ML17037C422
| ML17037C422 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/24/1977 |
| From: | LeBoeuf, Lamb, Leiby & MacRae, Niagara Mohawk Power Corp |
| To: | Rusche B Office of Nuclear Reactor Regulation |
| References | |
| Download: ML17037C422 (18) | |
Text
U.S. NUCLEAR REGULATORY C 'ISSION NUMBER NRC FoRM 195 I2.76)
MATERIAL'OCKET gO-2 Rd FILE NUMBER NRC DISTRIBUTE I'mW FOR PART 60 DOCKET TO: FROM: DATE OF DOCUMENT LeBoeuf, Lamb, Leiby & MacRae 3/24/77 Mr- Ben C. Rusche- Washington, D. C. DATE RECEIVED LeBoeuf, Lamb, Leiby & MacRae 3/25/77 8tl ETTER NOTOR IZED PROP INPUT FORM NUMBER OF COPIES RECEIVED
@ORIGINAL OCOPY~
5KINCLASSIFIED g/yes 4 DESCRIPTION ENCLOSU RE
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Ltr. trans the following: Amdt. to OL/change'o tech specs. "..."
concerns certain aspects of the controJ.
rod system ~ .notorized 3/21/77 '
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PLANT NAME Nine Mile Point Unit No. 1 RJL SAFETY FOR ACTION/INFORMATION ASSIGNED AD:
N PR C MANA E o ]c,C PROJECT MANAGER L C ASST LIC ASST INTERNAL DISTRIBUTION REG FILE SYSTEMS SAFETY PLANT SYSTEMS STE SA NRC PDR HEINEMAN TEDESCO I&E SCHROEDER OELD GOSSICK & STAFF ENGINEERING IPPOLITO MIPC MACARR ERNST CASE BOSNAK BALLARD SIHWEIL OPERATING REACTORS YOUNGBLOOD HARLESS PA CK STELLO SITE TECH PROJECT MANAGEMENT REACTOR SAFE OPERATING TECH GAMMILL BOYD ROSS EISENHUT STEPP P COLLINS NOVAK HULMAN HOUSTON ROSZTOCZY PETERSON CHECK SITE ANALYSIS MELTZ VOLLMER HELTEMES AT&I BUNCH SKOVHOLT SALTZMAN J COLLINS RUTBERG KREGER EXTERNAL DISTRIBUTION CONTROL NUMBER LPDRi NAT LAB > ) g8'7 0 /0 4 TIC: REG V IE ULR KSON OR NSIC 0 LA PDR ASLB: CONSULTANTS!
ACRS CYS Mba@/ E NRC FORM 195 {2 76)
>>ya LAW OFFICES OF LEBOEUF, LAMB,LEIBY 8c MAC RAE 1767 N SvREEv, N.W.
WAsHINGTQN, D. C. 20036 TCLCFHCNC 292 4d7.7SCC CAQLC ADCRCSS LEON A. ALLCN, JR. CAMERON F. MAcRAK 4 LCQWIN>>WASHIIIQTQII>>9 C '9 RANOALL J. LcQCCUF,JR. I929.IQ75 JOSEPH K. BACHELDER,ZB CAMKRON F. MAcRAE, IE 4 ERNEST S. BALLARD,JR. OERARD A.MAHER TCLCX> 449274 AORIAN C. LCIBY I9$ 2 l97B O S. PETER BKROCN 4 SHCILA H. MARSHALL
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DAVID P. DICKS JAMES O. McELROY TAYLOR R BRIGOS
~ JAMES P. McORANERY,JR 44 IIEITH BROWN L. MANNINO MUNTZINOi4 CHARLES N.BURGER JAMCS O MALLCY>>JR 4 WILLIAM O. DOUBi4 J. MICHAEL PARISH JACOB FRIKDLANDKR DONALD J ~ ORECNC PAUL O. RUSSELL HAROLD M SCIDEL March 24, 197
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JAMES A. GREER>>K 4 CHARLES P.SIFTON I 40 BROADWAY JOHN L. OROSE 4 HALCYON 0 SKINNER DOUOLAS W. HAWKS JOSEPH NCW YORK,N.Y. IOOOS CARL D HOBKLMAN
~ M.SUDDEN S.STRAUSS'AMUEL MICHAEL IOVKNKO EUGENE B. THOMAS, JR44 I TELEPHONE 2I2 259 IIOO JAMES F JOHNSON>> 4' LEONARD M. TROSTKN44.
RONALD D.JONES HARRY H. VOIOT 44 CABLE ADDRESS r
LCX K. LARSON%4 H. RICHARD WACHTEL LCBWIN, NCW YORK ORANT S. LEWIS OKRARD P. WATSON
,C~ TCLCX: 4234IS 4 RESIDENT PARTNERS WASHINOTON OFFICE 4 ADMITTED TO THC DISTRICT OF COLUMBIA BAR Mr. Ben C. Rusche 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'.'5'0-220
Dear Mr. Rusche:
As counsel for the above-named licensee, we hereby transmit thx'ee (3} originals and nineteen (19) copies of a proposed amendment to the Techn'ical Specifications for the above-named facility. Also transmitted are forty (40) copies each of Attachments A and B which 'are the supporting data for the'equested change.
The'roposed Technical Specifications deal with certain aspects of the control rod system.
~ g~ Very truly yours; LeBoeuf, Lamb, Attorneys fox'iagara QSI Leiby
~ ~cIIII.BBCRE 6 'MacRae Mohawk Power Corporation Enclosures
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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-poration, holder of Facility Operating License No. DPR-63, hereby requests that Specification 3.1.1, 4.1.1 and Bases of the Technical Specifications and Bases set forth in I
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 or any increase in the amounts of effluents or any change in the authorized power level of the facility.
F' 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 Gerald K. Rhode Vice President-Engineering Subscribed and sworn to before me on this u<~
day of March, 1977.
NOT RY PUBLIC HAZE L CARIIIOK-Notary Public in tho Stato cf Now York Qualified In Onon. Co. No. 4524460 Aly Commission Expiros Alar'ch 30,10'PK
~ Q Attachment A Niagara l1ohawk Power Corporation License No. DPR-63 Docket No. 50-220 Proposed Changes to Facility Operating License Attached are revisions to Pages 29, 35 and 37 of Appendix A to Facility Operating License
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE UIREMENT (iv) The rod block 'function of the rod worth minimizer shall be verified by attempting to with- ~
draw an out-of-sequence control rod beyond the block point.
( b) Whenever the reactor i s in the (b) If the rod worth minimizer is inoper-startup or run mode below 205 able while the reactor is in the rated thermal power, no control startup or run mode bel4w 20% rated rods shall be moved unless the thermal power and a second indepen-rod worth minimizer is operable. dent operator or engineer is being except as noted in 4.1 .1 .b(3)(a).(iv), . used he shall verify that all rod positions are correct prior to com-
'r a second independent operator or engineer verifies that the opera- mencing withdrawal of each rod group.
tor at the reactor console is following the control rod program.
The second operator may be used as a substitute for an inoper-able rod worth minimizer during a startup only if the rod worth minimizer fails after withdrawal of at least twelve control rods.
(4) Control rods shall not be with-drawn for approach to critica-'ity unless at least three source range channels have an observed count rate equal to or greater than three counts per second.
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BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SYSTEM (2) The rod housing support is provided to prevent control rod ejection accidents. Its design is discussed in Section VII-E.* Procedural control shall assure that the housing supports are in place for all control rods.
(3) Control rod withdrawal and insertion sequences are established to assure that the maximum in-sequence individual control rod or control rod segments which are withdrawn could not be worth enough to cause the core to be more than 0.013 delta k supercritical if they were to drop out of the core in the manner defined for the Rod Drop Accident(.~) These sequences are developed prior to initial operation of the unit following any refueling outa'ge and the requirement that an operator follow the sequences is backed up by the operation of the RWM. This 0.013 delta k limit, together with the integral rod velocity limiters and the action of the control rod drive system, limits potential reactivity insertion such that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal/gm. The peak fuel enthalpy content of 280 cal/gm is below the energy content at which rapid fuel dispersal and primary system damage have been found to occur based on experimental data as is discussed in reference l.
Recent improvements in analytical capability have allowed more refined analysis of the control rod drop accident. Theqe geggnjguqg have been described'n a topical report, two supplemen s and letters to the AEC.<1)~2~~3~~4~P). By using the analytical models described in these reports coupled with conservative or worst-case input parameters, st has been determ>ned that for power levels less than 20K of rated power, the specified limit on in-sequence control rod or control rod segment worths will limit the peak fuel enthalpy content to less than 280 cal/gm.
Above 20K power, even multiple operator errors cannot result in a peak fuel enthalpy content of 280 cal/gm should a postulated control rod drop. accident occur.
The following conservative or worst-case bounding assumptions have been made in the analysis used to determine the specificed 0.013 delta k limit on in-sequence control rod or control rod segment worths. The allowable boundary conditions used in the analysis are quantified in re-ferences (4) and (5). Each core reload will be analyzed to show conformance to the limiting parameters.
- FSAR 35
II BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SYSTEM The'RWM provides automatic supervision to assure that out-of-sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences.
.It serves as an independent backup of the normal withdrawal procedure followed by the operator.
In the event that the RWM is out of service when required, a second independent operator or engineer can manually fulfill the operator-follower control rod pattern conformance function of the RWM. In this case, procedural control is exercised by verifying all control rod positions after the withdrawal of each group, prior to proceeding to the next group. Allowing substitu-tion of a second independent operator or engineer in case of RWM inoperability recognizes the capability to adequately monitor proper rod sequencing in an alternate manner without unduly restricting plant operations. Above 20Ã power, there is no requirement that the RWM be operable since the control rod drop accident with out-of-sequence rods will result in a peak fuel energy content of less than 280 cal/gm. To assure high RWM availability, the RWM is required to be operating during a startup for the withdrawal of a significant number'of control rods for any startup (4) The source range monitor (SRM) system performs no automatic safety function. . It does provide the operator with a visual indication of neutron level which is needed for knowledgeable and efficient reactor startup at low neutron levels. The results of reactivity. accients are func-tions of the initial neutron flux. The requiremgnt of at least 3 cps assures that any transient at or above the initial value of 10 8 of rated power used in the analyses of tran-sients from cold conditions. One operable SRM channel would be adequate to monitor the approach to critical using homogeneous patterns of scattered control rods. A minimum of three operable SRM's is required as an added conservation.
- c. Scram Insertion Times The revised scram insertion times have been established as the limiting condition for operation since the postulated rod drop analysis and associated maximum in-sequence control rod worth are based on the revised scram insertion times. The specified times are based on design requirements for control=rod scram at reactor pressures above 950 psig. For reactor pressures above 800 psig and below 950 psig the measured scram times may be longer. The analysis discussed in the next paragraph is still valid since the use of the revised scram insertion times would result in greater margins to safety valves lifting.
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Attachment B Niagara Mohawk Power Corporation License No. DPR-63 Docket No. 50-220 Supporting Information On December 23, 1976, we were notified by our fuel supplier that the Rod klorth Minimizer should be operable at power levels up to 20 percent instead of the current 10 percent as required by the Technical Specifications.
This change is necessary to limit the maximum fuel energy content to 280 cal/gm during a postulated con-rod drop accident. Supplement 3 to NED0-20360, I'rol
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"GE BWR Generic Reload Application for 8 x 8 Fuel" Rev. 1 dated September 25, 1975, provides additional supporting information.
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