ML17037C421

Amendment to Operating License/Change to Appendix a Technical Specification Relates to Partial Loop Operation on the Emergency Core Cooling System with Supporting Info

ML17037C421
Person / Time
Site:	Nine Mile Point
Issue date:	04/14/1977
From:	Eric Thomas LeBoeuf, Lamb, Leiby & MacRae, Niagara Mohawk Power Corp
To:	Rusche B Office of Nuclear Reactor Regulation
References
Download: ML17037C421 (24)

Text

NRC'FDRM 196, . - . '., U.s, NUcLEAR REGULAToRv coMMIssIoN DOCKET NUMB'<

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'NUMBER DISTRIBU i LORY'FDR P T 50 DOCKET MATERIAL '~NRC TP o FROM: DATE OF DOCUMENT.

LeBoeuf, Lamb, Leiby & MacQ,ae 4/14/77 Mr. Ben C. Rusche Washington, Di C DATE RECEIVED Eugene Bi Thomas, Jr 4/14/77

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o trans the fol lowing: l Amdt~ to OL/change to Appendix'A tech-specsi ~ inotorized 4/12/77 ~ ~ irelates to partial loop operation. on the Emergency

.ACKJXlow i"DGFD Core Cooling. System with supporting infoi ~ ~ .

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PLANT NAME:

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-SAFETY. FOR ACTION/INFORMATION ASSIGNED AD:

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0 LAW OFFICES OF LEBOEUFI LAIYIB>LEIBY 84 MACRAE I757 N STREETf N.W.

WAsHINGToN, D. C. 20036 TELEPHONE 202 h57 7500 CABLE AOORCS5 LEON A. Al LEN, JR. CAMERON F. MACRAC 4 RANDALL J. LCBOEUP,JR. IQ29 l975 JOSEPH C.GACHELDER,TIE CAMCRON F. MAcRAC III 4 LEBWIN,WASHINCTONiD.c.

CRNCST S. GALLARO,JR. GERARD A.MAHCR TELCX: 440274 AORIAN C. LCIBY I952 ~ I975 O. S. PCTCR BERGEN 4 SHEILA H. MARSHALL GCOFFRY D. C. GEST JAMCS G.McCLROY DAVID P DICKS JAMES P. McQRANCRY, JR. CP CCVHSCL TAYLOR R. GRIGGS PHILIP PALMER McGUIGAN CHARLES N BURGER L. MANNINQ MUNTZING~4 ARVIN K. UPTON WILLIAM O. DOUGi 4 JAMES O'ALLEY.JR.4 CUGCNC R. FIDELL44 JACOB FRIEOLANOER J. MICHAEL PARISH JOHN A. RUDY PAUL G. RUSSELI.

April 14, 1977 DONALD J. QRCCNC HAROLD M. SCIDCI. Iho BROADWAY JAMES A QRCERoIE 4 CHARLES P. SIFTON JOHN L.GROSC 4 HALCYON Q. SKINNER NEW YORK, N.Y. IOOOS DOUGLAS W. HAWKS JOSEPH S. STRAUSS CARL D,HOGCLMAN SAMUEL M.SUGDCN TCLCPHONC 2I2 259 IIOO MICHAEL IOVCNKO EUGENE G. THOMAS, JR 4 4 CABLE ADDRESS JAMES F JOHNSON LEONARD M. TROSTEN 4 RONALD O.JONES HARRY H VOIQT 4

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i LEBWIN, NEW YORK LEX K. LARSON 4 H. RICHARD WACHTCL GRANT 5 LEWIS

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U. S. Nuclear Regulatory Commission s&~"

Nashington, D. C. 20555 Re: Ni,agara Mohawk Power Corporation Nine Mile Point Nuclear Station

'ni;t No. 1--Dock'et No." 5'0-220

Dear Mr. Rusche:

As counsel for Niagara Moha'wk,' hereby enclose the following:

('1) Three (3) originals and nineteen (19) copies of a, pleading requesting a change in the above-numbered facili,ty's opexating license DPR-63; (2) Forty ('.40) copies of the proposed Technical Specificati,ons filed herewith as Attachment A; and Forty (40) copies of the suppoxting infox-

'atfon .(.:3)as set forth i.n Attachment. B.

The proposed Technical Specification change relates to partial loop operation on the Emergency Core Cooling System.

Very truly yours, LEBOEUP, LAMB, LEXBY E MACRAE Attorneys for Niagara Mohawk Power Corporation Eugene B. Thomas, Jr.

Partner Enclosure

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

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NIAGARA MOHAWK 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, Niagara Mohawk Power Cor-H poration, holder of Facility Operating License No. DPR-63, hereby requests that Specification 3.1.7 and 4.1.7 of the Technical Specifications and Bases set forth in Appendix A to that License be amended. These proposed changes have been concurred with by the Site Operations Review Committee and Safety Review and Audit Board.

The proposed Technical. Specification changes are set forth in Attachment A to this application. Supporting Information, which demonstrates that the proposed changes do not involve a significant hazards consideration, is set forth in Attachment B. The, proposed change would not authorize 'any change in the types ox any increase in the amounts of effluents or any change in the authorized power level of the facility.

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

NIAGARA MOHAWK POWER CORPORATION Gerald K. ode Vice President-Engineering Subscribed and sworn to before me 'on this Pz~

day of April, 1977.

NO ARY PUBLIC HAZEL J. CARRICK Notary Public in the State of New Yotk Qualified in Onon. Co. No. 4524460 51y Commissiun bpires March 30,19'7k

Attachment A Corporation,

'iagara Mohawk Power License No. DPR-63

. Docket No. 50-220 Proposed Changes To Faci1ity Operating License The attached-pages 63, 70 and 70c are revisions to Appendix A of DPR-63.

I LIHITING CONDITIONS FOR OPERATION SUR>! EILLANCE RE(UIREYiENT 3.1:7 FUEL RODS 4.1.7 FUEL RODS A~ii bi1it A licabilit  :

The Limiting Conditions for Operation associated The Surveillance Requirements apply to the with the fuel rods apply to those parameters parameters which monitor the fuel rod operating which monitor the fuel rod operating conditions. conditions.

~0b'ective: ~0b'ective:

)The objective of the Surveillance Requirements The objective of the Limiting Conditions for Operation is to assure the performance of the fuel rods.

is to specify the type and frequency of surveillance to be applied to the fuel rods.

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~5if i ti ~RiH ti Avera e Planar Linear Heat Generation Rate a. Avera e Planar Linear Heat Generation Rate

a. APLHGR APLHGR During power operation, the APLHGR for each The APLHGR for each type of fuel as a a function of average planar function of average planar exposure shall type of fuel as be determined daily during reactor operation exposure shall not exceed the limiting value shown in Figures 3.1.7.a, 3.1.7.b, 3.1.7.c, at > 25$ rated thermal power.

3.1.7.d and 3.1.7.e. If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the APLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

During power operation with one recirculation line isolated, the APLHGR for each fuel type as a function of average planar exposure shall not exceed 98K of limiting value shown in Figures 3.1.7.a, 3.1.7.b, 3.1.7.c, 3.1.7.d and 63 in A

,n BASES FOR 3.1.7 and 4.1.7 FUEL RODS Avera e Planar Linear Heat Generation Rate APLHGR)

This specification assures tha't the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10CFR50, Appendix K.

T he peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than + 20 F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10CFR50, Appendix K limit. The limiting value for APLHGR is shown in Figure 3.1.7.

These curves are based on calculations usina ~he models described in References 1, 2, 3, 5 5 6.

Analysis has been performed (Reference 7) <which shows for isolation of 1 loop, operation limited to 98Ã of the limiting APLHGR shown in Figure 3.1.7 conservatively assures compliance with 10CFR50, Appendix K.

Linear Heat Generation Rate LHGR)

This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation even if fuel pellet densification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Reference 1'and in References- 2, .3 and 4, and assumes a linearly increasing variation in axial gaps between core bottom and top, and assures'ith a 95Ã confidence, that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking. The LHGR as a function of core height shall be checked daily during reactor operation at > 25K power to determine if fuel burnup or control rod movement has caused changes in power distribution.

Minimum Critical Power Ratio MCPR At core thermal power levels less than or equal 'to 25K, the reactor'ill be operating at a minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal-hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial startup testing 70

REFERENCES FOR BASES 3;.'1.7 and 4.1.7 FUEL RODS (1) "Fuel Densification Effects on General Electric Boilin'g Water Reactor Fuel," Supplements 6, 7 and 8, NEDM-10735, August 1973.

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

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

l (4) "General Electric Boiling to Revision 1, Mater Reactor Generic Reload Application for 8 x 8 Fuel," RE00-20360, .Supplement December 1974. ~

(5) "General Electric Company Analytical Model for Loss of Coolant Analysis in Accordance with 10CFR50 Appendix K,"

NE00-20566.

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

(7) September 26, 1975 letter, G. K. Rhode, Niagara Mohawk Power Corporation to G. Lear, United States Nuclear Regulatory Commission.

70c

d Attachment B Niagara Mohawk Power Corporation License No. DPR-63 Docket No. 50-220 Supporting Information 4

In responding to the Nuclear'Regulatory Commission (NRC) letter of June 18, 1975, Niagara Mohawk addressed the effect of partial loop operation on Emergency Core Cooling System (ECCS) analyses assuming the idle loop is not isolated and that recirculation flow was adjusted to be equivalent to five loop flow. The conclusion was that no safety analysis was affected by partial loop operation (four recirculation pumps in service with the inoperative loop not isolated).

This.,document discusses the effect on safety analyses of partial loop operation with the idle loop isolated. The models used in this analysis are described in References 1 and 2.

Core flow is unaffected by isolation of the idle loop, thus the bases for the conclusions of the previous partial loop analysis (Reference

3) concerning the pump seizure accident and the transient analysis are applicable with the idle loop isolated. The total vessel inventory is affected by isolation of the idle loop, thus an. analysis of the isolation of an idle recirculation line was performed to determine the effect upon the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR).

The blowdown analysis was performed with one recirculation pump inoperative and the mass between the recirculation suction and discharge isolation valves subtracted from thy total vessel inventory. The two worst small breaks (0.07 and 0.5 ft~) and the Design Basis Accident (Reference 4) were considered.. Recirculation flow was assumed to be at rated'onditions with the four operating pumps appropriately compensated. The mass removed due to the isolated loop is small compared to the total vessel coolant inventory, and therefore, the hot node uncovered only seconds earlier than the case with all recirculation loops operating. The core heat-up analysis was. calculated assuming this earlier uncovery time and results indicate an increase in the peak cladding 'temperature of 30 F, which is approximately equivalent to a 1.5 percent reduction in the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR).

0 To assure that the peak clad temperature will remain below the 2200 F 'Appendix K 'limit during steady state power operation with one recir'culation line isolated, the average linear heat generation rate (LHGR) of: all the rods in any fuel assembly, as a function of average planar exposure, at any axial location shall be limited to 98%%d of the maximum average planar LHGR applicable to 5 pump operation.

REFEREi'ICES

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, NED-20566 (Draft), submitted

. August 1974, and General Electric Refill Ref lood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L. Gyorey to Victor Stello, Jr., dated December 20, 1974.

2. GEGAP-III: A tlodel for the Prediction of Pellet-Claddin Thermal Conductance in BMR Fuel Rods, NED0-20181, November 1973.

3. Letter, G. K. Rhode, Niagara Mohawk Power Corporation to G. Lear, United States Nuclear Regulatory Commission dated September. 26, 1975.

"4. Letter, G. K. Rhode, Niagara Hohawk Power Corporation to G. Lear, United States Nuclear Regulatory Commission dated October 31, 1975.

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