ML19329F848
| ML19329F848 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/06/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19329F849 | List: |
| References | |
| NUDOCS 8007110254 | |
| Download: ML19329F848 (5) | |
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UNITED STATES A f
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CONSUMERS POWER COMPANY DOCKET NO. 50-255 PALISADES PLANT AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 58 License No. OPR-20 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The appl 1:ation for amendment by Consumers Power Company (the licensee) dated May 14, 1980, 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; 8.
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 ca'n be conducted without endangering the health and safety of the public, and (ii) that such activities j.
will be conducted in compliance with the Comission's regulations; 4
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.
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Accordingly..the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amandment and paragraph 3.B of Provisional Operating License-
"c. OPR-20 is hereby amended to read as follows:
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- 3. ' Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 58, are hereby incorpcrated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h
N Dennis M. Crutchfield, Chief Operating Reactors Branch #5 l
Division of Licensing j
Attachment:
Changes to the Technical 1
Specifications
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Date of Issuance: June 6,1980
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ATTACHMEf;T TO LICEf1SE AMEf;DMEf!T 'i05 8 PROVISIOilAL OPERATIt!G LICEtiSE ?!O. DPR-20 00CKET l10. 50-255 Revise Appendix A Technical Specifications by removing the following
? ages and by inserting the enclosed pages.
The revised pages contain tha captioned amendment number and marginal lines indicating the area Of change.
PAGES 3-59 3-66 t
t 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Centd) 3.10.3 Power Distribution Limits-(Contd) satisfy the criterien. Appropriate consideration shall be given to the following factors:
(1) A flux peaking augmentation factor of 1.0, (2) A measurement calculational uncertainty factor of 1.10, (3) An engineering uncertainty f actor (which includes fuel eclumn shortening due to densification and thermal expansion) of 1.03, and (4) A thermal power measurement uncertainty factor of 1.02.
b.
If the quadrant to core average pcwar tilt exceeds 13'.,
except for physics tes:s, then:
(1) The linear heat generation rate shall prcmptly be demonstrated to be less than that specified in Par: a, or a
i (2) Immediate ac:icn shall be initiated to reduce reactor pcwer to 75%
or less of rated power.
l c.
If the power in a quadrant exceeds core average by 10*. for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if the power in a quadrant exceeds core average by 20'. at any time. immediate action shall be initiated to reduce reactor pcwer belew 50*. until the situation is remedied.
d.
If :he power in a quadrant exceeds the core average by 15'. and if the linear heat generation rate cannot be demonstrated promptly to be within limits, then the overpower trip set point shall be reduced to 30'. and the thermal margin low pressure trip set point (PTrip) shall be increased by 400 psi.
e.
If the power in a quadrant exceeds core average by 5% for a period of 30 days, immediate action shall be initiated to reduce reactor pcwer to 75*.
or less of rated power.
f.
The part-length centrol rods will be completely withdrawn frem the core (except for red exercises and physics tests).
A g.
The calculated value of F shall be limited to < 1.45 (1.0 + 0.5 T*
(1 - P)), the calculated value of F shall be limited to < 1.77 (1.0 +
S 0.5 (1 - P)), and the calcula:ed value of F shall be limited to
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< 1.66 (1.0 + 0.5 (1 - P)), where P is the core thermal power in fractica i
of core rated thermal power (2530 MWt).
(*Fer the duration of Cy:1e-4 for H-fuel only, F.
for rods adjacent to the wide water gap shall be limited to 1.90 (1.0 ' O.5 (1 - P)).)
J 3-59 Amendment NO. 31,.43, 57, 58 n
'3.11 IN-CORE INSTRCMENTATION (Cen:d)
Spscification (Contd) a 10-hour period)-at least each two hours thereafter or the reactor power level shall be reduced to less than 50*. of rated power (63*. of rated power if no dropped or misaligned rods are present).
If readings indicate a local power level equal to or greater than the alarm se:
point, the action specified in 3.11.b shall be taken.
g
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F ", F/ and F shall be determined whenever :he core pcwer distribution r
e T
e H is found to be in excess of the limit is evaluated.
If aither F A,Cr., er r specified in Section 3.10. (g), within one hour thermal power shall be reduced to less than'-
a
/1 )
(1 - 2(F " - 1)) X 2530 MWt 7
1.1-T (2)
(1 - 2(F
- 1)) X 2530 Mut
- or p
1.77*
AN (3)
(1 - 2(F
- 1)) X 2S30 F..t r
I 70 er is 1cwer.
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- r ; N um.
" -; 7 50"I!.
a Basis A system of 43 in-core flux detector and thermecouple assemblies and a data display, alarm and record functions has been provided. A four level, five level or six level system =ay be used.(1)( ) The out-of-core nuclear 1
instrumentation calibration includes:
1 Calibration (axial and azimuthal) of the split detectors at initial a.
reactor start-up and during the.pcwer escaletion program.
A comparison check with the in-core instrumentation in the event abnormal b.
readings are cbserved on the out-of-core detectors during opera:1cn.
c.
Calibration check during subsequent reactor star:-ups, d.~ Confirm that readings frem the out-of-core split detectors are as expected.
Core power distribution verification includes:
a.
Measurement at initial reactor s: art-up to check that power distributien is censistent with calculations.
b.
Subsequent checks during cperation to insure that power distribution is consistent with calculations.
i Indication of power distributica in the even: that abnormal situations i
c.
occur during reactor operation.
i If the data logger for the in-core readout is not in operation for more than two hcurs, power will be reduced to provide margin between the actual peak linear heat generatica ra:es and the limit and the in-core readings
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j will be manually collected a: the terminal blocks in he control recm utilizing a suitable signal detector.
If this is not feasible with the 4
Amendment No. 3T, 43, 50, 57, 58 3-66
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UNITED STATES i
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NUCLEAR REGULATORY COMMISSION
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% *....9 SAFETY EVALLATI0fl BY THE OFFICE OF flVCLEAR REACTOR REGULATION SUPPORTING AMENDMEllT NO. 58 TO LICE!!SE t!0. DPR-20
, CONSUMERS POWER COMPANY PALISADES PLANT DOCKET fl0. 50-255
1.0 INTRODUCTION
910 DISCUSSION By letter dated May 14,1980 (Reference 1) Consumer's Power Company (CPCo),
(the licensee) requested an amendment to Appendix A of the Provisional Goerating License No. DPR-20 for the Palisades Plant.
This is the third in a series of related requests, pertaining to the peaking factors of the Cycle-4 H-design loading.
CPCo was requested by letter from D. Zienann (NRC) to D. Bixel dated July 11,1979 (Reference 2), to submit information which would provide assurance that water hole peaking is appropriately considered in the calculation of flux distributions.
CPCo's replies dated September 10, 1979 and February 26,1980 (letters D. Hoffman CPCo to D. Ziemann NRC, References 3 and 4 respectively) dealt with the 4
calculational procedure used to compute water hole peaking. CPCo by letter dated February 26, 1980 submitted information succorting the addition of the " Total Interior Rod Ldial Peaking Factor F$H". The licensee considered it appropriate to impose a limit on the product of total radial peaking factor times the interior pin local peaking facto.r to assure that the assumptions in the DNS analysis remain valid in all cases.
This proposed addition has been reviewed and accepted by the NRC staff (Reference 5).
The current request (Reference 1) concerns a change of the Palisades Plant Technical Specifications to increase the limit of the Total Radial Peaking Factor FrT for Type H fuel assembly rods adjacent to the wide water gap from 1.77 (1.0 + 0.5 (1-P)) to 1.90 (1.0 + 0.5 (1-P))
where P is the core thermal power in fraction of core rated thermal power (2530 Mwt). This increase is only for the Cycle 4 loading and will allow operation at full power for the total fuel cycle, whereas operation under the present Technical Specifications will result in plant operation derated by 12% power for part of this cycle.
2.0 _ EVALUATION d
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A c see o J e 1
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- f. The 1icensee responded IN DUPLICATE DOCUMENT
.,3 6 and 9 ). The followina Entire document previously entered into system under:
ANO Oh d
No. of pages:
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