. Sreenivas, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339
Enclosures:
- 1. Amendment No. 239 to NPF-7
- 2. License page
- 3. Errata Sheet
- 4. TS pages
- 5. Page 1 of SE cc w/encls: Distribution via Listserv DISTRIBUTION:
Public RidsAcrsAcnw_MailCTR Resource LPL2-1 RlF RidsNrrDorlLpl2-1 Resource RidsOgcRpResource RidsNrrDirsltsb Resource RidsNrrPMNorthAnna (hard copy)
RidsRgn2MailCenter RidsNrrLAMO'Brien (hard copy)
RidsNrrDorlDprResource LBenton NRRlDRA/AABD Amendment No'.. ML100670110 OFFICE LPL2-1/PM LPL2-1/LA LPL2-1/BC NAME VSreenivas (KCotton for)
MO'Brien GKulesa DATE 03/09/10 03/08/10 03/09/10 OFFICIAL RECORD COpy