ML20040B369
| ML20040B369 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/30/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20040B370 | List: |
| References | |
| NUDOCS 8201250476 | |
| Download: ML20040B369 (10) | |
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UNITED STATES y'
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NUCLEAR REGUL ATORY COMMISSION
- E WASHING TCN, D. C. 20555
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l NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO.1 AMENDMENT TO FACILITY OP! RATING LICENSE Amendment No. 5 3 License No. DPR-42 l.-
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for nendment by Northern States Power Company j
(the licensee) dated August 27, 1976, complies with the standards and requirements of the Atomic Energy Act of 195.4, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that tne activities authorized by this amendment can be conducted without endangering the health and safety cf 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.
8201250476 811230 PDR ADOCK 05000282 P
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l l 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. OPR-42 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 5 3, 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 the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
(
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g Robert A. Clark, Chief e
Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical
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Speci fica tions Date of Issuance: DEC 3 01981 t
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o UNITED STATES y
g fJUCLEAR REGULATORY COMMISSION 5-C WASHINGTON. D. C. 20555 o
I NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 4 7 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
1 A.
The application for amendment by Northern States Power Compt.ny
]
(the licensee) dated August 27, 1976, 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 i
Chapter I; i
B.
The facility will operate in confonnity 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 Connission'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.
kccordingly, 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 flo. DPR-60 is hereby amended to read as follows:
(2) Technical Specifications i
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 4 7, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical i
Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 0Ni /rkt e
q-gy Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing Attachnent:
Changes to the Technical Speci fica tions Date of Issuance: C;: 3 e '305 1
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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT N0. 5 3 TO FACILITY OPERATING LICENSE NO. DPR-42 AMENDMENT NO. 4 7 TO FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NOS. 50-282 AND 50-306 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 vertical lines indicating the area of change.
Remove Insert TS.3.4-2 TS.3.4-2 TS.3.6-3 TS.3.6-3 Table TS.4.1-1, (pg 3 of 6)
Table TS.4.1-1, (pg 3 of 6) i TS.4.4-6 TS.4.4-6 TS.4.4-9 TS.4.4-9 Table TS.6.7-1 I
1 l
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b I
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TS.3.4-2 For Unit 1 operation motor operated valves M732242 and MV32243 e.
shall have valve position monitor lights operable and shall be locked in the open position by having the motor control center supply breakers manually locked open.
For Unit 2, correspond-ing valve conditions shall exist.
f.
Essential features including system piping, valves, and inter-locks directly associated with the above components are operable, g.
Manual valves in the above systems that could (if one is im-properly positioned) reduce flow below that assumed for accident analysis shall be locked in the proper position for emergency use.
During power operation, changes in valve position will be under direct administrative control.
3.
Steam Exclusion System Both isolation dampers in each ventilation duct that penetrates rooms containing equipment required for a high energy line rupture outside of containment shall be operable or one damper in each duct with an inoperable isolation damper shall be closed.
4.
Radiochemistry The iodine-131 activity of the water on the secondary side of either steam generator for that reactor does not exceed 0.30 uci/cc.
B.
If, during startup operation or power operation, any of the conditions of Specification 3.4.A., except as noted below for 2.a. 2.b or 4 cannot be l
met, startup operations shall be discontinued and if operability cannot be restered within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the affected reactor shall be placed in the cold shutdown condition using nor6al operating procedures.
With regard to Specifications 2.a or 2.b. if a turbine driven AFW pump is not operable, that AFW pump shall be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the af f ected reactor shall be cooled to less than 350*F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If a motor driven AFW pump is not operable, that AlW pump shall be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or one unit shall be cooled to less than 350*F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Basis A reactor shutdown f rom power requires removal of decay heat.
Decay heat i
removal requirements are normally satisfied by the steam bypass to the condenser and by continued feedwater flow to the steam generators.
Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system.
I DPR-42 Amendment No. 17, 46, 5 3 DPR-60 Amendment No. 11, 40, 4 7
TS.3.6-3 9.
The valves and actuation circuits that isolate the auxiliary building normal ventilation system following an accident shall be considered operable for containment integrity if the ventilation system can be manually isolated within 6 minutes following an accident and the inoperable components are repaired within 7 days.
10.
The Auxiliary Building Special Ventilation Systen shall be considered operable only if the Turbine Building _
roof exhauster fans can be deenergized within 30 minutes of a loss of coolant accf ! ant.
B.
Containment Internal Pressure If the internal pressure of the containment vessel exceeds 2 psig whenever containment integrity is required, the condition shall be corrected within eight (8) hours or procedures shall be initiated tamediately to establish reactor conditions for which containment integrity is not needed.
C.
Containment and Shield Building Air Temperature If the average temperature of the air in the contaimment vessel exceeds 44*F above the average temperature of the air in the Shield Building whenever containment integrity is required, the condition shall be corrected within eight hours or procedures shall be initiated tamediately to establish reactor conditions for which containment integrity is not needed.
D.
Containment Shell Teeperature If the containment shell temperature becomes less than 30*F whenever containment integrity is required, the condition shall be corrected within eight hours or procedures shall be initiated t= mediately to establish reactor conditions for which containment integrity is not
- needed, i
DPR Amendment No. 17, 5 3 DPR Amendment No. 11,4 7 l
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TABil TS.4.1-1 (Par,e 3 of 5)
Channel Functional Response ee
&R Description Check Calihrate Test Test Remarks i
N ~
- 19. Radiation Monitoring
- D R
M NA Includes all channels used for leak detection per Spec. 3.1.D. and yy effluent release monitoring per Spec. 3.9 and 5.5.
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- 20. Boric Acid Make-up Flow NA R
NA NA Channel 19 n
- 21. Containment Sump Level NA R
R NA Includes Sumps A, B, and C z
f, O O
- 22. Accumulator Level and gg Pressure S
R R
NA k
- 23. Steam Generator Pressure S
R M
NA
- 24. Turbine First Stage S
R M
nA Pressure
- 25. Emergency Plan Radiation
- M R
M NA Includes those named in the emergency Instruments procedure (referenced in Spec. 6.5 A.6)
- 26. Protection Systems Logic NA NA M
NA Includes auto load sequencers Channel Testing
- 27. Turbine Overspeed Protec-NA R
H NA i
tion Trip Channel
- 23. Deleted
- 29. Deleted
- 30. Environmental Monitors M
NA NA NA Includes those used per Spec. 4.10
- 31. Seismic Monitors H
R NA NA Includes those reported in Item 4 of Table TS.6.7-1
- 32. Coolant Flow-RTD Bypass S
R H
NA Flowmeter
- 33. CRDM Cooling Shroud Exhaust S
NA R
NA FSAR page 3.2-56 Air Temperature
- 34. Reactor Gap Exhaust Air S
NA R
NA FSAR page 5.4-2 Temperature
l TS.4.4-6 E.
Containment Isolation Valves During each refueling shutdown, the containment isolation valves, shield building ventilation valves, and the auxiliary building normal ventila-tion system isolation valves shall be tested for operability by applying a simulated accident signal to them.
F.
Post Accident Containment Ven,tilation System During each refueling shutdown, the operability of system recirculating fans and valves, including actuation and indication, shall be demonstrated.
G.
Containment and Shield Building Air Temperature Prior to establishing reactor conditions requiring containment integrity, the average air temperature difference between the containment and its associated Shield Be *1dir.g shall be verified to be within acceptable limits.
H.
Containment Shell Temperature Prior to establishing reactor conditions requiring containment integrity, the temperature of the containment vessel wall shall be verified to be within acceptable limits.
Basis The containment systen consists of a steel containment vessel, a concrete shield building, the auxiliary building special ventilation cone (ABSVZ), a shield building ventilation system, and an auxiliary building special ventilation system.
In the event of a loss-of-coolant accident, a vacuum in the shield building annulus will cause most leakage f rom the containment vessel to be mixed in the annulus volume and recirculated througn a filter system before its deferred release to the environment through the exhaust fan that maintains vacuum.
Some of the leakage goes to the ABSVZ from which it is exhausted through a filter. A small fraction bypasses both filter systems.
The f rees tanding containment vessel is designed to acco=modate the fromtheDesignBasisAccident.7gyimum internal pressure that would result For initial conditions typical of normal operation, 120*F and 15 psia, an instan-taneous double-ended break with minimum safeguards results in a peak pressure of less than 46 psig at 268*F.
The containment will be strength-tested at 51.8 psig and Icak-tested at 46.0 psig to meet acceptance specifications.
The safety analysis (
(
is bdsed on a conservatively chosen reference set of assumptions regarding the sequence of events relating to activity release and attainment and maintenance of vacuum in the shield building annulus and the auxiliary building special ventilation zone, the effectiveness of filtering, and the leak rate of the containment vessel as a function of tLne.
The effects of variation in these assumptions, including that for leak rate, has been investigated thoroughly. A summary of the items of conservatism involved in the reference calculation and the magnitude of their ef fect upon off-site dose demonstrates the collective ef fectiveness of conservatism in these assumptions.
Unit 1 - Amendment No. 5 3 Unit 2 - Amendment No..i 7
TS.4.4-9 blocked off. Until these trays can be installed, to guarantee a representa-tive adsorbent sample, procedures should allow for the removal of a tray containing the oldest batch of adsorbent in each train, emptying of one bed from the tray, mixing the adsorbent thoroughly, and obtaining at least two samplea.
One sample will be submitted for laboratory analycis and the other held as a backup.
If test results are unacceptable, all adsorbent in the train will be replaced. Adsorbent in the tray removed for sampling will be renewed.
Any HEPA filters found defective will be replaced. Replacement charcoal adsorber and HEPA filters will be qualified in accordance with the intent of Regulatory Guide 1.52 - Rev. 1 June 1976.
If significant painting, fire, or chemical release occurs such that the HEPA filters or charcoal adsorbers could become contaminated from the fumes, chemicals, or foreign material, the same tests and sample analysis will be performed as required for operational use.
Operation of each train of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate operability of the system and remove excessive moisture which may build up on the adsorber.
Periodic checking of the inlet heaters and associated controls for each train will provide assurance that the system has the capability of reducing inlet air humidity so that charcoal adsorber ef ficiency is enhanced.
In-place testing procedures will be established utilizing applicable sections of ANSI N510 - 1975 standard as a procedural guideline only.
A minimum containment shell temperature of 30*F has been specified to provide assurance that an adequate margin above NDTT exists.
Evaluation of data collected during the first fuel cycle of Unit No. 1 shows that this Ibnit can be approached only when the plant is in cold shutdown.
Requiring containment shell temperature to be verified to be above 30*F prior to plant heatup f rom cold shutdown provides assurance that this temperature establishing conditions re' quiring containment integrity {9)above NDTT prior to A maximum temperature dif ferential between the average containment and annulus air temperatures of 44*F has been specified to provide assurance that offsite doses in the event of an accident remain below those calculated in the FSAR.
Evaluation of data collected during the first fuel cycle of Unit No. 1 shows that this limit can be approached only when the plant is in cold shutdown.
Requiring this temperature differential to be verified to,be less than 44*F prior to plant heatup from cold shutdown provides assurance that this para-meter is within accepgagle limits prior to establishing conditions requiring l
containment integrity i
References (1) FSAR, Section 5, and Appendix 14-C (2)
FSAR, Section 14, and Appendix G (3) Safety Evaluation Report, Sections 6.2 and 15.0 (4) FSAR, Section 14 (5) FSAR, Section 14.3.6 (6) Letter to NSP from AEC dated' November 29, 1973 i
(7)
NSP Report, " Prairie Island Containment Systems Special Analyses,"
dated April 9, 1976.
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i Unit 1 - Amendment No. 17, 5 3 Unit 2 - Amendment No. II, 4 7