ML19269E762

Requests Amend to License DPR-63 Re Tech Specs for Fuel Cladding & Fuel Rod.Forwards Proposed Tech Specs Changes & NEDO-24185

ML19269E762
Person / Time
Site:	Nine Mile Point
Issue date:	06/27/1979
From:	Dise D NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17053A757	List: ML17053A756 ML17053A758 ML19269E762
References
NUDOCS 7906290590
Download: ML19269E762 (9)

Text

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UNITED STATES OF AMERICA NtCLEAR REGULATORY COMMISSION In the Matter of )

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NI AGARA MOllAWl' POWER CORPORATION ) Docket No. 50-220 (Nine Mile Point Nuclear Station )

Unit No. 1) )

APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the Nuclear Regulatory Commission, hiagara Mohawk Power Corporation, holder of Facility Operating License No. DPR-63, hereby requests that Sections 2.1.2, 3.1.7, and Bases of the Technical Specifi-cations set forth in Appendix A to that License be amended.

This proposed change has been reviewed by the Site Operations Review Committee and the Safety Review and Audit Board.

The proposed Technical Specifications change is set forth in Attachment A to this application. Supporting information, which deronstrates that the proposed change does not involve a significant hazards consideration, is set forth in Attachment B. The proposed change would not authorize any change in the types or any increase in the amounts of effluents or any change in the authorized power level of the facility. Justification for classification of the 2142 310 7006290 670 .

. amendment pursuant to 10 C.F.R Section 170.22 is included as Attachment C. A check for the app Jopriate fee accompanies this application.

WHEREFORE, Applicant respectfully requests that Appendix A to Facility Operating License No. DPR-63 be amended in the form attached hereto as Attachment A.

NIAGARA MOHAWK POWER CORPORATION By Dud l.

Donald P. Dise Vice President - Engineering Subscribed and sworn to before me on this i? day of June, 1979.

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NOTARY PUBLIC 1

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ATTACHMEllT A liiagara Mohawk Power Corporation License 110. DPR-63 Docket tio. 50-220 Proposed Changes To The Technical Specifications (Appendix A)

Replace Pages 15, 20, 64c, 70a and 70c with the attached revised pages.

Pages 15 and 70a have been retyped in their entirety. Figure 3.1.7aa on Page 64c has been redrawn '1ith changes shown.

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3 BASES FOR 2~1.2

. FUEL CLADDING - LS ,

chambers provide the basic input signaTs, the AFFR system responds directly to average neutron -

T flux. During transients, the instantaneous rate of heat transfer frem the fuel (reactor D thermal power) is less than the instantaneous neutron flux due to the time constant of the O Therefore, durina abnormal operaticnai transients, tr.e thermal pcuer of th(e fuel w11 be less than that indicated by the neutrcn flux at the scram setting. Analyses o,6,8,9,0,11,13) fuel.

demonstrate that with a 120% scram trip setting, none of tne abnormal operational transients O analyted violate the fuel safety limit and there is a substantial nargin from fuel damage. -

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$ H: waver, in response to expressed beliefs that variation of APRM flux scram with' recircula- "

g* - tion flow is a prudent measure to assure safe plant operation during the design confirmation '

phase of plant operation, the scram setting will be varied with recirculation flow.

M P An increase in the APRM scram trip setting would decrease the margin present before the fuel

• ciadding integrity safety limit is reached. The APRM scram trip setting was dettrmined by an

- analysis of margins required to provide a reasonable range for maneuvering during cperation.

Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setti .g was selected because it provides adequate margin for the fuel cladding in-r0 tegrity safety limit yet allows operating margin that reduces the possibility of unnecessary

-n scrams.

The rcram trip setting must be adjusted to ensure that the LHGR transient peak is not increased u for any combination of MTPF and reactor core thermal power. The scram setting is adjusted in

- accordance with the formula in Figure 2.1.1 when the maximum total peaking factor is u ,

greater than the limiting total peaking factor,

b. I!ornal operation of the automatic recirculation pump control will be.in excess of 30% rated .

flow; therefore, little operation below 30% flow is anticipated. For operation in the start- .

up mode while the reactor is at low pressure, the IRM scram setting is 12% of rated neutron flux. Although the operator will set the IRM scram trip at 12% of rated neutron flux or less, the actual scram setting can be as much as 2.5% of rated neutron flux greater. This includes the margins discussed above. This provides adequate margin between the setpoint and the safety limit at 25% of rated power. The margin is adequate to accomT,odate anticipated maneuvers as-sociated with power plant startup. There are a few possible sources of rapid reactivity input to the system in the low power flow condition. Effects of increasing pressure at zero or icw 15

I

??ine Mile Point Unit 1 100 -

- 80 -

-g Limiting Power / Flow Line 8

5 5

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60 N

b

=

U O 40 N

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1. ' 20 -

4 l i l i l 1 0 20 40 60 80 100 .

Percent Rated Core Flow Figure 3.1.7.aa LIMITIr1G POWER FLOW LIllE 64c

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REFEREriCES FOR DASES 2.1.1 A!iD 2.1.2 FUEL CLADDIt?G -

(1) General Electric DUR Thermai Analysis Dasis (CETAB) Data, Correlation and Design Application, liEDD-10958 and T LIEGE-10958. C

" Analytical Methods of Plant Transient Evaluations for the General Electric Boilin9 Mater C

(2) Linford, R. B.,

neactor," IIED0-10801, February 1973.

y FSAR, Volume II, Appendix E.

O (3) y (4) FSAR, Second Supple:r.cnt. .. !CI3

==P (5) FSt.R, Volume II, Appendix E. g (6) FSAR, Second Supplement.

(7) Letters, Peter A. !! orris, Director of Reactor Licensing, USAEC, to John E. Lo9an, Vice-President, Jersey Central Pouer and Light Company, dated flovember 22, 1967 and January 9, 1968.

(8) Technical Supplement to Petition to Increase Power Level, dated April 1970.

3(9) Letter, T. J. Brosnan, fliagara Mohawk Power Corporation, to Peter A. Morris, Division of Reactor Licensin9, .

_p,. USAEC, dated February 28, 1972.

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{l0) Letter, Philip D. Raymond,fliagara Mohawk Power Corporation, to A. Giambusso, USAEC, dated October 15, 1973.

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]I1) lline 11ile Point tioclear Power Station Unit 1 Load Line I %it Analysis, flE00 24012, May,1977.

(12). Licensing Topical Report General Electric Boiling Water Reactor Generic Reload Fuel Application, tiEDE-24011-P-A, August, 1978.

(13) Nine :lile Point !?uclear Power Station Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal (Cycle 6), fiEDO-24185.. April 1979.

s M.endT.ent ilo. E, 23, 31 20

BASES FOR 3.7.7.AND 4.T.7 FUEL ROJS of the plant, a MCPR evaluation will be made at the 25% thermal power level with minimum recirculation pump T.

speed. The MCPR nargin will thus be demonstrated such that future MCPR evaluations below this power level C will be shovin to tc unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power cistribution shifts are very s'ow when there have not been significant power or C control red changes. The requirement for calculating MCPR when a limiting control rod pattern is approached Z ensures that MC?R will be known following a change in power or power shape (regardless of magnitude) that C could place operation at a thermal limit. ,

Z Figure 3.1.7-1 is used for calculating MCPR during operation at other than rated conditions. For the case c. g of automatic flow control, the K7 factor is determined such that any automatic increase in power (due to flow-2 control) will always result in arriving at the nominal required MCPR at 1007, power. For manual flow control, the Kr is determined such that an inadvertent increase in core flow (i.e., operator error or D

I recirculatica pump speed controller failure) would result in arriving at the 99.9% limit MCPR when core flow reaches the maximum possible core flow corresponding to a particular setting of the recirculaticn pump M3 set scoop tube maximum speed control limiting set screws. These screws are to be calibrated arld set to a particular value and whenever the plant is operating in manual flow control the Kf defined by that setting of the screws is to be used in the determination of required MCPR. This will assure that the reduction in

- MCPR associated with an inadvertent flow increase always satisfies the 99.9% requirement. Irrespective of the scoop tube setting, the required MCPR is never allowed to be less than the ncminal MCPR (i.e., F ' is never less than unity). .

Power / Flow Relationship The poveer/ fica curve is the locus of critical power as a function of flow frca which.the occurrence of abncrcal operating transients will yield results within defined plant safety limits. Each transient The and postulated accident applicable to operation of the plant was analyzed along the power / flow line.

1 analysis (7,8,9) justifies the operating envelope bounded by the power / flow curve as long as other operating

. limits are satisfied. Operation under the power / flow line is designed to enable the direct ascension to N full power within the design basis for the plant. .

u Reactor povier level in the one-locp-isolated mode is restricted to a power level which has been analyzed and

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found acceptable.

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REFERENCES FOR BASES 3.1.7 AND 4.1.7 FUEL RODS (1) " Fuel Censification Effects on General Electric Goiling Water Reactor Fuel," Supplements 6, 7 and 8, NECM-!0735, August 1973.

(2) Supplecent 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff).

(3) Com.vunica tion: V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.

(4) "Ceneral Electric Doiling Water Reactor Generic Reload Application for 8 x 8 Fuel," NED0-20360, Supplement 1 to 1:evision 1, Occember 1974.

(5) " General Electric Company Analytical Model for loss of Coolant Analysis in Accordance with 10CFRSO Appendix K,"

UECO-20SGS.

(6) General Electric Refill Reficed Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L. Gyorey to Victor StcIlo Jr., dated December 20, 1974.

(7) "nine Mile point Unclear power Station Unit 1, Load Line Limit Analysis," UE00-24012.

(8) Licensing Topical Report General Electric Boiling Hater Reactor Generic Reload Fuel Application, NEDE-240ll-p-A, August, 1978.

J (9) Nine Mile Point Nuclear Power Statien Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal -

(Cycle 6), NEDO-241CS, April 1979.

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-a Amendment No. ZA, 31 70c

ATTACHMENT B tiiagara Mohawk Power Corporation License flo. DPR-63 Docket flo. 50-220 Supporting Information Attachment A describes proposed changes to the Nine Mile Point Unit 1 Technical Specifications. These changes are required to provide for more flexible operation. The bases for the proposed Technical Specification changes are provided in the enclosed report "fline Mile Point fluclear Power Station Unit 1 Extended Load Line Limit Analysis Licensing Amendment Submittal, HEDO-24185."

The analyses were performed utilizing an extended pot;er/ flow line. The analysis results are provided in the same scquence as the standard reload format to assure that all aspects of reactar operation which are affected are considered. Future reload submittals will incorporate the use of the extended load line in the analysis.

The analysis in the attached ceport shows operatio^ near end-of-cycle requires s'.ightly lower power levels to assure a quate pressure margins during a postulated turbine trip without bypass transient than those calculated in Reference 2. This is due to the reduced core flow assumed in the calculation. To assure adequate pressure margins near end-of-cycle, Niagara Mohawk will utilize a power / flow ratio near end-of-cycle which is conservatively bounded by the analysis in the attached report and that of Reference 2.

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