ML20028D395
| ML20028D395 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/06/1983 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Power Authority of the State of New York |
| Shared Package | |
| ML20028D396 | List: |
| References | |
| DPR-59-A-072 NUDOCS 8301190066 | |
| Download: ML20028D395 (13) | |
Text
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UNITED STATES ge NUCLEAR REGULATORY COMMISSION gg-j W ASH WOTON, D. C. 20555 Y.. * [
POWER AUTHORITY OF THE STATE OF NEW YORK s..
DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No 72.
License No. DPR-59 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Power Authority of the State
, of New York (the licensee) dated February 20, 1981, as supplamented by letters dated November 18, 1981 and February 19, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954 as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter 1:
B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amndment can be conducted 6:ithout endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the 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 Comission's regulations and all applicable requirements have been satisfied.
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8301190066 830106 PDR ADOCK 05000333 P
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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility. 0perating License No. DPR-59 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
72, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
O 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: January 6,1983 l
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ATTACHMENT TO LICENSE AMENDMEtiT NO. 72
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FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Revise the Appendix "A" Technical Specifications as follows:
Remove Replace vii vii 6
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8 10 10 12 12 15 15 23 23 41 41 43 43 73 73
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1 JAFNPP LIST CF FIGUPIS Figure Title Page l
3.1-1 Manual Flow Control 47a 3.1-2a Operating Limit MCPR versus'TIFor 8X8 Fuel Types 47b 3.1-2b Operating Limit MCPR versus'tfFor 8X8R Fuel Types 47c 1
3.1-2c Operating Limit MCPR versus'{'For P'8X8R Fuel Types 47d 4.1-1 Graphical Aid in the Selection of an Adequate Interval 48 Between Tests 4.2-1 Test Interval vs. Probability of System Unavailability 87 3.4-1 Sodium Pentaborate Salution Volume-Concentration 110 Requirements 3.4-2 Saturation Temperature of Sodium Pentaborate Solution 111 3.5-3 MAPLHGR Versus Planar Average Exposure 135a
^
Reload 1, 8D274L 3.5-4 MAPLHGR Versus Planar Average Exposure 135b Reload 1, 8D274H 3.5-5 MAPLHGR Versus Planar Average Exposure 135c Reload 2, 8DRB265L t
l 3.5-6 MAPLHGR Versus Planar Average Exposure 135d Reload 2, 8DRB283 3.5-7 BAPLHGR Versus Planar Average Exposure 135e Reload 3, P8DRB265L I
3.5-8 MAPLHCR Versus Planar AVERAGE Exposure 135f Reload 3, P8DRB283 3.5-9 MAPLHGR Versus Planar Average Exposure 135g Reload 4, P8DRB284L 135h 3.5-10 MAPLHGR Versus Planar Average Exposure i
Reload 4, P8DRB299 3.6-1 Reactor Vessel Thermal Pressurization Limitations 163 4
4.6-1 Chloride Stres: Cerrecien Test Result? at 5000F 164 259 l
6.1-1 Management Organization Chart 260 6.2-1 Plant Staff Organization N,J,44, p 72 d
y11 Amendment No.
a JAFNPP surveillance tests, checks, calibrations, and V.
Electrically Disarmed Control Rod examinations shalI be performed within the i
specified surveillance intervals. These intervals To disarm a rcxl driven electrically, the tour may be adjusted + 25 percent. The interval as amphenol type plug connectors are removed pertaining to instrument and electric surveillance from the drive insert and withdrawal shall never exceed one operating cycle.
In cases solenoids rendering the rod incapable of where the elapsed interval has exceeded 100 withdrawal.
This procedure is equivalent percent of the specified interval, the next to valving out the drive and is preferred, surveillance interviil shall commence at the end Electrical disarming does not eliminate of the original specified interval.
position indication.
U.
Thermal Parameters W.
Iligh Pressure Water Fire Protection System 1.
Minimum critical power ratio (MCPR)-Ratio The liigh Pressure Water Fire Protection of that power in a fuel assembly which is System consists of:
a water source and calculated to cause some point in that fuel pumps; and distribution system piping with assembly to experience boiling transition associated post indicator valves (isolation to the actual assembly operating power as valves).
Such valves include the yard calculated by application of the GEXL hydrant curb valves and the first valve correlation (Reference NEDE-10958).
ahead of the water flow alarm device on each sprinkler or water spray subsystem.
2.
Fraction of Limiting Power Density - The ratio of the linear heat generation rate X.
Staggered Test Basis (LilGR) existing at a given location to the design LIIGR for that bundle tyl... The A Staggered Test Basis shall consist er:
design LilGRis 13.4 KW/f t for DX8, 8X8R and P8X8R bundles.
a.
A test schedule for a system, sub-systems, trains or other designated 3.
Maximum Fraction of Limiting Power Density-components obtained by dividing the The tiaximum Fraction of Limiting Power specified test interval into n equal Density (MPLPD) is the hightest value exist-subintervals.
ing in the core of the Fraction of Limiting Power Density (FLPD).
b.
The testing of one system, subsystem, train or other designated component 4.
Transition Boiling - Transition boiling means at the beginning of each subinterval.
the boiling region between nucleate and film boiling. Transition boiling is the region Y.
Rated Recirculation Flow in which both nucleate and film boiling occur intermittently with neither type being com-That drive flow which produces a core flow 6
pletely stable.
of 77.0x10 lb/hr.
Amendment No. #f,p f 72 6
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.l AFill'I' 1.1 (cont'd) 2.1 (cont'd)
A.1.b.
APitM P1ux Scram Trip SeLting (RoIue1 or_
Start & ilot Standby Mode)
APRM - The APRM flux scram setting shalI be < 15 percent of rated neutron flux with the Reactor Mode Switch in Startup/Ilot 13.
Core Thermal Power Limi.t (Reactor Pressure Standby or Refuel.
L 785 psig) c.
APRM Flux Scram Trip Settings (Run 11 ode)
When the reactor pressure is < 785 psig or core flow is less than 10% of rated, the (1) Flow Referenced ticutron Flux Scram core thermal power shall not exceed 25 Trip Setting I ercent of rated thermal power.
When the Mode Switch.is in the RUtl C.
Power Transient position, the APRM flow referenced flux scram trip setting shall be:
To ensure that the Safety Limit established in Specification 1.1. A and 1.1.H is not S < 0.66 W + 54%
e:<ceeded, each required scram shall be initiated by its expected scram signal, where:
The Safety himit shall be assumed to be e<ceeded when scram is accomplished by a S = Setting in percent of rated ruoans other than the expected scram signal.
thermal power (2436 Mwr)
W= Recirculation flow in percent l
of rated For no combination of recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exc;eed 117% of rated thermal power.
AmenilmentNo.1/, ajf, yf 72 8
9
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1.1 (cont'd) 2.1 (cont'd) j A.1.d.
APRM Rod Trip Setting The APIW1 Rod block trip setting shall be:
j S < O.66 W + 42%
where:
s = Rod block setting in percent ot-thermal power (2436 MWT)
W = Recirculation flow rate in percent l
of rated In the event of operation with a maximum fraction limiting power density (MPLPD) greater than the fraction of rated imt' r (FRP), the setting shall be modified as totlows:
S<
(0.66 W + 42%)f FRP I
[s f1FLPD where:
FRP = fraction of rated thermal power (2436 MWT)
MFLPD = maximt.m fraction of limiting power density where the limiting power density is 13.4 KW/ft for fiX8, 8XilR and P8X8R fuel.
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The ratio of FRP to MPLPD shall be set equal y
to 1.0 unless the actual operating value is
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less than the design value of 1.0, in which case the actual operating value will be used.
Amendment No.
g g,g 72 10 i
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JAl'! app 1.1 BASES A.
Reactor Pressure >785 psig and' Core Flow >
I 10% of Rated e
l.1 FU15L CtADDI!!G INTEGRITY Onset of transition boiling results in a de-The fuel cladding integrity limit is set such crease in heat transfer from the clad and, that no calculated fuel damage would occur as therefore, elevated clad temperature and the a result of an abnormal operational transient.
possibility of clad failure. However, the Because fuel damage is not directly observ-existence of critical power, or boiling trans-i able, a step-back approach is used to establish ition, is not a directly observable parameter I
a Safety Limit such that the minimum critical in an operating reactor. Therefore, the mar-power ratio (MCPR) is no less than 1.07.
MCPR>
gin to boiling transition is calculated from l
1.07 represents a conservative margin relative plant operating parareters such as core power, to the conditions required to maintain fuel core flow, feedwater temperature,-and core l
cladding integrity. The fuel cladding is one power distribution.
The margin for each fuel i
of the physical barriers which separate radio-assembly is characterized by the critical power i.
active materials from the environs. The in-ratio (CPR) which is the ratio of the bundle tegrity of this cladding barrier is related to power which would produce onset of transition l
its relative freedom from perforations or boiling divided by the actual bundle power.
cracking.
Although some corrosion or use re-The minimum value of this ratio for any bundle lated cracking may occur during the life of in the core is the minimum critical power ratio i
the cladding, fission product migration from (MCPR).
It is assumed that the plant operation l
this swirce is incrementally cumulative and is controlled to the nominal protective r,et-continuously measurable.
Fuel cladding per-toints via the instrumented variable, i.e.,
the forations, however, can result from thermal operating domain.
The current load line 1imit I
stresses which occur from reactor operation analysis contains the current operating donuin j
significantly above design conditions and the map.
The Safety Limit (MCPR of 1.07) has sufficient protection system safety settings.
While conservatism to assure that' in the event of an i
fission product migration from cladding per-abnormal operational transiant initiated from the l
foration is just as' measurable as that from MCPR operating limits specified for the normal j
use related cracking, the thermally caused operating conditions in speci fication 3.1.h, cladding' perforations signal a threshold, be-more than 99.9 % of the fuel rods in the core are yond which still greater thermal stresses may expected to avoid boiling transition. The margin cause gross rather than incremental cladding between MCPR of 1.0 (onset of transition boiling) i deterioration. Therefore, the fuel cladding and the Safety Limit is derived from a detailed v
Safety Limit is defined with margin to the statistical analysis.considering alJ of the un-conditions which would produce onset of trans-certainties in monitoring the core operating i
j ition boiling, (MCPR of 1.0).
These conditions state including uncertainty in the boiling transit-represent a significant departure from the ion correlation as described in Reference 1.
The j
condition intended by design for planned uncertainties employed in deriving the' Safety Limit are oleration.
Amendment No. J, jd, g, Jr6, f$ 72
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.l AFilP P BASES 2.1 FUEL CLADDIt3G INTEGRITY The abnormal operational transients applicable The most limiting transients have been to operation of the FitzPatrick Unit have analyzed to determine which result in the been analyzed throughout the spectrum of largest reduction in CRITICAL POWER RATIO.
planned operating conditions up to the thermal The type of transients evaluated were in-power condition 2535 MWt.
The analyses were based crease in pressure and lower, positive ulon plant operation in accordance with the operating reactivity insertion, and coolant temper-map given in the current load line limit analysis.
ature decrease. The limiting transient In addition, 2436 is the licensed maximum power level yicids the largest delta MCPR.
When4added of FitzPatrick, and this represents the maximum to the Safety Limit, the required ope at-steady-state power which shall not knowingly be ing limit MCPR of Specification 3.1.H is obtained.
exceeded.
Fuel cladding integrity is assured by the The evaluation of a given tr utsient begins operating limit MCPR's for steady state with the system initial parameters shown in conditions given in Specification 3.1.B.
the current reload analysis and reference These operating limit MCPR's are derived 2 that are input to a core dynamic behavior from the established fuel cladding integrity transient computer program described in Safet9 Limit, and an analysis of abnormal references 1 and 3.
The, output of these operational transients.
For any abnormal programs along with the initial MCPR form operating transient analysis evaluation with the the input for the further analyses of the initial condition of the reactor being thermally limited bundle with a single at the steady state operating limit, it channel transient thermal hydraulic code.
is required that the resulting MCPR The principal result of the. evaluation is does not decrease below the Safety Limit the reduction in MCPR caused by the transient.
MCPR_ at any time during the transient.
Amendment No. pf,.f4 72 15 C
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f Amendment No. J4 j( 72 23 (next page is 27) j
.I A FriPP TAlti.E 3.1-1 REACIOR PROTECTION SYSTEM (SCRAf t) INSTRUllENTATIO!! REQUIREMErff Minimum No.
flodes in Which Total of Operable Trip I.cVel Function Must be Number of i
Instrument Trip Function Setting operable Instrument ArtIon Channels
_ _ _ _Refue1 Startup Run Provided Channels (1) per Trip System (1)
(6) by Design for Itoth Trip Systems i
1 Mode Switch in X
X X
1 Mode Switch A
I Shutdown (4 Sections) i j
1 Manual Scram X
X X
2 Instrument A
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Channals
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3 IRM liigh Flux s120/125 of full scale X
X 8 Instrument A
Channels 3
IRM Inoperat tve X
X 8 Instrument A
Channels APRM Neutron Flux-s 15% Power X
X 6 Instrument A
2 Startup(15)
Channels l
2 APRM Flow Referenced S 5_(0. 66W+547.) x X
6 Instrument A or 11 Neutron Flux (12) (13)
FRP Channels tlFLPD
=
(14) (Not to exceed 117%)
2 APRM Fixed liigh Neutron s120% Power X
6 Instrument A o r 11 Flux (14)
Channels l
2 APRM Inoperative (10)
X X
X 6 Instrument A or B Channels 1/),.h. 4d 72 l
Amendment No.
x
JAFt2PP
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TABLE 3.1-1 (cont'd)
REACTOR PROTECTION SYSTEM (g RAM) IriSTRUMENTATIOri REQUIREMENT NOTES OF TABLE 3.1-1 (cont'd)
C.
liigh Flux IRM t
D.
Scram Discharge Volume fligh Level E.
APRM 15% Power Trip 7.
t;ot required to be operable when primary containment integrity is not required.
8.
Not required to be operable when the reactor pressure vesselhead is not bolted to the vessel.
9.
The APRM downscale trip is automatically bypassed when the IRM Instrumentation is operable and not high.
10.
An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 11 LPRM i
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inputs of t.he normal complement.
11.
See Section 2.1.A.1 9
12.
This equation will be used in the event of operation with a maximum fraction of limiting power density
(!!PLPD) greater than the fraction of rated power (FRP).
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where:
FRP = Fraction of rated thermal power (2436 31WT) r
' 1PLPD = Maximum fraction of limiting power density where the limiting power density is 13.4 i
i KW/ft for 8X8, 8X8R and P8X8R fuel The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
W = Recirculation flow in percent of rated S = Scram setting in percent of initial l
13.
The Average Power Range Monitor scram function is varied as a function of recirculation l
flow (W).
The. trip setting of this function must be maintainell in accordance with Specification 2.1. A. l.c.
Amendment No. jW, k,'
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JAFf1PP TAllI.E 1.2-3 (Cont'd)
IllSTRUMEfJTATIOr1 TI'.AT IllITIATES COf1 TROT. ROD BLOCKS tJOTES FOR TABLE 3.2-3 (cont'd)
The APRM and RRM rod blocks need not be operable in start-up mode.
From and after the time it is found that tt$e first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that curing that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped.
From and after the time it is found that the first column cannot be met for both trip systems, the systems shall be tripped.
2.
IRM downscale is bypassed when it is on its lowest range.
3.
This function is bypassed when the count is > 100 cps.
4.
One of the four SRM inputs may be bypassed.
5.
This SRM Function is bypassed when the IRM range switches are on range 8 or above.
6.
The trip is bypassed when the reactor power is < 30%.
s 7.
This function is bypassed when the Mode Switch is placed in Run.
8.
S = Rod Block Monitor Setting in percent of initial.
W = Recirculation flow in percent of rated K = Intercept values of 39%, 40%, 41% and 42% can be used with appropriate MCpR Limits from Section 3.1.B.
0 9.
When the reactor is subcritical and the reactor water temperature.is less than 212 F, the, control rod block is required to be operable only if any control rod in a control cell containing fuel is is not fully inserted.
10.
When the control rod block function associated with scram discharge instrument volume high water level is not operable when required to be operable, the trip system shall be tripped.
Amendment No. g, g. 72 73 9
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