ML20140B950
| ML20140B950 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 03/17/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Nebraska Public Power District (NPPD) |
| Shared Package | |
| ML20140B954 | List: |
| References | |
| DPR-46-A-096, TAC 60925 NUDOCS 8603250039 | |
| Download: ML20140B950 (5) | |
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION E
WASMNGTON, D. C. 20666 g
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NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION Al ADMENT TO FACILITY OPERATING LICENSE Amendment No. 96 License No. DPR-46 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Nebraska Public Power District dated March 11, 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 Comission; i
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 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 licensee is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-46 is hereby amended to read as follows:
8603250039 860317 PDR ADOCK 03000290 P
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(2) Technical Specification The Technical-Specifications contained in Appendices A and B, as revised through Amendment No. 96, are hereby incorporated in the license.
The licensee shall operate the facility in accordance l
with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION wf
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Daniel R. Muller, Director BWR Project Directorate #2 i
Division of BWR Licensing I
Attachment:
Changes to the Technical a
Specifications I
j Date of Issuance:
March 17,1986 4
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ATTACHMENT TO LICENSE AMENDMENT NO. 96 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by rnarginal lines.
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COOPER
' LEAR STATION
,2 TABl.E
.A (Pcgs I)
PRIMARY CONTAINHENT AND REACTOR VESSEL ISOLATION INSTRUHENTATION Minimum Number Action Required &
of Operalite Components Component Operabil:
Instrument Instrument I.D. No.
Setting Limit Per Trip System (1) is Not Assured (2)
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Hain Steam Line liigh RHP-RH-251. A.B.C,&D
< 3 Times Full Power 2
A or B Rad.
R rctor I.ow Water Level NBI-LIS-101, A.B.C,&D 1+12.5"' Indicated Level 2(4)
A or B Bretor Low Low Water NBI-LIS-57 A & B #2
>-37" Indicated Level 2
A or B Level NBI-LIS-58 A & B #2
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Rxctor Low Low Low Water NBI-LIS-57 A & B #1
>-145.5" Indicated Level Level NBI-LIS-58 A & B #1 2
A or B Main Steam Line Leak HS-TS-121, A.B.C,&D
< 200*F 2(6) 3 Detection 122, 123, 124, 143, 144 145, 146, 147, 148, 149, j
150 Main Steam Line liigh MS-dPIS-Il6 A,B,C,&D
< 150% of Rated Steam 2(3)
B Flow 117, 118, 119 Flow l
jj Main Steam Line Low HS-PS-134, A,B,C,&D
> 825 pelg 2(5)
B
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Prsssure liigli Drywell Pressure PC-PS-12. A B.C,&D
< 2 psig 2(4)
A or B' I'
Illgh Reactor Pressure RR-PS-128 A & B
< 75 psig 1
D i
Hain Condenser Low HS-PS-103, A.B.C,&D 1 7" Hg (7) 2 A or B Vacuum j
Rxctor Wator Cleanup RWCU-dPIS-170 A & B
< 200% of System Flow
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System liigli Flow I
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3.2 ' BASES:
(Cent'd) and the guidelines of 10CFR100 vill not be exceeded. For large breaks up to the complete circumferential break.o.f a 28-inch recirculation line and vith the trip setting given above, CSCS initiation and primary system
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isolation are initiated in time to meet the above criteria.
Reference Paragraph VI.5.3.1 USAR.
The high dryvell pressure instrumentation is a diverse signal for mal-functions to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 and 6 iselation valves.
For the-breaks discussed above, this instrumentation vill generally initiate CSCS operation before th'a low-low-low water level instrumentation; thus the results given above are applicable here also. The water level instrumen-tation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instru-
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. mentation is to detect a break in the main steam line.
For the worst case of accident, main steam line break outside the dryvell, a trip setting of 150I of rated steam flow in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel clad ' temperatures peak at approximately 1000*F and release of radioactivity to the environs is below 10CTR100 guidelines.
Reference Section IIV.6.5 USAR.
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Temperature mesitoring instrumentation is provided in the main steam tunnel and along the steam line in-the turbine building to detect leaks
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in these areas.
Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.
See Spec. 3.7 for Valve Group. The setting is 200*F for the main steam leak detection system.
For large breaks, the high steam flow instrumentatign is a backup to the temp. instrumentation.
High radiation monitors in the main steam tunnel have been provided to detect gross fuel failure as in the control rod drop accident.
With the established setting of 3 times normal background, and main steam line isolation valve closure, fission product release is limited so that 10CTR100 guidelines.are not exceeded for this accident. Reference Sec-tion IIV.6.2 USAR.
Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below Speci-fication 2.1.A.6.
The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the turbine bypass valves when not in the RUN Mode is less severe than the loss of feedvater analyzed in Section IIV.5 of the USAR, therefore, closure of the Main Steam Isolation valves for thermal transient protection when not in RUN mode is not required.
The Reactor Water Cleanup System high flow and temperature instrumentation are arranged similar to that for the HPCI. The trip settings are such that core uncovery is prevented and fission product release is within limits.
i n nda.nt no. f.g sv. 96 i 7
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