ML20035F200
| ML20035F200 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/14/1993 |
| From: | Hubbard G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20035F201 | List: |
| References | |
| GL-88-01, GL-88-1, NUDOCS 9304210021 | |
| Download: ML20035F200 (6) | |
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UNITED STATES I
NUCLEAR REGULATOP.Y COMMISSION
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NEBRASKA PUBLIC POWER DISTRICT l
DOCKET NO. 50-298 COOPER NUCLEAR STATION i
l AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.161 License No. DPR-46 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nebraska Public Power District (the licensee) dated February 1, 1993, 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 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 license 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.
9304210021 930414 PDR ADOCK 05000298 P
PDR L
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i' 2.
Accordingly, the license is amended by changes to the Technical Specifi-l cations as indicated in the attachment to this license amendment and i
Paragraph 2.C.(2) of Facility Operating License No. DPR-46 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.161, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
l 3.
The license amendment is effective 30 days after its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 4
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George T. Hubbard, Acting Director Project Directorate IV-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 14, 1993 t
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ATTACHMENT TO LICENSE AMENDMENT NO.161 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 pages are identified by Amendment number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES 135 135 c
137 137 149 149
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LIMITING CONDTTIONS FOR OPERATION SURVEILLANCE REOUIREMENTS l
3.6.C.
Coolant Leakare 4.6.C.
Coolant 12akare l
1.
Any time irradiated fuel is in the 1.
Reactor coolant system leakage shall the reactor vessel and reactor cool-be checked by the sump flow measur-ant temperature is above 212'F, ing systems and drywell air sampling j
reactor coolant leakage into the system and recorded at least once primary containment shall not exceed per shift, not to exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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a 5 gpm unidentified leak rate, 25 gpm identified leak rate, or a 2 gpa increase in unidentified leak rate within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
If these limits cannot be met, an orderly SHUTDOWN shall be initiated and the reactor shall be in a COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
t 2.
Each of the sump flow measuring systems shall be operable during REACTOR POWER OPERATION.
From and after the date that one of these systems is made or found to be inop-erable for any reason and the sump j
flow leak rate cannot be quantified, REACTOR POWER OPERATION is permis-sible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless the system is sooner l
i made OPERABLE.
If leakage can be l
quantitatively measured by manually i
pumping the sump or measuring the difference in sump level, then REAC-TOR POWER OPERATION is permissible during the succeeding 30 days, un-less the sump flow measuring system is sooner made OPERABLE.
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3.
The drywell air sampling system i
shall be OPERABLE during REACTOR POWER OPERATION. From and after the date that this system is made or found to be inoperable for any rea-i son, REACTOR POWER OPERATION is permissible only during the succeed-1 irg 30 days unless the system is sooner made OPERABLE.
4 If the requirements of specification 3.6.C.2 or 3 cannot be met, an or-derly SHUTDOWN.shall be initiated and the reactor shall be in a COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Amendment No. 4, 161
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LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS 3.6.E.
Jet Puros 4.6.E.
Jet Pumns i
1.
Whenever the reactor is in the 1.
Whenever there is recirculation flow STARTUP or RUN MODES, all jet pumps with the reactor in the STARTUP or shall be OPERABLE. If it is deter-RUN modes, jet pump OPERABILITY mined that a jet pump is inoperable, s} all be checked daily by verifying or if two or more jet pump flow that the following conditions do not l
INSTRUMENT failures occur and cannot occur simultaneously:
l be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an orderly SHUTDOWN shall be initiated a.
The recirculation pump flow differs and the reactor shall be in a COLD by more than 15% from the esta-SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, blished speed flow characteristics.
b.
The indicated value of core flow rate varies from the value derived from loop flow measurements by more than 10%.
c.
The diffuser to lower plenum differ-ential pressure reading on an indi-vidual jet pump varies from the mean of all jet pump differential pres-sures by more than 10%.
F.
Egeirculation Pump Flow Mismatch F.
Recirculation Pumo Flow Mismatch 1.
Following one recirculation pump 1.
Deleted.
operation, the discharge valve of the low speed recirculation pump may not be opened unless the speed of the faster pump is equal to or less than 50% of its rated speed.
C.
Inservice Inspection G.
Inservice Inspection l
1.
To be considered OPERABLE, compo-nents shall satisfy the requirements 1.
Inservice inspection shall be per-l contained in Section XI of the ASME formed in accordance with the re-Boiler and Pressure Vessel Code and quirements for ASME Code Class 1, 2, applicable Addenda for continued and 3 components contained in Sec-service of ASME Code Class 1, 2, and tion XI of the ASME Boiler and Pres-3 components except where relief has sure Vessel Code and applicable been granted by the Commission pur-Addenda as required by 10 CFR 50, suant to 10 CFR 50, Sec-Section 50.55a(g),
except where i
tion 50.55a(g)(6)(i).
relief has been granted by the Com-mission pursuant to 10 CFR 50, Sec-tion 50.55a(g)(6)(i).
2.
The inservice inspection program for '
piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule, methods, personnel, and sample expansion included in this generic letter.
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l Amendment No. 161 137-
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f 3.6.C & 4.6.C BASES (cont'd.)
indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or
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stress corrosion cracking or some other mechanism characterized by gradual crack growth.
This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks, associated with such leakage would grow rapidly. However, the establishment t
of allowable unidentified leakage greater than that given in 3.6.C on the basis of i
the data presently available would be premature because of uncertainties associated with the data.
leakage limits of 2 gpm increase within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and a maximum of 5 gpm are specified in 3.6.C and are also supported Cener$c Letter 88-01.
The experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for i
l rapid propagation.
Leakage less than the magnitude specified can be detected
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reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time the plant should be SHUTDOWN to allow further investigation and corrective action.
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The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
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The ccpacity of the drywell floor sump pumps is 50 gpm and the capacity of the I
l drywell equipment sump pumps is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with margin.
Reactor coolant leakage is also sensed by the drywell air sampling system which detects gaseous, particulate, and iodine radioactivity. Leakage can also be detected by area temperature detectors, humidity detectors and pressure instrumentation. Due to the many and varied ways of detecting primary leakage, a 30 day allowable repair time is justified.
D.
Safety and Relief Valves The safety and relief valves are required to be OPERABLE above the pressure l
(113 psig) at which the core spray system is not designed to deliver full flow. The pressure relief. system for Cooper Nuclear Station has been sized to meet two design bases.
First, the total safety / relief valve capacity has been established to meet the overpressure protective criteria of the ASME code.
Second, the distribution of this required capacity between safety valves and relief valves has been set to meet design basis IV.4.2.1 of the USAR which states that the nuclear system relief valves l
shall prevent opening of the safety valves during normal plant isolations and load rej ections.
The details of the analysis which shows compliance with the ASME code requirements is presented in subsection IV.4 of the FSAR and the Reactor Vessel Overpressure Protection Summary Technical Report presented in question 4.20 of Amendment 11 to the FSAR. Results of the overpressure protection analysis are provided in the current reload license document.
Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
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Amendment No. 161
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