ML20133D606
| ML20133D606 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/22/1985 |
| From: | Butcher E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20133D610 | List: |
| References | |
| TAC-56844, TAC-56845, NUDOCS 8508070388 | |
| Download: ML20133D606 (8) | |
Text
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UNITED STATES g
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NUCLEAR REGULATORY COMMISSION y
- j WASHINGTON, D. C. 20555
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WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. DPR-24 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (thelicensee)datedJanuary 30, 1985 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 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.
8508070388 850722
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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 3.8 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. 94, 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 Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: July 22, 1985 i
1
8 UNITED STATES
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3 NUCLEAR REGULATORY COMMISSION 5
- 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. 98 License No. DPR-27 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applicat'on for amendment by Wisconsin Electric Power Company (the licensee) dated January 30, 1985 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.
J 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.98
. 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 Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: July 22, 1985 1
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ATTACHMENT TO LICENSE AMENDMENTS NO. 94 AND ga TO FACILITY OPERATING LICENSE NO. DPR-24 AND DPR-27 DOCKET N05. 50-266 AND 50-30I Revise Appendix A as follows:
Remove Pages Insert Pages 15.2.3-3 15.2.3-3 15.2.3-7 15.2.3-7 15.3.1-3b 15.3.1-3b 1
i I
(c) for each percent that the magnitude of q, - qu exceeds -17 percent, the AT trip setpoint shall be automaticaTly rsduced by an equivalent of 2.0 percent of rated power.
1 (5) Overpower AT (.)
3) saTo {K4-KI5T5 *I} (I
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5 6
5 4
4 where ATo = indicated AT at rated power, OF T
= average temperature, OF 0
T'
= 574.2 F K
1
= 1 089 of rated power 4
K
= 0.0262 for increasing T 5
= 0.0 for decreasing T K
= 0.00123 for T 1 T' g
= 0.0 for T < T' T
= 10 sec 5
f (al) as defined in (4) above, 3
= 2 sec for Rosemont or equivalent RTD T
= 0 sec for Sostman or equivalent RTD 4
= 2 sec for Rosemont or equivalent RTD v
= 0 sec for Sostman or equivalent RTD (6) Undervoltage - >75 percent of normal voltage (7)
Indicated reactor coolant flow per loop - >90 percent of normal indicated loop flow (8) Reactor coolant pump motor breaker open (a) Low frequency set point > 55.0 HZ l
(b) Low voltage set point > 75% of normal voltage 15.2.3-3 Unit 1 - Amendment No. 3, 28,Eh, H, 94 Unit 2 - Amendment No. 32, 70, 77,98
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,.
l which assures a trip signal before the pump inertia is reduced to an unacceptable value. The high pressurizer wtter level reactor trip protects the pressurizer safety valves against water relief. The specified set point 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 set point 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.I9)
~~
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 set point above which these trips are unblocked assures their avail-ability in the power range where needed. Specifications 15.2.3.2.A(1) and 15.2.3.2.C have a i 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.
I Sustained operating with only one pump will not be permitted above 10% power.
If a pump is lost while operating between 10% and 50% power, an orderly and immediate reduction in power level to below 10% 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 maximum enthalpy rise which occurs during full I
power and full flow operation.
References (1) FSAR 14.1.1 (4) FSAR 14.3.1 (7) PSAR 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. $$,94 Unit 2 - Amendment No.S@,98 15.2.3-7 c
m y
I 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 vill 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 (1) i 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 I
coolant pump is lost during operation between 10% and 15% of rated power.
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 value.(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 i-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 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 i
available for decay heat removal, this Specification requires both residual l
heat is*<<a1 loops to be operable unless the reactor system is in the refueling j
shutdown condition with the refueling cavity flooded and no operations in progress rhich 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 RER loop. Also, one residual heat removal loop may be temporarily out of service due to l
l Unit 1 - Amendment No. 55,66,76,93,94
- 15. 3.1-3b Unit 2 - Amendment No. 66,71,80,97, 98 l
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