ML20203B208
| ML20203B208 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/03/1986 |
| From: | Colburn T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203B212 | List: |
| References | |
| TAC-57555, TAC-57556, NUDOCS 8607180084 | |
| Download: ML20203B208 (9) | |
Text
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NUCLEAR REGULATORY COMMISSION g
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WASHINGTON, D. C. 20555
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e WISCONSIN ELECTRIC POWER COMPANY DOCKET N0. 50-266 POINT BEACH NUCLEAR FLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 103 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 April 10, 1985 as modified February 14 and March 10, 1986 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 'his 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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, 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-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.103, 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 fromsthe date of issuance.
FORTHENUCLEARREGULATORY'tQtMISSION N
bYw Timothy G. Colburn, Project Manager s
PWR Project Directorate #1
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Division of PWR Licensing-A N
Attachment:
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Changes to the Technical x '
Specifications Date of Issuance:
July 3, 1986
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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WISCONSIN ELECTRIC POWER COMPANY DOCKET N0. 50-301 POINT BEACH NUCLEAR PLANT, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 106 License No. DPR-27 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated April 10, 1986 as modified February 14, and March 10, 1986 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 (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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
. 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 through Amendment No.106, 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 issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f'
A -$
Timothy G. Colburn, Proj anager i
PWR Project Directorate #1 Division of PWR Licensing-A
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 3, 1986 I
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i ATTACHMENT TO LICENSE AMENDMENT NOS.103 AND106 TO FACILITY OPERATING LICENSE NOS. OPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendmant number and contain marginal lines indicating the area of change.
REMOVE INSERT 15.2.3-7 15.2.3-7 15.3.1-1 15.3.1-1 15.3.1-3b 15.3.1-3b 15.3.1-3d 15.3.1-3d 4
3 4
the reactor coolant pump breaker opening as actuated by either high current, low supply voltage or low electrical frequency, or by a manual control switch. The significant feature of the breaker trip is the frequency setpoint, 55.0 HZ, which assures a trip signal before the pump inertia is reduced to an unacceptable value. The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. The specified setpoint allows adequate operating instrument error (2) and transient overshoot in level before the reactor trips.
The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.(9)
Numerous reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations.
The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed. Specifications 15.2.3.2.A(1) and 15.2.3.2.C have 1% tolerance to allow for a 2% deadband of the P10 bistable which is used to set the limit of both items. The difference between the nominal and maximum allowed value (or minimum allowed value) is to account for "as measured" rack drift effects.
Sustained operation with only one pump will not be permitted above 3.5 percent power. If a pump is lost while operating between 3.5 percent and 50 percent power, an orderly and immediate reduction in power level to below 3.5 percent is allowed. The power-to-flow ratio will be maintained equal to or less than unity, which ensures that the minimum DNB ratio increases at lower flow because the maximum enthalpy rise does not increase above the <naximum enthalpy rise which occurs during full power and full flow operation.
References (1) FSAR 14.1.1 (4) FSAR 14.3.1 (7) FSAR 3.2.1 (2) FSAR, Page 14-3 (5) FSAR 14.1.2 (8) FSAR 14.1.9 (3) FSAR 14.2.6 (6) FSAR 7.2, 7.3 (9) FSAR 14.1.11 Unit 1 Amendment No. 86, H, 103 15.2.3-7 I
15.3 LIMITING CONDITIONS FOR OPERATION 15.3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.
Objective To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe reactor operation.
Specification A.
OPERATIONAL COMP 0NENTS 1.
Coolant Pumps
- a.
When the reactor is critical, except for tests, at least one reactor coolant pump shall be in operation.
(1) Reactor power shall not be maintained above 3.5% 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 3.5% of rated power.
(3) If both reactor coolant pumps cease operating and power is greater than 3.5% of rated power, but less than 10% of rated power, reactor shutdown will comence immediately and verify the reactor trip breakers are opened within one hour.
b.
When the reactor is subcritical and the average reactor coolant temperature is greater than 350'F, except for tests, at least one reactor coolant pump shall be in operation.
(1) Both reactor coolant pumps may be deenergized provided:
a.
No operations are permitted that would cause dilution of the reactor coolant system boron concentration.
b.
Core outlet temperature is maintained at least 10*F below saturation temperature, and c.
The reactor trip breakers are open.
c.
At least one reactor coolant pump or residual heat removal system shall be in operation when a reduction is made in the boron concentration of the reactor coolant.
2.
- a.
One steam generator shall be operable whenever the average reactor coolant temperature is above 350 F.
3.
Corrponents Required for Redundant Decay Heat Removal Capability
- a.
Reactor coolant temperature less than 350 F and greater than 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
- Applicable only when one or more fuel assemblies are in the reactor vessel.
15.3.1-1 Unit 1 Amendment No. 44,56,82,103 Unit 2 Amendinent No. 49, 7J, 86,106
Specification 15.3.1.A.1 requires that at leart one reactor coolant pump must be operating whenever the average reactor coolant temperature"is above 350 F unless the listed restrictions are established. This is required'sc that the FSAR zero power transients (rod withdrawal from subcritical and rod ejection) are addressed from conservative conditions. With the reactor subcritical, with required shut-down margin, and with the trip breakers open, a single rod ejection will not result in criticality being reached. With the reactor subcritical and the average reactor coolant temperature above 350*F, a single reactor coolant pump provides sufficient decay heat removal capability. Heat transfer analysesII) show that reactor heat equivalent to 3.5% of the rated power can be removed with natural circulation only.
F Items 15.3.1.A.I.a.(2) permits an orderly reduction in power if a reactor,
coolant pump is lost during operation between 3.5% and 50% of rated pcwer.
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 enthalpy rise does not increase above its normal full-flow maximum lue.(2)
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 hert; however, single failure consideraticns require that at least two decay heat removal methods be avail-able. 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 atmosphere, 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 operable unless the ree.ctor system is in the refueling shutdown conditior, with the refueling cavity flooded and no operations in progress which could cause an increase in reactor decay heat load or a decrease in boron concentration.
In this condition, the reactor vessel is essentially a fuel storage pool and removing a RHR 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 du,e to Unit 1 Amendment No. 55, 66, 76, 15.3.1-3b N, H,103 Unit 2 Amendment No. 60, 71, 80, 97, 98, 106
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i restricts leakage so that, in the event of a pipe break or isolation valve failure, makeup water for the leakage can be provided by a single coolant charging pump. If a RCGVS vent path from either the pressurizer or reactor vessel head is inoperable, Specification 15.3.1.A.7.c requires the remotely operable valves in that inoperable path to be shut with power removed.
If a vent path from the common header to the pressurizer relief tank or contain-ment atmosphere is inoperable, the isolation valve in that path must be shut but reactor operations may continue.
If both vent paths to or both vent paths from the common header are inoperable, the RCGVS is inoperable and the steps in specification 15.3.1.A 7.d must be taken.
F e
(1) FSAR Section 14.1.11.
(2) FSAR Section 7.2.3.
Unit 1 Amendment No. 93, 103 Unit 2 Amendment No. 97, 106 15.3.1-3d
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