ML20078G824
| ML20078G824 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/06/1983 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078G827 | List: |
| References | |
| TAC-51596, TAC-51597, NUDOCS 8310130202 | |
| Download: ML20078G824 (10) | |
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UNITED STATES
[ {%,;. 7 i NUCLEAR REGULATORY COMMISSION C
WASHINGTON, D. C. 20555 7
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DOCKET NO. 50-266 a
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POINT BEACH NUCLEAR PLANT, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE s
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Amendment No. 76 License No. OPR-24 1.
The Nuclear Regulatory Commission (the Commission) has found that:
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A.
fheapplicationforamendmentbyWisconsinElectricPowerCompany (the licensee) dated May 4, 1983, 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;
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The facility will operate in conformity with the application, B.
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; l
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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8310130202 831006 PDR ADOCK 05000266 P
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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 amen 4d to read as follows:
B.
Technical specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 76, 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
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1 James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 6, 1983
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C, UNITED STATES i
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DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 80 License No. OPR-27 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated May 4, 1983, 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 i
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 i
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. OPR-27 is hereby amended to read as follows:
B.
Technical specifications l
t The Technical Specifications contained in Appendices A and B, as revised througn Amendment No. 80, 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
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,. ~ ~~ h James R.' Miller, Chief Operating Reactors Branch #3 Division of Licensing I
Attachment:
Changes to the Technical 1
Specifications Date of Issuance: October 6,1983 i
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ATTACHf1ENT TO LICENSE AMENDMENTS AMENDMENT N0. 76 TO FACILITY OPERATING LICENSE NO. DPR-24 AMENDMENT N0. 80 TO FACILITY OPERATING LICENSE N0. DPR-27 DOCKET N05. 50-266 AND 50-301 Revise Appendix A as follows:
Remove Pages Insert Pages 15.3.1.3a 15.3.1-3a 15.3.3-9 15.3.3-9 15.3.10.-6 15.3.10-6 15.3.10-7 15.3.10-7 Table 15.4.1-1 Table 15.4.1-1 Page 4 of 4 Page 4 of 4 4
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 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 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 reactor systen is in the refueling shutdown condition 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 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, therefore, 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 - Amendment No. 35, 66, 76 Unit 2 - Amendment No. 60, 71, 80
following combinations will provide sufficient cooling to reduce containment pressure:
(1) four fan coolers, (2) two containment spray pumps, (3) two fan coolers plus one containment spray pump.( )
Sodium hydroxide addition via one spray pump reduces airborne iodine activity sufficiently to limit off-site doses to acceptable values.
One of the four fan coolers is permitted to be inoperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during power operation.
l The component cooling system is different from the other systems discussed above in that the components are so located in the Auxiliary Building as 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,(6) in one unit together with a hot shutdown condition in the other unit.
l References (1) FSAR Section 3.2.1 i
(2) FSAR Section 6.2 (3) FSAR Section 6.3.2 (4) FSAR Section 6.3 (5) FSAR Section 9.3.2 f
(6) FSAR Section 9.6.2 15.3.3-9 Unit 1 - Amendment No. - 66, 76 Unit 2 - Amendment No. 11, 80
a.
The RCCA does not drop upon removal of stationary gripper coil voltage.
b.
The RCCA does not step in properly when the proper voltage sequences are applied to the control rod drive mechanism coils.
It shall then be assumed inoperable until it has been tested to verify that it does drop.
c.
If the bank demand position is greater than or equal to 215 steps, or, less than or equal to 30 steps, and the rod position indicator channel shows a misalignment from the bank demand position of 15 inches.
The RCCA shall be assumed inoperable until it has been tested to verify that it does step properly.
d.
If the bank demand position is between 215 steps and 30 steps, and the rod position indicator channel shows a misalignment from the bank demand position of 7.5 inches. The RCCA shall be assumed inoperable until it has been tested to verify that it does step properly.
2.
Specification 15.3.10.C.l.b can be modified by the following:
a.
If an RCCA does not step in upon demand, up to six hours is 4
allowed to determine whether the problem with stepping is an electrical problem.
If the problem cannot be resolved within six hours, the RCCA shall be assumed inoperable until it has been verified that it will step in or would drop upon demand.
b.
If more than one RCCA does not step in, apparently due to electrical problems, the situation shall be rectified or clearly defined that it is an electrical problem and the RCCAs are l
capable of dropping upon demand or an orderly shutdown shall commence within six hours.
3.
No more than one inoperable RCCA shall be permitted during sustained I
power operation.
4.
When it has been determined that an RCCA does not drop on removal of 4
stationary gripper coil voltage, the shutdown margin shall be main-tained by boration as necessary to compensate for the withdrawn worth of the inoperable RCCA.
If sustained power operation is
<4ticipated, the insertion limit shall be adjusted to reflect the worth of the inoperable RCCA.
i 15.3.10-6 Unit 1 - Amendment No. 49, 76 Unit 2 - Amendment No. $$, 80 i
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D.
Misaligned or Dropped RCCA 1.
If the rod position indicator channel is functional and the asso-ciated RCCA is more than 7.5 inches indicated out of alignment with its bank demand position and cannot be aligned when the bank demand position is between 215 steps and 30 steps, then unless the hot channel factors are shown to be within design limits as specified in Section 15.3.10.B-1 within eight (8) hours, power shall be reduced to less than 75% of Rated Power. When the bank demand position.is l
greater than or equal to 215 steps, or less than or equal to 30 steps, the allowable indicated misalignmen,t is 15 inches between the rod position indicator and the bank demand position.
l 2.
To increase power above 75% with an RCCA more than 7.5 inches indi-cated out of alignment with its bank demand position when the bank demand position is between 215 steps and 30 steps, an analysis shall first be made to determine the hot channel factors and the resulting allowable power level based on Section 15.3.10.B.
When the bank demand position is greater than or equal to 215 steps, or less than i
or equal to 30 steps, the allowable indicated misalignment is 15 inches between the rod position indication and the b ak demand position.
3.
If it is determined that the apparent misalignment or dropped RCCA indication was caused by rod position indicator channel failure, sustained power operation may be continued if the following condi-l tions are met:
a.
For operation between 10% power and Rated Power, the position of the RCCA(s) with the failed rod position indicator channel (s) will be checked indirectly by core instrtmentation (excore detectors, and/or thermocouples, and/or moveable incore detec-tors) every shift'and after associated bank motion exceeding 24 steps in one direction.
b.
For operation below 10% of Rated' Power, no special monitoring is required.
E.
RCCA Drop Times l.
At operating temperature and full flow, the drop time of each RCCA shall be no greater than 1.8 seconds from the loss of stationary gripper coil voltage to dashpot entry.
15.3.10-7 Unit 1 - Amendment No. 49, 76 Unit 2 - Amendment No. 53, 80
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TABLE 15.4.1-1 (Continued)
S - Each shift M - Monthly D - Daily P - Prior to each startup if not done previous week.
W - Weekly R - Each Refueling Shutdown (But not to exceed 18 months).
l B/W - Biweekly N.A. - Not applicable.
Not required during periods of refueling shutdown, but must be performed prior to starting up if it has not been performed during the previous surveillance period.
Not required during periods of refueling shutdown if steam generator vessel temperature is greater than 70*F.
When used for the overpressure mitigating system each PORV shall be demonstrated operable by:
Performance ~ of a channel functional test on the PORV actuation channel, but excluding valve a.
operation, within 31 days prior to entering a condition in which the PORV is required operable and at least once per 31 days thereafter when the PORV is required operable.
b.
Testing valve operation in accordance with the inservice test requirements of the ASME Boiler and Pressure Vessel Code,Section IX.
Unit 1 - Amendment No. 3$, f/, 55, 76 Unit 2 - Amendment No. 50, 55, 60, 80
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