ML17053A756
| ML17053A756 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/28/1979 |
| From: | Eric Thomas LEBOEUF, LAMB, LEIBY & MACRAE |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17053A757 | List: |
| References | |
| NUDOCS 7906290584 | |
| Download: ML17053A756 (24) | |
Text
REGULATORY INFnRMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:7906290580 Db,DATE: 79/06/28 NOTARIZED.
NO DOCKET FACIL:50-220 Nine Mile Point Nuclear Stationi Unit 1~ Niagara Powe 05000220 AUTH,NAME, AUTHOR AFFILIATION THOMASiE.B, LeBoeuf~
Lamb~ Leiby L MacRae REC IP. NAME RECIPIENT AFFILIATION DENTONgH RE Offic~ of Nuclear Reactor Regulation
SUBJECT:
Forwards application to amend License DPR-63.
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LEBOEUF, LAMB7LEIBY 8o MACRAE l333 NEW HAMPSHIRE AVENUE,N. W.
WAsHINGTONID. C. 20036 TELESHOHC ROR 457 ~7500 CASLE ADDRESS LEON A. ALLEN'R.
JOSEPH K,BACHELDER>3IE ERNEST S. BALLARD,JR.
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GEOFFRY D. C. BEST DAVID P. BICKS TAYLOR R. BRIGGS CHARLES N.BURGER THOMAS E BURKE JOHN B. CHASE ROGER O. FELDMAN a
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JACOB FRIEDLANDER GERARD GIORDANO DONALD J ~ GREENE JAMEs A ~ GRKKR.xc a JOHN L.GROSE a
DOUGLAS W.HAWKS CARI. D ~ HOBKLMAN MICHAEL IOVENKO JAMES F'JOHNSON
~ 4' RONALD D.JONES JAMES A. LAPKNN LEX K ~ LARSON 4
ORANT S LEWIS KIMBAW LOVEJOY CAMERON F, MACRAK S
CAMERON F,MAORAE,IK s GERARD A.MAHER SHEILA H. MARSHALL JAMES G. MOELROY JAMES P.MOGRANERY,JR+ a PHILIP PALMER MOGVIGAN E. ELLSWORTH MOMKKN,XX WILLIAMD MORRISON HARVEY A ~ NAPIER JAMEs 0 MALLEY,JR. a J. MICHAEL PARISH JOHN C. RICHARDSON a
WILLIAMW ROSKNBI.ATT JOHN A RVDY PATRICK J
~ SCOGNAMIGLIO HAROLD M. 5 E IDE L HALCYON O. SKINNER JOSEPH S STRAUSS SAMUEL M SVGDEN EUGENE B THoMAs JR.
4 LEONARD M ~ TROSTEN a a HARRY H VOIGT a
H. RICHARD WACHTEI.
GKRARD P.WATSON THOMAS A. ZIERK LESWIN,WASHIHOTOHsD.C.
TELEX: 440274
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~ RESIDENT PARTNERS WASHINGTON OFFICE RESIDENT PARTNERS LONDON OFFICE s ADMITTED TO THE DISTRICT OF COLUMBIA BAR June 28, 1979 Mr. Harold R. Denton Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Re:
Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station, Unit No.'1 Docket No'.0'-'220"
Dear Mr. Denton:
As counsel for Niagara Mohawk Power Corporation, X enclose the following:
I (1) Three (3) originals and nineteen (19) copies of an Application for Amendment to Operating License to amend Sections 2.1.2 and 3.1.7 and Bases of the Techinical Specifications to utilize, an extended power/flow line which will allow rated power operation at, reduced core flow; and (2) Forty (40) copies each of three (3) documents designated Attachments A, B, and C which set forth the re-quested changes in the Technical Specifications along with its technical basis, and supporting information, which demonstrates that the proposed changes do not involve T906890 5'd'/
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Mr. Harold R. Denton June 28, 1979 Page Two significant hazards consideration, nor would authorize any change in the types or any increase in the amount of effluents or any change in the authorized power level of the facility.
The proposed amendment to the Operating License has been evaluated and determined to fall within the defini-tion of Class III of 10 C.F.R. 5 170.22; therefore, a check in the amount of
$4,000.00 is enclosed to cover the appropriate fee.
Enclosures by Very truly yours,
- LeBOEUF, LAMB, LEIBY 6 MacRAE Eugene, B. Thomas, Jr.
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UNITED..STATES OF, AMERICA NUCLEAR REGULATORY COMMISSION In. the.Matter of
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'NIAGARA. MOHAWK POWER CORPORATION
)
(Nine Mile,Point, Nuclear Station
).
Unit No.
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Docket.No.
50-220 APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the Nuclear Regulatory Commission, Niagara 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 demonstrates 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 V906890 SQO
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amendment pursuant to 10 C.F.R.Section. 170.22 is included as Attachment C.
A check for the appropriate 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 PGWER'ORPORATION By Donald P. Disc Vice President.
Engineering Subscribed and sworn to before me on this 47+day of June, 1979.
NOTARY PUBL C PHYLLIS D. VOYTKO Notary Public In the Stete rtf Now YorR'vollltott In Onon. Co. No. 84;948653@g My Cotnrnlt>ton Rxnlrar March 30, 19~
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ATTACHNENT A Niagara Mohawk Power Corporation License No.
DPR-63 Docket No. 50-220 Pro osed Chan es To The Technical S ecifications A
endix 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 with changes shown.
BASES FOR 2.1.2 FUEL CLADDING - LS3 chambers provid'e the,basi'c input signal's, the APRN system responds directly to average neutron fl'ux.
During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) i's 1'ess than the instantaneous neutron flux due to the time constant of the fue1
'. Therefore, during abnormal'perationa1 transients, the thermal poner of t'p~ fuNI uI 11 be 1'ess than that indicated by the neutron flux at the scram setting.
Analyses demonstrate that with a T20/ scram trip setting, none of the abnormal operational transients-anal'yzed violate the fuel safety limit and there is a substantial margin from fuel damage.
However, in response to expressed beliefs that variation of APRtl flux scram with recircula-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.
An increase in the APRH scram trip setting would decrease the margin present before the fuel cladding integrity safety limit is reached.
The APRN scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.
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 APRH scram trip setting was selected because it provides adequate margin for the fuel cladding in-tegrity safety limit yet allows operating margin that reduces the possibility of unnecessary scrams.
The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of HTPF 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 greater than the limiting total peaking factor.
b.
flormal 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 IRN scram setting.is 12% of rated neutron flux.
Although the operator will set the IRt1 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 be. ween the setpoint and the safety limit at 2SX of.rated power.
The margin is adequate to accomm'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 condi iion.
Effects of increasing pressure at zero or low
100 Nine Mile Point Unit 1
a CP CX O
80 Limiting Power/Flow Line CY UJ CD 0
60 LJ CD 40 20 0
40 60 Percent Rated Core Flow 80 100 Figure 3.1.7.aa LIMITING POii<ER FLOM LINE 64c
REFERENCES FOR BASES 2.1.1 AHD 2.1.2 FUEL CLADDING (1)
General Electric DWR Thermal Analysis Dasis (GETAD) Data, Correlation and Design Application, HEDO-10958 and HEDE-10950.
(2)
Linford, R. 8., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor,"
HE00-10001, February 1973.
{3)
FSAR, Volume II, Appendix E.
(4)
FSAR, Second Suppl ment,.
(5)
FSAR, Volume II, Appendix E.
(6)
FSAR, Second Supplement.
(7)
Letters, Peter A. Morris, Director of Reactor Licensing, USAEC, to John E. Logan, Vice-President, Jersey Central Pow r and Light Company, dated November 22, 1967 and January 9, 1968.
{8)
Tcchnical Supplement to Petition to Increase Power Level, dated April 1970.
{9)
Letter, T. J.
- Drosnan, Niagara Mohawk Power Corporation, to Peter A. Morris, Division of Reactor Licensing, USAEC, dated February 28, 1972.
{10) Letter, Philip D. Raymond, Niagara Mohawk Power Corporation, to A. Giambusso, USAEC, dated October 15, 1973.
{11) Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis, HEDO 24012, May, 1977.
{12). Licensing Topical Report General Electric Boiling Water. Reactor Generic Reload Fuel Application, HEDE-24011-P-A, August, 1978.
(13) 'line Nile-Point Nuclear Power Station Unit 1; Extended Load Line Limit.Analysis, License Amendment Submittal (Cycle 6),
HE00-24185., April 1979.
Amendment Ho. 5, gg, 31 20
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BASES FOR 3.T.7.Ni 9 4.7.7 FUEL RODS of the plant, a
HCPR evaluation will be made at the 25/ thermal power level with minimum recirculation pump
- speed.
The HCPR margin will thus be demonstrated such that future NCPR evaluations below this power level will be shown to be unnecessary.
The daily requirement for calculating MCPR above 25/ rated thermal. power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that HCPR will be known following a change in power or pointer shape (regard'less of magnitude) that could place operation at a thermal limit.
0 Figure 3.1.7-1 is used for calculating MCPR during operation at other than rated conditions.
For the case of automatic flow control, the Kf factor is determined such that any automatic increase in power (due to flow control) will always result in arriving at'the nominal required HCPR at 100/. power.
For manual flow control, the Kf is determined such that an inadvertent increase in core flow (i.e., operator error or recirculation pump speed controller failure) would result in arriving at the 99.9% limit MiCPR when core flow reaches the maximum possible core flow corresponding to a particular setting of the recirculation pump HG set scoop tube maximum speed control limiting set screws.
These screws are to be calibrated arid 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 HCPR associated with an inadvertent flow increase always satisfies the 99.9Ã requirement.
Irrespective of the scoop tube setting, the required HCPR is never allowed to be less than the nominal MCPR (i.e,, Kf is never less than unity).
Power/Flow Relationshi The power/flow curve is the locus of critical power as a fUnction of flow from which. the occurrence of abnormal operating transients will yield results within defined plant safety limits.
Each transient and postulated accident applicable to operation of the plant.was analyzed along the pokier/flow line.
The analysis (7i8i9) justifies the operating envelope bounded by the power/flow curve as long as other operating limits are satisfied.
Operation under the power/flow ling'is designed to enable the direct ascension to full power within the design basis for the plant.
Reactor power'evel in the one-loop-isolated mode is restricted to a power level which has been analyzed and found acceptable.
70a
REFERENCES FOR BASES 3. 1.7 AND 4.1.7 FUEL RODS (1)
"Fuol Densification Effects on General Electric Boiling Mater Reactor Fuel," Supplements 6,
7 and 8, HEDH-10735, August 1973.
(2)
Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (USA R gula'ory Staff).
{3)
Co-,;.unication:
V. h. Hoore to I. S. Hitchell, "Hodified GE tiodel for Fuel Densification," Docket 50-321, Harch 27, 1974.
(4)
"General Electric Boiling Mater Reactor Generic Reload Application for 8 x 8 Fuel," HED0-20360, Supplement 1 to Revision 1, December 19/4.
(5)
"Goneral Electric Company Analytical Hodel for Loss of Coolant Analysis in Accordance with 10CFR50 Appendix i<,"
HE00-20566.
(6)
General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by
- letter, G. L. Gyorey t:o Yictor Stello Jr., dated December 20, 1974.
{7)
"Nine labile Point thrclear Power Station Unit 1, Load Line Limit Analysis," HE00-24012.
(8)
Licensing Topical Report General Electric Boiling Mater Reactor Generic Reload Fuel Application, HEOE-240ll-P-A, August, 1978.
(9)
Nine Nile Point Nuclear Power Station Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal (Cycle 6), NED0-24185, April 1979.
Amendment No. gg, 31 70c
ATTACHMENT 8 Niagara Mohawk Power Cor'poration fl License No.
DPR-63 Docket No. 50-220 Su ortin 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 "Nine Mile Point Nuclear Power Station Unit 1 Extended I oad Line Limit Analysis Licensing Amendment Submittal, NED0-24185."
The analyses were performed utilizing an extended power/flow line.
,The analysis results are provided in the same sequence as the standard reload format to assure that all aspects of reactor 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 report shows operation near end-of-cycle requires slightly lower power levels to assure adequate 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.
I