ML17037C414
| ML17037C414 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 07/18/1977 |
| From: | Eric Thomas LeBoeuf, Lamb, Leiby & MacRae |
| To: | Case E Office of Nuclear Reactor Regulation |
| References | |
| Download: ML17037C414 (30) | |
Text
U.S. NUCLEAR REGULATORY COM ION NRC FORM 196
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NRC DISTRIBUTION FoR PART 50 DOCKET MATERIAL DOCKET NUMBER 8'0-2Z 0 FILE NUMBER kRETTE R ORIGINAL QCOPY DESCRIPTION r&NOTOR IZED PION C LASS IF IE0 TO!~
Mr. Edson G. Case PROP INPUT FORM ENCLOSURE FROM:
- LeBoeuf, Lamb, Leiby & MacRae Washington, D
C Eugene Bi Thomas, Jri DATE OF DOCUMENT'/18/77 DATE RECEIVED 7/19/77 NUMBER OF COPIES RECEIVED Sl 4 slKO
@Op EPQQ License No+
DPR-63 'Appl for Amend: tech specs proposed change concerning Load Line Limit Analysis
~ ~ notorized 7/ll/77 ~ ~ ~ ~ ~ ~ ~
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PLANT NAME.
Nine Mile Point Unit No 1
RJL 7/19/77 ECT MANAGER:
FOR ACTION/INFORMATION ASSIGNED AD!
BRANCH CHIEF:
PROJECT MANAGER:
ENVIRHNMENTAL V+ MOORE LTR CEN ING ASSISTANT!
LICENSING ASSISTANT:
INTERNALD S
SAFETY HEINEMAN EDE ENGINEERING Bo HARLESS ISTRIBUTION PLANT SYSTEMS TEDESCO BENAROYA IPPOLITO OPERATING REACTORS SITE SAFETY &
ENVIRON ANALYSIS DENTON & MULLER ENVIRO TECH ERNST BALLARD GBLOOD TIC AT I TZMAN BER EXTERNALDISTRIBUTION NSIC BAER B
ER IME GAMMILL 2
SITE ANALYSIS VOLLMER BUNCH J i COLLINS KREGER CO ROL NUMBER REG IV Ja HANCHETT
- 16 CYS ACRS SENT CAT GO Y NRC FORM 19S (2-75) 2ooo2m
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LEBOEUFI LAMB,LEIBY K MACRAE I757 N STREET, N.W.
WAsHINGToN,D. C. 20036 LEON A. ALLEN,JR.
JOSEPH E. BACHELDKR.ZIE ERNCST S
BALLARDoJR.
G. S. PETER BCRGCN GEOFFRY D.C BEST DAVID P. DICKS TAYLOR R. BRIGGS CHARLES N.BURGER THOMAS E.BURKE ROGER D.FCLDMANF EUGENE R. FIDELL~4 JACOB FRIKDLANDKR DONALD J. GREENE JAMES A. GRCCR,XE 4 JOHN L. GROSE 4
DOUGLAS W. HAWKS CARL D. HOBKLMAN MICHAEL IOVENKO JAMES F. JOHNSON> 4' RONALD D. JONES LKX K ~ LARSON44 GRANT S. LEWIS CAMERON F. MACRAK 4 CAMKRON F. MACRAC,1IE 4 GERARD A.MAHER SHEILA H. MARSHALL JAMES G. McKLROY JAMES P. McGRAN CRY. JRF4 PHILIP PALMER McGVIGAN JAMES O'ALLEY,JR. 4 J. MICHACL PARISH JOHN A.RUDY PAUL G. RUSSELL HAROI,D M. SEIDEL CHARLES P SIFTON HALCYON G. SKINNER JOSCPH S. STRAUSS SAMUEL M.SUGDKN EUGENE B THOMAS'RF4 LEONARD M. TROQTCN 4
HARRY H VOIGT 4 H. RICHARD WACHTKI GERARD P. WATSON FRKSIDCNT PARTNERS WASHINGTON OFFICE I ADMITTED TO THC DISTRICT OF COLUMBIA BAR TELEPHONE 202 457 7500 CABLE ADDIIESS LEBWINsWASHINCTONi D. C.
TELEX'40274 July 181977 Ikt<O))5
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~Sod RANDALLJ. LEBOEUF, JR. I929 I975 ADRIAN C LEIBY I952 I970 CF CCVNSEL ARVIN C. UPTON I40 BROADWAY NEW YORK,N.Y. I0005 TELEPHONE 2I2 299 IIOO CABLE ADDRESS LEBWIN,NEW YORK TELEX: 4234IB Mr. Edson G.
Case Acting Director Office of Nuclear Reactor'egulation'.S.
Nuclear Regulatory Commission Washington, D.C.
,20555 Re:
Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station, Uni,t No.
1'Do'cke't No'.
5'0-'220'ea,r Mr, Ca,se:
3,ow3;n,g As counsel for Niagara Mohawk, I enclose the fol-Q)
Three Q3). ori'ginals and nineteen Q9) copies of an apple",cation for amendment to operating license; and (2)
Forty (40} copies each of two (2) documents entitled Attachments A, and B which "set forth the requested change zn the 'Technical Specifications along with its technical basis.
Very truly yours,
- LeBOEUF, LAMB, LEXBY 6 MacRAE Eug e B. Thomas, Jr.
Enclosures I 772000245
E 1
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A, I,
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.
1)
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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 Cor-poration, holder of Facility Operating License No.
DPR-63, hereby requests that Sections 2.1.1, 2.1.2, 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 the Safety Review and Audit Board.
The proposed Technical Specification changes are set forth in Attachment A to this application.
Supporting Information, which 'demonstrat'es that the proposed changes do not involve a significant hazards consideration, is set forth in Attachment B.
The proposed changes 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.
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WHEREFORE, Applicant respectful ly requests that Appendix A to Facility Operating License No.
DPR-63 be amended in the form attached here to as Attachment A.
NIAGARA MOHAWK POWER CORPORATION B'erald Rhode Vice President Engineering Subscribed agd sworn to before me this //~ day of July, 1977.
NOTARY PUBLIC PHYLL(S O. VOYTKO NOtary PublfC In thO ~erato Ol t aW YOrk Qrralt(ieo in anon.Co.
tro. 34 943553a My Commraalon Cxplrea March 30. 19 7j
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ATTACHMENT A Niagara Mohawk Power Corporation License No.
DPR-63 Docket No. 50-220 Pro osed Chan es To The Techni:cal' eci;fications (A
endix A)
Replace Pages 8,
20,
- 64a, 64b,
- 64c, 70a,
- 70b, and 70c with the attached revised pages.
Remove Pages
- 64d, 64e and 64f.
160
~ ICO 120 ttOTES I. RAT EO POP/E R IS 1850 tiwt
- 2. DESIGN PLOY/ IS 07.5 x 10G Ib/br
- 3. DESIGN TOTALPEAKINGFh TOR ~ PF
- 4. COfIE PRESSURE IS ~ 800 p:is SCRN 0 '
I 100 0
O 80 D
z0 D
60
.z ROD B 40 20 S - TPF S
TPF 308 for 7y'7FUEL h4TPF TPF 3.02 Ior8x8 FUFL S~" x So WHEREt S"
THE NEV/SCRAM AND ROD BLOCK MTPF CALCUI ATED MA:rltIUMTOTALPEAKING FACTOR S
SCAhtif 5 ROO BLOCK SHOWt/ ABOYE 0
0
'10 20 30 40 60 60
.70 80 110 120
'ECIRCULATIONFLOV(, PFRCENT OF OES!GN Fi gare 2'.l.3.
F1o,v Biased ScrafA and APRIL Rod BIock
REFERENCES FOR BASES 2.1.1 AND 2.1.2 FUEL CLADDING (1)
General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO-10958 and NEDE-10958.
(2)
Linford, R. B., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor,"
NED0-10801, February 1973.
(3)
FSAR, Volume II, Appendix E.
(4)
FSAR, Second Supplement.
(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 Power and Light Company, dated November 22, 1967 and January 9, 1968.
(8)
Technical Supplement to Petition to Increase Power Level, dated April 1970.
(9)
Letter, T. J.
- Brosnan, 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.
! (ll) Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis, NEDO 24012, May, 1977.
20
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT hlinimum Critical Power Ratio (MCPR) hlinimum Critical Power Ratio (MCPR)
During power operation hlCPR shall be 1.37 for 7x7 fuel and >
1.38 for 8x8 fuel at rated power and flow. If at any time during power operation it is determined by normal surveillance that these limits are no longer met, action shall be initiated within 15 minutes to restore operation to within the prescribed limits'f the operating hlCPRs are 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.
hlCPR shall be determined daily during reactor power operation at ) 25'o rated thermal power.
de Power Flow Relationshi Compliance with the power flow relationship in section 3.1.7.d shall be determined daily during reactor operation.
For core flows other than rated the hlCPR limits shall be the limits identified above times Kf where Kf is as shown in Figure 3.1.7-1.
Power Flow Relationshi Durin Power 0 eration The power/flow relationship shall not exceed the limiting values shown in Figure 3.1.7.f.
64a
LihfITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREh1ENTS If'at any time. during power operation it is determined by normal surveillance that the limiting value for the power/flow relationship is being exceeded action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the power/
flow relationship 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.
e.
Re ortin Re uirements If any of the limiting values identified in Specification 3.1.7.a, b, c and d are exceeded, a Reportable Occurrence report shall be submitted.
If the corrective action is taken, as described, a thirty-day written report will meet the requirements of this specification.
64b
~Fi
.1.7.
NINE NILE POINT UNIT 1
LII)ITIN(jPONE(
FLOW LINE 100 80 S-60 5
I 40 O
Limiting Power/Flow Line 20 20 40 60 80 100 PERCENT RATED CORE FLOlJ 64c
BASES FOR 3. 1.7 AND 4. 1.7 FUEL RODS of the plant, a
MCPR evaluation will be made at the 25K thermal power level with minimum recirculation pump speed.
The HCPR margin will thus be demonstrated such that future HCPR evaluations below this power level will be shown. to be unnecessary.
The daily requirement for calculating NCPR about 25K 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 HCPR when a limiting control rod pattern is approached ensures that HCPR will be known following a chang'e in power or power shape (regardless of magnitude) that could place operation at a thermal limit.
Figure 3.1.7-1 is used for calculating HCPR 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 MCPR at 100K 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 99.9X limit NCPR when core flow reaches the maximum possible core flow corresponding to a particular setting of the recirculation pump MG set scoop tube maximum speed control limiting set screws.
These screws are to be calibrated and 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 HCPR.
This will assure that the reduction in MCPR associated with an inadvertent flow increase always satisfies the 99.9X requirement.
Irrespective of the scoop tube section, 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 power/flow line.
The analysis(>) 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 full power within the design basis for the plant.
7oa
BASES FOR 3.1.7 AND 4.1.7 FUEL RODS Re ortin Re uirements The LCO's associated with monitoring the fuel rod operating conditions are required to be met at all times, i.e., there is no allowable time in which the plant can knowingly exceed the limiting values of HAPLHGR, LHGR, and NCPR Power/Flow Ratio.
It is a requirement, as stated in Specifications 3.1.7.a, b, c
& d that if at any time during power operation, it is determined that the limiting values for MAPLHGR, LGHR or h)CPR Power/Flow Ratio are exceeded, action is then initiated.to restore operation to within the prescribed limits.
This action is initiated as soon as normal surveillance indicates that an operating limit has been reached.
Each event involving operation beyond a specified limit shall be reported as a Reportable Occurrence.
If the specified corrective action described in the LCO's was taken, a thirty-day written report is acceptable.
70b
REFERENCES FOR BASES 3. 1.7 AND 4. 1.7 FUEL RODS (1)
"Fuel Densification Effects on General Electric Boiling 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:
V. A. Moore to I. S. Mitchell, "Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.
(4)
"General Electric Boiling Water Reactor Generic Reload Application for 8 x 8 Fuel," NED0-20350, Supplement 1 to Revision 1, December 1974.
(5)
"General Electric Company Analytical Model for Loss of Coolant Analysis in Accordance with 10CFR50 Appendix K,"
NED0-20566.
(6)
General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by
- letter, G. L. Gyorey to Victor Stello Jr., dated December 20, 1974.
(7)
"Nine Mile Point Nuclear Power Station Unit 1, Load Line Limit Analysis," NED0-24012.
70c
ATTACHMENT B Niagara Mohawk Power Corporation 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 re-quired 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 Load Line Limit Analysis License Amendment Submittal, NEDO 24012."
Each transient and accident was analyzed along the new power/flow line.
The res'ults show that the 100/100 percent power/flow point is the most limiting for all accidents and transients except for the Rod Withdrawal Error.
Results of the Rod Withdrawal Error analysis show a slightly higher
(+0.01)
QCPR at the 91/75 percent power/flow point for 7 x 7 fuel.
The analysis also indicates that the new power/flow envelope maintains previously established safety limits.
To take advantage of as much of the analyzed envelope as possible, a change 'is required in the APRM flow biased rod block line and the 120 percent flow biased flux scram line.
The current APRM rod block and flux scram lines are overly conservative at low power/low flow conditions.
The new rod block and scram lines maintain previously estab-lished safety limits.
Currently, power'oid limits (B-Factors) are utilized at Nine Mile Point Unit 1 to assure compliance with assembly void fraction assumptions used in the Loss of Coolant Accident analysis.
Inclusion of these limits in the Technical Specifications are 'unnecessary since MCPR opera-ting limits and the proposed power/flow limit line provide adequate assurance that operating conditions will be more conservative than the initial conditions assumed in the Loss of Coolant Acci'dent analysis for Nine Mile Point Unit 1.
The initial MCPR value used in the Loss of Coolant Accident analysis was 1.19.
In determining the analysis
- basis, the bundle axial power shape was varied in both posi-tion and peaking factor in order to determine a conservative void fraction consistent with this MCPR.
The Technical Specifications on MCPR operating limits at rated conditions are in excess of 1.3.
Thes'e higher MCPR values would result in void fraction lower than those calculated from a 1.19 MCPR.
Furthermore, the fact that the operating MCPR at reduced power and flow must increase in accordance with the Kf factor in the Technical Specifications also con-tributes to a lower void fraction.
Since the time to dryout is directly proportional to the volume of water initially in the 'fuel bundle, a lower void fraction would result in a longer dryout time.
Hence; the current analysis basis is conservative.'he Loss of Coolant Accident analysis performed at the 100/100 percent power/flow point has been found to be limiting in relation to other points along the proposed power/flow curve.'y specifying that operation of Nine Mile Point Unit 1 will be within the 'envelope bounded by the proposed power/flow curve, additional assurance is provided that the current Loss of Coolant Accident analysis basis is conservativ'e 'with respect to all operating con-ditions.
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