ML20069L742
| ML20069L742 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  (DPR-24-A-066, DPR-24-A-66, DPR-27-A-071, DPR-27-A-71) |
| Issue date: | 11/08/1982 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20069L745 | List: |
| References | |
| TAC-42114, TAC-42115, TAC-52663, TAC-52664, NUDOCS 8211170562 | |
| Download: ML20069L742 (15) | |
Text
[
'q UNITED STATES
)' p, NUCLEAR REGULATORY COMMISSION
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-l WASHINGTON,0. C. 20555
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l WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 License No. DPR-24 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated November 16, 1981 as modified May 3, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules an1 ragulations 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.
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8 PDR ADOCK 05000266 P
. 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 Facility Operating License No. OPR-24 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 66, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective 20 days from the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION L (.
[
Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 8, 1982
t$"4 UNITED STATES NUCLEAR REGULATORY COMMISSION g
'j WASHINGTON, D. C. 20555
~%,,,,,/
WISCONSIN ELECTRIC POWER COMPANY DOCKET N0. 50-301 POINT BEACH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. DPR-27 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company
' he licensee) dated November 16, 1981 as modified May 3, 1982,
- nplies with the standards and requirements of the Atomic
.nergy Act of 1954, as amended (the Act) and the Commission's rules and regalations 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
. 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 Facility Operating License No. DPR-27 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised t'.irough Amendment No. 71, are hereby incorporated in the license. The licensee shall operate the facilit." in accordance with the Technical Specifications.
3.
This license amendment is effective 20 days from the.date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 4
Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing ant:
> to the Technical specifications Date of Issuance: November 8, 1982
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 66 TO FACILITY OPERATING LICENSE NO. DPR-24 AMENDMENT NO. 71 TO FACILITY OPERATING LICENSE NO. DPR-27 DOCKET NOS. 50-266 AND 50-301 1
i Revise Appendix A as follows:
Remove Pages Insert Pages 15.3.1-1 15.3.1-1 15.3.1-2 15.3.1-2 15.3.1-3 15.3.1-3 15.3.1-3a 15~3.1-3a 15.3.1-3b 15.3.3-2 15.3.3-2 15.3.3-2a 15.3.3-2a 15.3.3-8 15.3.3-8 15.3.3-9 15.3.3-9 15.3.8-1 15.3.8-1 4
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15.3 LIMITING CONDITIONS FOR OPERATION 15.3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.
Obiective To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe reactor operation.
Specification A.
OPERATIONAL COMPONENTS 1.
Coolant Pumps a.
At least one reactor coolant pump or the residual heat removal system shall be in operation when a reduction is made in the boron concentration of the reactor coolant.
b.
When the reactor is critical and above 1% of rated power except I
for natural circulation tests, at least one reactor coolant pump shall be in operation.
c.
(1) Reactor power shall not be maintained above 10% of rated power unless both reactor coolant pumps are in operation.
(2)
If'either reactor coolant pump ceases operating, immediate power reduction shall be initiated under administrative control as necessary to reduce power to less than 10% of rated power.
2.
One steam generator shall be operable whenever the average reactor coolant temperature is above 350*F.
3.
Components Required for Redundant Decay Heat Removal Cag, ability Reactor coolant temperature less than 350*F and greater than a.
140*F.
(1) At least two of the decay heat removal methods listed shall be operable.
(a) Reactor Coolant Loop A, its associated steam generator and either reactor coolant pump (b) Reactor Coolant Loop B, its associated steam generator and either reactor coolant pump 15.3.1-1 Unit 1 - Amendment No. 44, 66 Unit 2 - Amendment No. 49, 71
(c) Residual Heat Removal Loop (A)*
(d) Residual Heat Removal Loop (B)*
(2)
If the conditions of specification (1) above cannot be met, corrective action to return a second decay heat removal method to operable status as soon as possible shall be initiated immediately.
(3) At least one of the above decay heat removal methods shall be in operation except when required to be secured for testing.
(4)
If no decay heat removal metnod is in operation, all operations causing an increase in the reactor decay heat load or a reduc-tion in reactor coolant system boron concentration shall be suspended. Corrective actions to return a decay heat removal method to operation shall be initiated immediately.
b.
Reactor Coolant Temperature Less Than 140'F (1)
Both residual heat removal loops shall be operable except as permitted in items (3) or (4) below.
(2)
If no residual heat removal loop is in operation, all operations causing an increase in tne eactor decay heat load or a reduc-tion in reactor coolant system boron concentration shall be suspended.
Corrective actions to return a decay heat removal method to operation snall be initiated immediately.
(3)
One residual heat removal loop may be out of service when the reactor vessel head is removed and the refueling cavity flooded.
(4) One of the two residual heat removal loops may be temporarily out of service to meet surveillance requirements.
4.
Pressurizer Safety Valves I
a.
At least one pressurizer safety valve shall be operable whenever the teactor head is on the vessel.
b.
Both pressurizer safety valves shall be operable whenever the reactor is critical.
- dechanical and electrical design provisions of the residual heat removal system afford the necessary flexibility to allow an operable residual heat removal loop to consist of the RHR pump from one loop coupled with the RHR heat exchanger from the other loop and to allow the normal or emergency power source to be in-operable or tied together when the reactor coolant temperature is less than 200*F.
15.3.1-2 Unit 1 - Asendment No. 44, 55, 66 Unit 2 - Amendment No. 49, $$. 71
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5.
Pressurizer Power Operated Relief Valves (PORV) and PORV Block Valves I
a.
Two PORVs and their associated block valves shall be operable.
(1)
If a PORV is inoperable, the PORV shall be restored to an operable condition within one hour or the associated block valve shall be closed.
(2)
If a PORV block valve is inoperable, the block valve shall be restored to an operable condition within one hour or the block valve shall be closed with power removed from the block valve; otherwise the unit shall be in hot shutdown within the next six hours.
6.
The pressurizer shall be operable with at least 100 KW of pressurizer I
heaters available and a water level greater than 10% and less than 95% during steady-state power operation. At least one bank of pressurizer heaters shall be supplied by an emergency bus power supply.
Basis When the boron concentration of the reactor coolant system is to be reduced, the ess must be uniform to prevent sudden reactivity changes in the reactor, of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the primary system volume in approximately one-half hour. The pressurizer is of little concern because of the lower pressurizer volume and because pressurizer boron concentration normally will be higher than that of the rest of the reactor coolant.
Specification 15.3.1.A.1 requires that a sufficient number of reactor coolant pumps be operable to provide core cooling in the event a loss of power occurs.
The flow provided in each case will keep DNBR well above 1.30 as discussed in FFDSAR, Section 14.1.9.
Therefore, cladding damage and release of fission products to the reactor coolant will not occur. Heat transfer analyses ( } show that reactor heat equivalent to 10% of rated power can be removed with natural circulation only; hence the specified upper limit of 1% rated power without operating pumps provides a substantial safety factor.
Item 15.3.1. A.l.c.(2) permits an orderly reduction in power if a reactor coolant pump is lost during operation between 10% and 50% of rated power.
15.3.1-3 Unit 1 - Amendment No. d4, 55, 66 Unit 2 - Amendment No. A9, 69, 71
f
)
Above 50% power, an automatic reactor trip will occur if either pump is lost.
The power-to-flow ratio will be maintained equal to or less than 1.0, which ensures that the minimum DNB ratio increases at lower flow since the maximum en**
lpy rise does not increase above its normal full-flow maximum value.( )
Specification 15.3.1.A.3 provides limiting conditions for operation to ensure that redundancy in decay heat removal methods is provided. A single reactor coolant loop with its associated steam generator and a reactor coolant pump or a single residual heat removal loop provides sufficient heat removal capacity for removing the reactor core decay heat; however, single failure considerations require that at least two decay heat removal methods be available. Operability of a steam generator for decay heat removal includes two sources of water, water level indication in the steam generator, a vent path to htmosphere, and the Reactor Coolant System filled and vented so thermal convection cooling of the core is possible.
If the steam generators are not available for decay heat removal, this Specification requires both residual heat removal loops to be oper-able unless the reactor system is in the refueling shutdown condition with the refueling cavity flooded and no core alterations in progress. In this condition, the' reactor vessel is essentially a fuel storage pool and removing a RRR loop from service provides conservative conditions should operability problems develop in the other RHR loop. Also, one residual heat removal loop may be temporarily out of service due to surveillance testing, calibration, or inspection requirements.
The surveillance procedures follow administrative controls which allow for timely restoration of the residual heat removal loop to service if required.
Each of the pressurizer safety valves is designed to relieve 288,000 lbs. per hour of saturated steam at setpoint.
If no residual heat is removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve, there-fore, provides adequate defense against overpressurization. Below 350*F and 400 psig in the Reactor Coolant System, the residual heat removal system can remove decay heat and thereby control system temperature and pressure.
A PORV is defined as OPERABLE if leakage past the valve is less than that allowed in Specification 15.3.1.D and the PORV has met its most recent channel test as
]
specified in Table 15.4.1-1.
The PORVs operate to relieve, in a controlled 15.3.1-3a Unit 1 - Amenc' ment No. 44, 55, 66 Unit 2 - Amendment No. 49, 69, 71
manner, reactor coolant system pressure increases below the setting of the pressurizer safety valves. These PORVs have remotely operated block valves to provide a positive shutoff capability should a PORV become inoperable.
The requirement that 100 KW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain pressure control and natural circulation at hot standby.
References (1) FSAR Section 14.1.6 (2) FSAR Section 7.2.3 i
15.3.1-3b Unit 1 - Amendment No. 55, 66 Unit 2 - Amendment No. 69, 71
safety injection system are in the open position.
g.
All valves, interlocks, and piping associated with the above components and required to function during accident conditions are operable.
h.
During conditions of operation with reactor coolant system pressure in excess of 1,000 psig, the source of AC power shall be remcved from the accumulator isolation valves MOV-841A and B at the motor control center and the valves shall be open.
1.
Power may be restored to MOV-841A and B for the purpose of valve testing or maintenance providing the testing and maintenance is completed and power is removed within four hours.
2.
During power operation, the requirements of 15.3.3.A.1, Items b and c, l
may be modified to allow one of each of the following components to be inoperable at any one time.
If the system is not restored to meet the requirements of 15.3.3.A.1 within the time period specified, the reactor shall be placed in the hot shutdown condition.
If the requirements of 15.3.3.4.1 are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor thall be placed in the cold shutdown condition.
One accumulator may be isolated for a period of up to one hour to a.
permit a check valve leakage test.
Before isolating an accumulator, the other accumulator isolation valve shall be checked open.
b.
One safety injection pump may be out of service, provided the pump is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The other safety injection pump shall be tested to demonstrate operability prior to initiating repair of the inoperable pump.
c.
Any valve in these systems required to function during accident conditions may be inoperable provided repairs are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Prior to initia ing repairs, all valves in the system that provide the duplicate function shall be tested to demonstrate operability.
3.
During power operation, the requirements of 15.3.3.A.1, Items d and e, may be modified to allow one of each of the following components to be inoperable at any one time.
If the component is not restored to meet 15.3.3-2 Unit 1 - Amendment No. 66 Unit 2 - Amendment No. 71
the requirementa of 15.3.3.A.1 within the time specified, the reactor shall be placed in the hot shutdown condition.
If the requirements of 15.3.3.A.1 are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be maintained in a condition with reactor coolant temperatures between 500 and 350*F, unless one residual heat removal loop is being relied upon to provide redundancy for decay heat removal.
In this case the reactor shall be maintained between 350* and 140*F.
a.
One residual heat removal pump may be out of service, provided the pump is restored to operable statos within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The other residual heat renoval pump shall be tested to demonstrate operability prior to initiating repair of the inoperable. pump.
b.
One residual heat exchanger may be out of service for a period of no more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
c.
Any valve in the system, required to function during accident condi-tions, may be inoperable provided repairs are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Prior to initiating repairs, all valves in the system that provide the duplicate function shall be tested to demonstrate operability.
15.3.3-2a Unit 1 - Amendment No. 66 Unit 2 - Amendment No. 71
Assuming the reactor has been operating at full rated power for at least 100 days, the magnitude of the decay heat decreases as follows after initiating hot sh'utdown.* i Time After Shatdown Decay Heat % of Rated Power 1 min.
3.6 30 min.
1.55 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.25 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.7 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.4
- Based on ANS 5.1-1979, " Decay Heat Power in Light-Water Reactors" Thus, the requirement for core cooling in case of a postulated loss-of-coolant accident while in the hot shutdown condition is significa,ntly reduced below the requirements for a postulated loss-of-coolant accident during power operation.
Putting the reactor in the hot shutdown condition significantly reduces the poten-tial consequences of a loss-of-coolant accident, and also allows more free access to some of the engineered safety system conponents in order to effect repairs.
Failure to complete safety injection system repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to l
the hot shutdown condition is considered indicative of a requirement for major maintenance and, therefore, in such a case, the reactor is to be put into the cold shutdown condition. When the failures involve the residual heat removal system, in order to insure redundant means of decay heat removal, the reactor system may remain in a condition with reactor coolant temperatures between 500 and 350*F so that the reactor coolant loops and associated steam generators may be utilized for redundant decay heat removal. However, when the remaining RHR loop must be relied upon for redundant decay heat removal capability, reactor coolant temperatures shall be maintained between 350*F and 140*F.
With respect to the core cooling function, there is some functional redundancy for certain ranges of break sizes.( )
The containment cooling function is provided by two independent systems:
(a) fan coolers and (b) containment spray which, with sodium hydroxide addition, provides the iodine removal function. During normal power operation, only three of the four fan coolers are required to remove heat lost from equipment and piping within the containment.(
in the event of a Design Basis Accident, any one of the 15.3.3-8 Unit 1 - Amendment No. 66 Unit 2 - Amendment No. 71 i
l i
following combinations will provide sufficient cooling to reduce containmen't pressure:
(1) four fan coolers, (2) two containment spray pumps, (3) two fan coolers plus one containment spray pump.(4)
Sodium hydroxide addition via one spray pump reduces airborne iodine activity sufficiently to limit off-site doses to acceptable values. One of the fotr fan coolers is permitted to be inoperable when the reactor is made critical and during power operation.
The component cooling system is different from the other systems discussed above in that the components are so located in the Auxiliary Building aa to be acces-sible for repair after a loss-of-coolant accident. One component cooling water pump together with one component cooling heat exchanger can accommodate the heat removal load on one unit either following a loss-of-coolant accident, or during normal plant shutdown.
If during the post-accident phase the component cooling water supply is lost, core and containment cooling could be maintained until repairs were effected.(5)
A total of six service water pumps are installed, only three of which are required to operate during the injection and recirculation phases of a postulated loss-of-coolant accident,( } in one unit together with a hot shutdown condition in the other unit.
References (1) FSAR Section 3.2.1 (2) FSAR Section 6.2 (3) FSAR Section 6.3.2 (4) FSAR Section 6.3 (5) FSAR Section 9.3.2 (6) FSAR Section 9.6.2 15.3.3-9 Unit 1 - Amendment No. 66 Unit 2 - Amendment No. 71
15.3.8 REFUELING AND SPENT FUEL ASSEMBLY STORAGE Applicability:
Applies to operating limitations during refueling operations and to operating limitations concerning the movement of heavy loads over or into the spent fuel storage pools.
Obiective:
To ensure that no incident could occur during refueling operations, or during auxiliary building crane operations that would affect public health and safety.
Specificationi:
A.
During refueling operations:
1.
The equipment hatch shall be closed and the personnel locks shall be capable of being closed. A temporary third door on the outside of the personnel lock shall be in place whenever both doors in a personnel lock are open (except for initial core loading).
2.
Radiation levels in fuel handling areas, the containment and spent fuel storage pool shall be monitored continuously.
3.
Core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audibic indication in the containment available whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service, i
4.
At least one residual heat removal loop shall be in operation. However, if refueling operations are affected by the residual heat removal loop flow, the operating residual heat removal loop may be removed from opera-tion for up to one hour per eight hour period.
5.
During reactor vessel head removal and while loading and unloading fuel from the reactor, a minimum boron concentration of 1800 ppm shall be maintained in the primary coolant system.
15.3.8-1 Unit 1 - Amendment No. 35, 66 Unit 2 - Amendment No. 41, 71 l
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