ML20236X374
| ML20236X374 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/03/1998 |
| From: | Kuo P NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20236X377 | List: |
| References | |
| NUDOCS 9808070335 | |
| Download: ML20236X374 (13) | |
Text
,
p+wn
[
4 UNITED STATES g
j NUCLEAR REGULATORY COMMISSION s.
4 t
WASHINGTON, D.C. 20066 4 001 p*'+,
.....,o!
i VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE l
DOCKET NO. 50-338 NORTH ANNA ROWER STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 213 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., Ghe licensse) dated November 5,1997, complies with the standarGs 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 I
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.
i
{
9808070335 980803 PDR ADOCK 05000338 P
pm
- v.,
D, 2.
Accordingly, the license is amended by changes to the Technical Specifications 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:
(2)
Technical Specifications The Technical Specifications contaired in Appendices A and B, as revised through Amendment No. 213, are hereby incorporated in the 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 30 days, j
I FOR THE NUCLEAR REGULATORY COMMISSION
/?7/
I P T. Kuo, Acting Director i
Project Directorate 11-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: August 3,1998 L
s.
ATTACHMENT TO LICENSE AMENDMENT NO. 213 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 verticallines indicating the areas of change.
Remove Paaes insert Paoes 3/4 9-7 3/4 9-7 B 3/4 9-2 B 3/4 9-2 B 3/4 9-3 8 3/4 9-3 i
1 l
l l
i
v..
REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL PIT LIMITING CONDITION FOR OPERATION l
3.9.7 Loads in excess of 2500 pounds shall be prohibited from travel over irradiated fuel assemblies in the spent fuel pit. This does not apply to movement of any spent fuel pit gate provided 1
eac'h of the following is satisfied:
- a. the top of the gate (excluding lifting lugs)is no higher than 15 inches above the top of the moveable platform crane deck support beam while over irradiated fuel,
- b. the gate is rigged to slack-free safety cables while over irradiated fuel, irradiated fuel containing Rod Control Cluster Assemblies are excluded along the load c.
path where the gate is moved, and
- d. irradiated fuel is prohibited in the cask area when the gate is lifted over the spent fuel cask handling area. There is no restriction on lift height.
APPLICABILITY:
With irradiated fuel assemblies in the spent fuel pit.
ACTION:
With the requirements of the above specification not sr.iisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
l SURVEILLANCE REQUIREMENTS 4.9.7.1 Loads other than the spent fuel pit gates shall be verified to be less than 2500 pounds l
prior to movement over irradiated fuel assemblies in the spent fuel pit.
l 4.9.7.2 For movement of any of the spent fuel pit gates:
gate lift height and slack-free redundant rigging shall be verified prior to moving over
{
a.
irradiated fuel, j
l
- b. load paths shall be verified not to have irradiated fuel with Rod Control Cluster Assemblies present in the gate load path, and i
- c. the spent fuel cask handling area shall be verified to have no irradiated fuel present prior to moving a gate over the area.
Amendment No. 8, 213 NORTH ANNA - UNIT 1 3/4 9-7
REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE OPERABILITY The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movement of control rods and fuel assemblies; 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
1/4 9.7 CRANE TRAVEL - SPENT FUEL PIT The restriction on movements of the spent fuel pit gates over irradiated fuel ensure that redundant rigging will be used in order to prevent a gate drop caused by hoist failure. As shown by calculation,in the event the load is dropped (1) the spent fuel storage racks limit gate penetration and prevent the impact load from being applied to stored fuel,(2) fuel spacing will not be changed and (3) impact loading to the spent fuel pit structure is acceptable.
The restriction on movement of other loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped,1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.
3ff.Sl RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.
After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 6 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140 F. Since the decay heat power production rate decreases with time after shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removal is provided as long as the reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140 F. The reduced flow rate provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation. During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circu!ation to minimize the effect of a boron dilution incident and to prevent boron stratification.
The requirement to have two RHR loops OPERABLE when there is less than 23 feet of i
l water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel i
head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus,in the event of a failure of th ' operating RHR loop, adequate time
{
is provided to initiate emergency procedures to cool the core
{
NORTH ANNA - UNIT 1 B 3/4 9-2 Amendment No. 12r44, 213
_fL
REFUELING OPERATIONS BASES 1/4 9 9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.
3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND SPENT FUEI PIT The restrictions on minimum water level ensure that sufficient water depth is available to remove 99c/c of the assumed 10c7c iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.
The minimum water level for movement of fuel assemblies (23 feet above the vessel flange) assures that sufficient water depth is maintained above fuel elements being moved to or from the vessel. With the upper intemals in place, fuel assemblies and control rods cannot be removed from the vessel. Operations involving the lifting of control rods with the vessel upper in:ernals in place may proceed with less than 23 feet of water above the vessel flange provided that 23 feet of water is maintained above all irradiated fuel assemblies within the reactor vessel.
3/4.9.12 FUEL BUILDING VENTILATION SYSTEM The limitations on the fuel building ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the auxiliary building HEPA and charcoal filter assemblies prior to discharge to the atmosphere. The Fuel Handling Accident analysis does not require filtration of the fuel building exhaust in order to meet the analysis criteria.
However, the OPERABILITY of this system and the resulting iodine removal capacity provide additional conservatism compared with the assumptions of the accident analyses.
l 1
l 1
t NORTH ANNA UNIT 1 B 3/4 9-3 Amendment No. 445Agg, 213
I o rar p
UNITED STATES g
j NUCLEAR REGULATORY COMMISSION e
WASHINGTON, D.C. 30006 4 001 l
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. I94 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 al., (the licensee) dated November 5,1997, 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.
s-
2.
Accordingly, the license is amended by changes to the Technical Specifications 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 Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.194, are hereby incorporated in the 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 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION G
7 T. Kuo, A ting Director roject Directorate ll-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
l I
l I
i
ATTACHMENT TO LICENSE AMENDMENT NO.194 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 verticallines indicating the areas of change. Overleaf page 3/4 9-7 is included for document completeness.
Remove Paaes insert Paaes 3/4 9-8 3/4 9-8 8 3/4 9-2 B 3/4 9-2 B 3/4 9-3 B 3/4 9-3 I
l l
l i
!~u__
(
REFUELING OPERATIONS MANIPULATOR CRANE OPERABILITY LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane and auxiliary hoist shall be used for movement of control rods or fuel assemblies and shall be OPERABLE with:
The manipulator crane used for movement of fuel assemblies a.
having:
1.
A minimum capacity of 3250 pounds, and 2.
An overload cut off limit less than or equal to 2850 pounds.
b.
The auxiliary hoist used for movement of control rods having:
1.
A minimum capacity of 700 pounds, and 2.
A load indicator which shall be used to prevent lifting loads in excess of 600 pounds.
s APPLICABILITY:
During movement of control rods or fuel assemblies within the reactor pressure vessel.
ACTION:
With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable manipulator crane and/or auxiliary hoist from operations involving the movement of control rods and fuel assemblies within the reactor pressure vessel.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS l
4.9.6.1 Each manipulator crane used for movement of fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a load test of at least j
3250 pounds and demonstrating an automatic load cut off when the crane load j
exceeds 2850 pounds.
4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of control rods within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a load test of at least 700 pounds.
I NORTH ANNA - UNIT 2 3/4 9-7 I
i
r
=
REFUELING OPERATIONS CR ANE TR AVEL - SPENT FUEL PIT LIMITING CONDITION FOR OPERATION 1
3.9.7 Loads in excess of 2500 pounds shall be prohibited from travel over irradiated fuel assemblies in the spent fuel pit. This does not apply to movement of any spent fuel pit gate l
provided each of the following is satisfied:
the top of the gate (excluding lifting lugs) is no higher than 15 inches above the top of a.
the moveable platform crane deck support beam while over irradiated fuel,
- b. the gate is rigged to slack-free safety cables while over irradiated fuel, irradiated fuel containing Rod Control Cluster Assemblies are excluded along the load c.
path where the gate is moved, and
- d. irradiated fuel is prohibited in the cask area when the gate is lifted over the spent fuel cask handling area. There is no restriction on lift height.
APPLICABILITY: With irradiated fuel assemblies in the spent fuel pit.
ACTION:
With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.7.1 Loads other than the spent fuel pit gates shall be verified to be less than 2500 pounds l
prior to movement over irradiated fuel assemblies in the spent fuel pit.
4.9.7.2 For movement of any of the spent fuel pit gates:
gate lift height and slack-free redundant rigging shall be verified prior to moving over a.
irradiated fuel,
- b. load paths shall be verified not to have irradiated fuel with Rod Control Cluster Assemblies present in the gate load path, and the spent fuel cask handling area shall be verified to have no irradiated fuel present prior c.
to moving a gate over the area.
l I
I l
l NORTH ANNA - UNIT 2 3/4 9-8 Amendment No.194 S
f j
i REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE OPERABILITY The OPE.RABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movement of control rods and fuel assemblies; 2) each crane has sufficient load capacity to lift a control rod or fuel as embly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT The restriction on movements of the spent fuel pit gates over irradiated fuel ensure that redundant rigging will be used in order to prevent a gate drop caused by hoist failure. As shown l'
by calculation, in the event the load is dropped (1) the spent fuel storage racks limit gate penetration i
and prevent the impact load from being applied to stored fuel, (2) fuel spacing will not be changed and (3) impact loading to the spent fuel pit structure is acceptable.
The restriction on movement of other loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped,1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.
l 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures l
that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.
After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 6 is permitted, provided there is sufficiem decay heat removal to maintain the RCS temperature less than or equal to 140 F. Since the decay heat l
power production rate decreases with time after shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removal is provided as long as the reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140*F. The reduced flow rate provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation. During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a l
boron dilution incident and to prevent boron stratification.
The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR f
loop will not result in a complete loss of residual heat removal capability. With the reactor vessel i
head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.
g NORTH ANNA - UNIT 2 B 3/4 9-2 Amendment No. 440.194
i REFUELING OPERATIQMS BASES 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the l
containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.
3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND SPENT FUEL PIT The restrictions on minimum water level ensure that sufficient water depth is available to -
remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.
The minimum water level for movement of fuel assemblies (23 feet above the vessel flange) assures that sufficient water depth is maintained above fuel elements being moved to or from the vessel. With the upper internals in place, fuel assemblies and control rods cannot be removed from the vessel. Operations involving the lifting of control rods with the vessel upper internals in place may proceed with less than 23 feet of water above the vessel flange provided that 23 feet of water is maintained above all irradiated fuel assemblies within the reactor vessel.
3/4.912 FUEL BUILDING VENTILATION SYSTEM The limitations on the fuel building ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the auxiliary building HEPA and charcoal filter assemblies prior to discharge to the atmosphere. The Fuel Handling Accident analysis d' s not require filtration of the fuel building exhaust in order to meet the analysis criteria.
However, the OPERABILITY of this system and the resulting iodine removal capacity provide additional conservatism compared with the assumptions of the accident analyses.
NORTH ANNA - UNIT 2 B 3/4 9-3
' Amendment No.
94-l@.194
.