ML20101B375
| ML20101B375 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/06/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20101B377 | List: |
| References | |
| NUDOCS 9603140238 | |
| Download: ML20101B375 (16) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20066-0001
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VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 199 License No. NPF-4 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated September 19, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
9603140238 960306 PDR ADOCK 05000338 p
1 i
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.D.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:
4 (2) Technical Snecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 199, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
i 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMISSION
.b awc] 0. Y ser Eugene V. Imbro, Director Project Directorate II-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 5,1996 i
1 ATTACHMENT TO LICENSE AMENDMENT NO. 199 TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 Replace the following pages of the Appendix "A" Technical Specifications with 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 3/4 7-1 3/4 7-1 3/4 7-2 3/4 7-2 3/4 7-3 3/4 7-3 B 3/4 7-1 B 3/4 7-1 B 3/4 7-2 B 3/4 7-2
)
3/_41 PLANT SYSTEMS 4
3/4.7,1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 4
3.7.1.1 All main steam line code safety valves associated with each steam generator of an unisolated reactor coolant loop shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY:
MODES 1,2 and 3.
l ACTION:
\\
With one or more main steam line code safety valves inoperable, operation in l
a.
MODES 1,2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
/
NORTH ANNA - UNIT 1 3/4 7-1 Amendment No. 199 I
3 TABLE 3.7-1 MAXIMUM AT LOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE S AFETY VALVES l
Maximum Number ofInoperable Safety Maximum Allowable Power Range eu r n X
Poi Valves on Any Ooerating Steam Generator gp gg p
1 52 2
37 1
3 21 l
l i
l l
NORTH ANNA - UNIT 1 3/47-2 Amendment No. 199
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4 i
l 1
l l
This Page Deleted i
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l NORTH ANNA - UNIT 1 3/47-3 Amendment No. 199 l
l
S 314 1 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensure that the secondary j
system pressure will be limited to within 110% of the system design pressure, during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition. The total 6
relieving capacity for all safety valves on all of the steam lines is 12.83 x 10 lbs/hr which is greater 6
than the total secondary steam flow of 12.77 x 10 lbs/hr at 100% RATED THERMAL POWER.
A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived from the following conservative calculation such that the maximum power level allowed for operation with inoperable MSSVs is below the heat removing capability of the operable MSSVs.
In order to calculate these setpoints, the governing equation is the relationship: q = m Ah, where q is the heat input from the primary side, m is the steam flow rate, and Ah is the heat of the vaporization at the steam relief pressure Therefore, the equation used in defining the revised setpoint values is:
(*
b, N)
Hi 4 = 100 g
xs K
NORTH ANNA - UNIT 1 B 3/4 7-1 Amendment No. 84, 199
L
)
e PLANT SYSTEMS BASES I
Where:
l Hi @ = Safety analysis power range high neutron flux setpoint, percent Q
= Nominal NSSS power rating of the plant (including reactor coolant pump heat, MWt K
= Conversion factor,947.82 (Btu / sec) / MWt g
= Minimum total steam flow rate capability of the operable MSSVs on any w
one steam generator at the highest MSSV operating pressum including tolerance and accumulation, as appropriate, Ibm / see hrg
= Heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu / lbm N
' = Number ofloops in plant The resulting values calculated from this equation are reduced by 9% power to account for l
instrument and channel uncertainties. With the revised values, the maximum plant operating power level would be lower than the reactor protection system setpoint by an appropriate operating margin.
3/4.7.1.2 AUXII IARY FFFDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss of off-site power.
l The original design basis of the AFW system provided for two motor driven AFW pumps l
(AFWP) each capable of delivering 340 gpm and a single turbine driven AFWP capable of delivering 700 gpm to the steam generators during accident conditions. The design basis accidents for the AFW system are the loss of normal feedwater (LONF), the loss of offsite power (LOOP),
which are ANS Condition II events, and the main feedline break (MFLB), which is an ANS Condition IV event.
,4..,
Current analyses of the design basis accidents for the AFW system have shown that the applicable accident analysis acceptance criteria are met, including the effects of a single active j
failure of any AFWP to start, if each AFWP is capable of delivering 2 300 gpm' to its respective l
steam generator at the safety valve set pressure (including the effects of setpoint drift).
l l
NORTH ANNA - UNIT 1 B 3/4 7-2 Amendment No. -174,199 l
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UNITED STATES y
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NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D.C. 2066&4001
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VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 180 License No. NPF-7 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
i The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated September 19, 1995, 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 (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.
O 5
I 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.C.(2) of Facility Operating License No. NPF-7 is hereby amended to read as follows:
(2)
Technical Snecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 180, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
i FOR THE NUCLEAR REGULATORY COMMISSION b W0 0 hr Eugene V. Imbro, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 6, 1996 l
9 5
ATTACHMENT TO LICENSE AMENDMENT NO. 180 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the following pages of the Appendix "A" Technical Specifications with 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 3/4 7-1 3/4 7-1 3/4 7-2 3/4 7-2 3/4 7-3 3/4 7-3 8 3/4 7-1 B 3/4 7-1 B 3/4 7-2 B 3/4 7-2 i
4
s 3/4J PLANT SYSTEMS 4
3/4.7.1 TURBINE CYCLE SAEETY VALVES.
LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator of an unisolated reactor coolant loop shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY:
MODES 1,2 and 3.
ACTION:
With one or more main steam line code safety valves inoperable, operation in l
a.
MODES 1,2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
NORTH ANNA - UNIT 2 3/47-1 Amendment No.180 4-
S TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE S AFETY VALVES l
Maximum Number ofInoperable Safety Maximum Allowable Power Range Valves on Any Ooerating Steam Generator (Percen of RA WER) 1 52 2
37 3
21 NORTH ANNA - UNIT 2 3/47-2 Amendment No.180
o t
This Page Deleted NORTH ANNA - UNIT 2 3/473 Amendment No.180
1 3/11 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES l
The OPERABILITY of the main steam line code safety valves ensure that the secondary system pressure will be limited to within 110% of the system design pressure, during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser)
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition. The total 6
relicving capacity for all safety valves on all of the steam lines is 12.83 x 10 lbs/hr which is greater 6
than the total secondary steam flow of 12.77 x 10 lbs/hr at 100% RATED THERMAL POWER.
A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived from the following conservative calculation such that the maximum power level allowed for operation with inoperable MSSVs is below the heat removing capability of the operable MSSVs.
In order to calculate these setpoints, the governing equation is the relationship: q = m Ah, where q is the heat input from the primary side, m is the steam flow rate, and Ah is the heat of the vaporization at the steam relief pressure. Therefore, the equation used in defining the revised setpoint values is:
Hi 4 = 100
(*s h, N) f x
NORTH ANNA - UNIT 2 B 3/4 71 Amendment No. -74, 180
d PLANT SYSTEMS e
BASES Where:
Hi @ = Safety analysis power range high neutron flux setpoint, percent Q
= Nominal NSSS power rating of the plant (including reactor coolant pump heat, MWt K
= Conversion factor,947.82 (Btu / sec) / MWt s
= Minimum total steam flow rate capability of the operable MSSVs on any w
one steam generator at the highest MSSV operating pressum including tolerance and accumulation, as appropriate, Ibm / see hrg
= Heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu / lbm N'
= Number ofloops in plant The resulting values calculated from this equation are reduced by 9% power to account for instrument and channel uncertainties. With the revised values, the maximum plant operating j
power level would be lower than the reactor protection system setpoint by an appropriate operating margin.
3/4.7.1.2 AUXII IARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensums that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a totalloss of off-site power.
The original design basis of the AFW system provided for two motor driven AFW pumps (AFWP) each capable of delivering 340 gpm and a single turbine driven AFWP capable of delivering 700 gpm to the steam generators during accident conditions. The design basis accidents for the AFW system are the loss of normal feedwater (LONF), the loss of offsite power (LOOP),
which are ANS Condition II events, and the main feedline break (MFLB), which is an ANS Condition IV event.
Current analyses of the design basis accidents for the AFW system have shown that the applicable accident analysis acceptance criteria are met, including the effects of a single active failure of any AFWP to stan, if each AFWP is capable of delivering 2 300 gpm to its respective steam generator at the safety valve set pressure (including the effects of setpoint drift).
NORTH ANNA - UNIT 2 B 3/4 7-2 Amendment No. 4-M,180