ML20069E181
| ML20069E181 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/26/1994 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20069E183 | List: |
| References | |
| NUDOCS 9406070063 | |
| Download: ML20069E181 (11) | |
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VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. SS-US NORTH ANNA POWER STATION. UNIT NO. 1 MiENDMENT TO FACILITY OPERATING LICENSE Amendment No.183 License No. NPF-4 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated October 4,1993, 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.
t 9406070063 940526" PDR ADOCK 05000338 P-PDR.
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2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.0.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.183, are hereby incorporated in the l
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 l
be implemented within 30 days.
(
FOR THE NUCLEAR REGULATORY COMMISSION rbert N. Be kow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 26, 1994
ATTACHMENT TO LICENSE AMENDMENT NO. 183 TO FACILITY OPERATING LICENSE N0. NPF-4 1
DOCKET NO. 50-338 4
Replace the following pages of the Appendix "A" Technical Specifications with i
the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Paaes Insert Paaes 5-4 5-4 6-17a 6-17a l
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DESGN FEATURES DESIGN PRESSURE ANDTEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 45 psig and a temperature of 280'F.
53 REACTORCORE
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FUEL ASSEMBUES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4 or ZlRLO. Each fuel rod shall have a nominal active fuel l
length of 144 inches and contain a maximum total weight of 1780 grams uranium. The initial core loading shall have a maximum enrichment of 3.2 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.3 weight' percent U 235.
i i
i CONTROL ROD ASSEAEUES l
5.3.2 The reactor core shall contain 48 full length control rod assemblies. The full length l
control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal 1
values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRNURE AND TEMPERATURE l
5.4.1 The reactor coolant system is designed and shall be maintained:
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i ADMINISTRATIVE CONTROLS (Cont'd) j 2a. WCAP 9220-P A. Rev.1," WESTINGHOUSE ECCS EVALUATION MODEL -
1981 VERSION", February 1982 (E Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor).
j l
2b. WCAP 9561 P A, ADD. 3, Rev.1, *BART A 1: A COMPUTER CODE FOR THE BEST 2 STIMATE ANALYSIS OF REFLOOD TRANSIENTS - SPECIAL i
REPORT: THIMBLE MODElJNG IN E ECCS EVALUATION MODEL", JULY, l
1986, (E. Proprietary).
l (Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor).
i 2c. WCAP-10266 P A, Rev. 2. "The 1981 Version of the Westinghouse CCCS Evaluation Model Using the BASH Code", March 1987 (E Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channe! Factor).
j-2d. WCAP-10054 P A, "Wastinghouse Small Break ECCS Evaluation Model 1
Using the NOTRUMP Code," August 1985 (E Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor).
i 2e. WCAP 10079 P A, "NOTRUMP, A Nodal Transient Small Break and General Network Code', August 1985 (E Proprietary).
(Me'nodology for LCO 3.2.2 - H6at Flux Hot Channel Factor).
1 2f. WC AP-12610, " VANTAGE + FUEL ASSEMBLY REPORT," June 1990 (E Froprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor.)
I Sa. VEP NE 2 A, ' Statistical DNBR Evaluation Methodology", June 1987.
(Methodology for LCO 3.2.3, Nuclear Enthalpy Rise Hot Channel Factor),
i 3b. VEP-NE-3-A, "Oualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code", July 1990.
4 (Methodology for LCO 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor).
4.
VEP NE-1 A, "Vopeo Relaxed Power Distribution Cor#ol Methodclogy and Associated FO Surveillance Technical Specifications," March 1986.
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor and LCO 3.2.1 -
. Axial Hux Difference.)
NORTH ANNA-UNIT 1 6-17a AmendmentNo. U6,183
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WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION. UNIT N0. 2 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 164 License No. NPF-7 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
l A.
The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated October 4,1993, complies with the l
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 (1) 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.
4 i-
' 4 2.
Accordingly, the license is amended by changes to the Technical Speci-I fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of facility Operating License No. NPF-7 is hereby j
amended to read as follows:
(2) Technical Soecifications i
The Technical Specifications contained in Appendices A and B, as i
revised through Amendment No.164
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
j 3.
This license amendment is effective as of its date of issuance and shall 4
be implemented within 30 days.
j F0 T E NUCLEAR REGULATORY COM>ilSSION i
Herbert N. Berkow, Director Project Directorate Il-2 Division of Reactor Pt/ojects - I/II j
Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications Date of Issuance: May 26, 1994 3
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l ATTACHMENT TO LICENSE AMENDMENT NO.164 4
TO FACILITY OPERATING LICENSE NO. NPF,
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DOCKET NO. 50-339 3
- Replace the following pages of'the Appendix "A" Technical Specifications with i
the enclosed r'.aes as indicated. The revised pages are identified by i
amendment nunb e and contain vertical lines indicating the area of change.
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The corresponding overleaf page is also provided to maintain document
- completeness.
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I lecimt McIA - UltIT 2 5-3 Amendment No. 3J,53
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i DESGN FEATURES 4
l 53 REACTORCORE FUELASSEMBUES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel cssembly containing 264 fuel rods clad with Zircaloy-4 or ZlRLO. Each fuel rod shall have a nominal active fuel l
length of 144 inches and contain a maximum total weight of 1780 grams uranium. The initial core loading shall have a maximum enrichment of 3.2 weight percent U 235. Reload fuel shall l
be similar in physical design to the initial core loading and shall have a maximum enrichment of j
4.3 weight percent U 235.
4 CONTROL ROD ANMRf IFA i
5.3.2 The reactor core shall contain 48 full length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal a
values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
5.4 REACTORCOOLANTSYSTEM i
DESIGN PRERSURE AND TEMPERATURE i
l 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the apptbable Surveillance
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Requirements,
- b. For a pressure of 2485 psig, and
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- c. For a temperature of 650'F, except for the pressurizer wl'ich is 680*F.
l 1
j VOLUME
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5.4.2 The total water and steam volume of the reactor coolant system is approximately 10,000 cubic feet at nominal operating conditions.
i 3
t NORTH ANNA-UNIT 2 5-4 Amendment No. E, 75,777,7A6,164 4
i
ADMINISTRATIVE COqROLS (Cont'd) 2a. WCAP 9220.P-A. Rev.1, ' WESTINGHOUSE FCCS EVALUATION MODEL -
1981 VERSION", February 1982 (E Proprietary).
(Methodology for LCO 3.2.2 Heat Flux Hot Channel Factor).
2b. WCAP 9561 P A, ADO. 3 Rev.1, 'BART A 1: A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS - SPECIAL REPORT: THIMBLE MODELING IN E ECCS EVALUATION MODEL", JULY, 1986, (W. Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor).
2c. WCAP 10266 P A, Rev. 2, "The 1981 Version of the Westinghouse ECCS '
Evaluation Model Using the BASH Code", March 1987 (E Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor).
2d. WCAP 10054 P A,." Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 (E Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor).
2e. WCAP 10079 P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code", August 1985 (E Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor).
2f. WCAP-12610, ' VANTAGE + FUEL ASSEMBLY REPORT," June 1990 (E Proprietary).
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor.)
3a. VEP NE 2 A, " Statistical DNBR Evaluation Methodology *, June 1987.
(Methodology for LCO 3.2.3, Nuclear Enthalpy Rise Hot Channel Factor).
3b. VEP NE 3-A, "Oualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code", July 1990.
(Methodology for LCO 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor).
4.
VEP NE 1-A, "Vepco Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications," March 1986.
(Methodology for LCO 3.2.2 - Heat Flux Hot Channel Factor and LCO 3.2.1 -
Axial Flux Difference.)
NORTH ANNA-UNIT 2 6 17a Amendment No. 739,164
- ..