ML19351F339
| ML19351F339 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 12/12/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19351F338 | List: |
| References | |
| NUDOCS 8101120279 | |
| Download: ML19351F339 (13) | |
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M UNITED STATES
[,,, d/f(j NUCLEAR REGULATORY COMMISSION l
WASHINGTON, D. C. 20555 a
%,..... f NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 39 License No. OPR-63 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated September 25,1979, as supplemented October 22, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the appli-cation, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; i
D.
The issuance of this amendment will not be inimical to the l
comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-63 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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. 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION YtM v[b, Thoma
. Ippolito, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: December 12, 1980 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 39 FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Revise Appendix A as follows:
Remove Insert 63 63 64a 64a 64b 64b 64c 64c, d, e 70a 70a 70b 70b 70c 70c,d 1
4 1
Marginal lines indicate area of change.
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m m-I LlilITillG Coll 0ITI0tt FOR OPERAT10tl SURVEILLAl{CE REQUIREMENT 11.7 filEL It005 4.1.7 rilEL It00s gglicability:
hglicability:
The Limiting Conditions for Operation associated
' The Surveillance Requirements apply to t'he with the fuel rods apply to those parameters parameters which monitor the fuel rod operating which monitor the fuel rod operating conditions.
conditions.
i Objective:
Objective:
The objective of the Limiting Conditions for lhe objective of the Surveillance Requirements Operation 1:: to as:;ure the performance of the is to specify the type and frequency of fuel rods, surveillance to be applied to the fuel rods.
Speci fica tion:
Specification:
a.
Average Planar Linear IIcat Generation Itate a.
Average Planar I.inear llcal Generation state lhliiiGit)
[diliiiGii) i During power operation, the APLilGR for each The APlllGR for each type of fuel as a type of fuel as a function of average planar function of average planar exposure shall exposure shall not exceed the limiting value he determined daily during reactor operation shosm in Figures 3.1.7a 3.1.7h and 3.1.7c.
at > 25% rated thermal power.
if at any time during power operation it is
' determined by normal surveillance that the 2
limiting value for APillGit is being exceeded E
at any unde in the core, action shall be M
initiated within 15 minutes to restore S
operaLion to within the prescrlhed 1imits.
If z
the apt.llGit at all nodes in the core is not re-P turned to within the prescribed limits within tuo (2) hours, reactor power reductions shall i..
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he infilated at a : ate riot less than 101 per M
hour until APlllGR at all nodes is within the prescrlhed limits.
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT If at any time during power operation it is detennined 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, reactor power reductions shall be initiated at a rate not less than 10% per hour until the power / flow relationship is within the prescribed limits.
c.
Partial Loop Operation During power operation, partial loop operation is pennitted provided the following conditions are met.
When operating with three or more recirculation loops in operation and the remaining loops unisolated, the reactor may operate at 100 percent of full licensed power level in accordance with Figure 3.1.7aa.
No reduction in the APLitGR for each fuel type is required.
When operating with four recirculation loops in operation and one loop isolated, or with three recirculation loops in operation and one loop isolated and the remaining loop unisolated, the reactor nay operate at 100 percent of full licensed power in accordance with Figure 3.1.7aa and an APLHGR not to exceed 98 percent of the limiting values shown in Figures 3.1.7a, 3.1.7b, and 3.1.7c, provided the following conditions are met for the isolated loop.
1.
Suction valve, discharge valve and discharge bypass valve in the isolated loop shall be in the closed position and the associated motor breakers shall be locked in the open position.
Amendment No. 39 64b
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2.
Associated pump motor circuit breaker shall be opened and the breaker removed.
If these conditions are not met, core power shall be restricted to 90.5 percent of full licensed power.
When operating with three recirculation loops in operation and the tvlo remaining loops isolated, the reactor may operate at 100 percent of full licensed power in accordance with Figure 3.1.7aa and an APLilGR not to exceed 96 percent of the limiting values shown in figures 3.1.7a, 3.1.7b, and 3.1.7c, provided conditions 1 and 2 above are net for the isolated loops.
If these conditions are not met, core power shall be restricted to 90.5 percent of full licensed power.
Power operation is not permitted with less than three recirculation loops in operation.
If at any time during power operation it is determined by normal surveillance that the limiting valde for APlllGR under one and two isolated loop operation is being exceeded at 'any node in the core, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLilGR at all nodes in the core is not returned to within the prescribed limits for one and two isolated loop operation within two (2) hours, reactor power reduction shall be initiated at a rate not less than 10 percent per hour until APLilGR at all nodes is within the prescribed limits.
Amendment No. 39 64c
LIl41 TING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT f.
Recirculation Loops During all operating conditions with irradiated fuel in the reactor vessel, at least two (2) secirculation loop suction valves and their associated discharge valves will be in the full open position except when the reactor vessel is flooded to.a level above the main steam nozzles or when the steam separators and dryer are restioved.
g.
Reporting Requirements if any of the limiting values identified in Specification 3.1.7.a. b, c, d, and e 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.
Amendment No. 39 64d I
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o Figiere 3.1.7.
itIliF. li!LE P0lflT Uiill 1 LillITit;G P0llER FL0il lit!E
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20 40 60 80 100 PERCEtlT RATED CORE FL0ll Fmendment ilo. J'[, 39 64e
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DASES FOR 3.1.7 AllD 4.1.7 FUEL RODS of the piant, a IICPR evaluation will be made at the 25% tiscrmal power Icvel uith minimum recirculation pump speed.
The llCPit saargin ull) thus he demonstrated such th.it future llCPR evaluations below this power level The daily requirement f.ir calculating itCPR above 25% rated thermal power ulll he shown to be unnecessary.
is sulficient :.ince pouer distribution slij f ts are very slow uhesi Lliere liave not been significant liouer or The requirement for calculottinij llCP!! sdien a lisiilting control rod pattern is approach ~cd control rod changes.
ensures that itLPit will he tnoun folloulng a change in luar:r or pouer shape (regardless of magnitude) that could place operation at a Lliermal limit.
figure 3.1.7-1 is used for calcblating llCPR durinq operation at other than rated conditions.
for the case of automatic fisas control the Kg factor is delcrmined such that, any automatic lucrease in pouer (due to flow centrol) will always restalt in erriving at the unminal required HCPR at 1002 ouer.
for manual flow is determlau:d such that an inadvertent increase in core flou (i.e., operator crror or control, the Kf recirculation pump speed controller f ailure) uould result in arriving at the 99.91 Ilmit IICPil uhen core f ino r eache, the maximum possthle core flou correspennling to a particular setting of the recirculation pump 110. set scoop tube maximum speed control limiting set screus. These scrcus are to be calibrated and set to a particular value and uliencver the plant is operati:19 in manisal flou control the Kr deflued by that setting of the screas is to be used in liie determination of requircilllCPR.
This ullt assure thH 'im reduction in ItCrit associated with an inadvertent flou increase always satisfics the 99.9% respatrcmcn!.
brespective of siever less than inilly)g, the rctguired flCPfl is never alloucd to he less thain tiie nniiiinal 11rM the scoop tube settin
- Pouer/finu licialinnshigt 1.he poucr/floy ciarve is the locus of critical pouer as a function of flow from uhtch the occurrence of ahnonaal operating translents ull) ylcld results within deflued plant safety Ilmits.
Each transient and pos:ulated accident applicabic to operation of the plant was asiaiyzed along the power / flow line.
The analysid7) justifies the operating envelope hounded by the poucr/ flow curve as long as other operating lir.iits are satisfied.
Operation under the pouer/flou line is designed to enable the direct ascension to full poucr uithin line design basis for the plant.
l Amendment No. 39 70a an.
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BASES FOR 3.1.7 AND 4.1.7 FilFt RODS Partial Loop Operation i
The requirements of Specification 3.1.7e for partial loop operation in which the idle loop is isolated precludes the inadvertent startup of a recirculation punp with a cold leg. However, if these conditions cannot be met,' power level is restricted to 90.5 percent power based on current transient analysis (Reference 9).
{
The results of the ECCS calculation are not affected by one or more r,ecirculation loops being unisolated and out of l
service. This is due to the fact that no credit is taken for extended nucleate boiling caused by flow coastdown in the unbroken loops (Reference 10).
The results of the ECCS calculations are affected by one or nere recirculatton loops being isolated and out of service.
The mass of water in the isolated loops unavailable during blowdown results in a slightly earlier uncovery time for" the hot node. This results in an increase in the peak clad temperature. To assure peak clad' temperatures renain below 2200 F during steady state power operation with one or two recirculation loops isolated, analysis has shown that' the average linear lieat generation rate for each fuel type shall be reduced 2 percent and 4 percent respectively.
1 Partial loop operation and its effect on lower plenum flow distribution is sunnerized in Reference 11. Since the lower plenum hydraulic design in a.non-jet pump reactor is virtually identical to a jet pump reactor, application of these results is justified. Additionally, non-jet pump plants contain a cylindrical baffle plate which surrounds the guide tubes and distributes the impinging water jet and forces. flow in a circumferential direction around the outside of the j
baffle.
g Recirculation Loops Requiring the suction and discharge for at least two (2) recirculation loops to be full open assures that an adequate flow path exists from the annular region between the pressure vessel wall and the core shroud, to the core region.
This provides for communication between those areas thus assuring that reactor water level instrument readings are indicative of the water level in the core region.
When the reactor vessel is flooded to the level of the main steam line nozzle, communica, tion between the core region and annulus exists above the core to ensure that indicative water level nonitoring in the core region exists. When the steam separators and dryer are removed, safety limit 2.1.1d and e requires water level to be higher than 9 feet below minimum normal water level (Elevation 302'9").
This level is above the core shroud elevation which would ensure communication between the core region and annulus thus ensuring indicative water level monitoring in the core region.
Therefore, maintaining a recirculation loop in the full open position in the:e two instances are not necessary to ensure indicative water level nonitoring.
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Amendment No. 39 70b 1
i
BASES FOR 3.1.7 AND 4.1.7 FUEL RODS Reporting Requirenents I
The LCO's assoc'iated with monitoring the fuel rod operat'ing conditions are required to be met at all times, t.e.,
there is no allowable tisie in which the plant can knowingly exceed the limiting values of MAPLilGR, LHGR, MCPR, or Power / Flow Ratio.
it is a requirement, as stated in Specifications 3.1.7a, b, c and d that if at any time during power operation, it is detennined that the limiting values for MAPLilGR, LHGR, MCPR, or Power / Flow Ratio are exceeded, action is then initiated to restore operation to within the prescribed limits. This action is initiated i
as soon as normal surveillance indicates that an operating limit has been reached. E*ach event involving operation beyond a specified limit shall be reported as a Reportable Occurrence.
If the specified corrective action described in the LC0's was taken, a thirty-day written report is acceptable.
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q Amendment No. 39 70c l
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REFEREllCES FOR BASES 3.1.7 Atl0 4.1.7 FUEL RODS (1) " fuel Densification Ef fects on General Electric Boiling Water Reactor fuel," Supplements 6, 7 and 8, flEDM-10735, August 1913.
(2) Sup;>1ccent I to Technical Report on Densifications of General Electric Reactor fuels December 14, 1974 (USA Regulatory Staff).
(3) Com.'unica t ion:
V. A. Bloore to 1. S. Ilitchc11. "ttodified GE flodel for fuel Densification," Docket 50-321, 1: arch 27, 1974.
(4) " General Electric Dolling llater Reactor Generic Reload Appilcation for 8 x 0 fuel," fled 0-20360, Suppicment 1 to I:evision 1, December 1974.
I (S) " General Electric Company Analytical Model for loss of Coolant Analysis in Accordance with 10CfR50 Appendix K,"
lit 00-20566.
(6) General Electric Refill Reflood Calculation (Suppicment to SAFE Cod,e Description) transmitted to the USAEC by letter,- G. L. Gyorcy to Victor Stello Jr., dated December 20, 1974.
(7) "fline Ilile Point flucicar Pouer Station Unit 1, l.oad Line Limit Analysis," fled 0-24012.
'0) Licensing Topical lleport Generdi Electric Bolling Water Reactor Generic Reload fuel Appilcation.
titDE-240ll-P-A, August,1978.
(9) Final Safety Analysis Report, Nine. Mile Point Nuclear Station, fliagara Mohawk Power Corporation, June 1967.
(10) NRC Safety Evaluation, Anendment No. 24 to DPR-63 contained in a letter from George Lear, NRC, to D. P. Dise dated May 15, 1978.
(11) " Core F1mv Distribution in a General Electric Boiling Water Reactor as Measured in Quad Cities Unit 1,"
fled 0-10722A.
Amendinent No. 7A, )d', 39
'70d