ML20040G454
| ML20040G454 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/29/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Power Authority of the State of New York |
| Shared Package | |
| ML20040G455 | List: |
| References | |
| DPR-59-A-064 NUDOCS 8202160055 | |
| Download: ML20040G454 (34) | |
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7 UNITED STdTES NUCLEAR REGULATORY COMMISSION y
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WASHINGTON, D C.20666
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POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amenchent No. 64 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 ofNewYork(thelicensee)datedNovember 18, 1981 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regula-tions set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations. of the Commission;
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C.' There is reasonable assurance (1) 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 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 CPR Part 51 of the Comission's regulations and all appifcable mquimments have been satisfied.
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O B202160055 820129 PDR ADOCK 05000333 P
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2.' 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 Operating 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. 64 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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 29, 1982 e
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' ATTACHMENT TO LICENSE AMENDMENT NO. 64 FACILITY OPERATING LICENSE NO. DPR-59
' DOCKET NO.~50-333 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages:
Remove Insert vii vii 6
6 9
9 10 10 13 13 15 15 20 20 29 29 30 30 31 31 31a 35 35 43 43 47b i
47c 47d 123 123 124 124 130 130 131 1 31 i
134-134 135-135 135a 135a 135b 135b 135c 135c l
135d 135d 135e 135f 1359 135h 245 245
. _. _ _ _ _ ~ -.
t-JAFNPP
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LIST OF FIGURES Figure Title Page 1.1-1 APIM Flow Bias Scram Relationship to Normal Operating 23 Conditions 3.1-1 Manual Flow Control 47a 3.1-2a Operating Limit MCPR versus't' For 8X8 Fuel Typee 47b 3.1-2b Operating Limit MCPR versusT For 8X8R Fuel Types 47c 3.1-2c Operating Limit FCPR versusT For P8X8R Fuel Types 47d 4.1-1 Graphical Aid in the Selecticn of an Adequate Interval 48 Between Tests 4.2-1 Test Interval vs. Prnhahility of Systan'Unavailmhility 87 3.4-1 Sodium Pentahnrate Solution Volume-Concentration 110 Requirements 3.4-2 Saturstion Tsuperature of sodium Pentaborate. Solutien 111 l
3.5-3 MAPUIGR Versus Planar Average Exposure 135a Reload 1, 8D274L 3.5-4 MAPUiGR Versus Planar Average Exposure 135b Reload 1, 8D274H 3.5-S MAPUIGR Versus Planar Average Exposure' 135c Reload 2, 8DRB265L 3.5-6 MAPUIGR Versus Planar Average Exposure 135d Reload 2, 8DRB283 3.5-7 MAPUIGR Versus Planar Average Exposure 135e Reload 3, P8DRB?65L 3.5-8 MAPUIGR Versus Planar Average Exposure 135f Reload 3, P8DRB28 3 3.5-9 MAPUIGR Ve.rsus Planar Average Exposure 1359 Reload 4,P8DRB284L 3.5-10 MAPMiGR Versus Planar Average Exposure 135h Peload 4,P8DRB299 3.6-1 Peactor Vessel 'Ihermal Pressurization Limitations 163 4.6-1 Chloride Stress Cbtrosion Test Results at 500*F 164 6.1-1 Management Organization Chart 259 6.2-1 Plant Staff Organization 260
, ((, [, 64 Amnrlmnt No.
vil
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surveillance tests, checks, calibrations, and V.
Electrically Disarned Cbntrol Ibd ex minations shall be performed within the specified surveillance intervals. 'Ibese intervals
'Ib disarm a rod drive electrically, the four may be adjusted + 25 percent. 'Ihe interval as m yimusil type plug connectors are ru oved pertaining to iniitrunent and electric surveillance fran the drive insert and withdrawal shall never exceed one operating cycle. In cases solenoids rendering the rod irrapable of where the elapsed interval has exceeded 100 per-withdrawal. 'Ihis procedure is equivalent cent of the specified interval, the next surveil-to valving out the drive and is preferred.
lance interval shall conmence at the end of the Electrical disarming does not eliminate original specified interval.
position indication.
U.
'Ihermal Parameters W.
High Pressure Water Fire PrOLection Systen 1.
Minimtun critical power ratio (MCPR)-Ratio
'Ihe High Pressure Water Fire Protection of that power in a fuel assmbly which is Systen consists of: a water source and calculated to cause sane point in that fuel pmps; and distribution systen piping with assenbly 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.
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correlation (Reference NEDL'-10958).
ahead of the water ficw alarm &vice on each sprinkler or water spray subsysten.
2.
Fraction of Limiting Power Density
'Ihe ratio of the linear heat generation rate X.
Staw M Test Basis (UlGR) existing at a given location to the design U1GR. 'Ihe design UlGR is 13.4 KW/ft A St&mM Test Basis shall consist of:
for 8x8, 8x8R.and P8x8R bundles.
a.
A test schedule for a systens, sub-3.
Maxinun Fraction of Limiting Power Deasity-systens, trains or other designated
'Ihe Max.mun Fraction of Limiting Power umyc==i:nts obtained by dividing the Density (MFLPD) is the highest value exist-specified test interval into n equal subintervals.
ing in the core of the Fraction of Limiting Power Density (FLPD).
b.
'Ihe testing of one systen, subsysten, 4.
Transition Boiling - Transition boiling means train or other designated ocuponent the boiling region between nucleate and film at the beginning of each subinterval.
boiling. Transition boiling is the region in which both nucleate and film boiling occur 3
?
intermittently with neither type being ocm-pletely stable.
Amendment No.M 64 6
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JAENPP 1.1 (cont'd) 2.1 (mnt'd)
In the event of operation with a maximm
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D.
Reactor Water Ievel (Hot or Cold Shutdown Oondition) fraction of limiting power density (MPIPD) greater than the fraction of rated power Whenever the reactor is in the shutdown FRP), the setting shall be modified as condition with irradiated fuel in the follows:
reactor vessel, the water level shall rsot be less than that corresponding to S < (0.66 W + 54%)
FRP 18 in. (-146.5 in. indicated level)
HEIPD above the top of the active fuel when it is seated in the core.
where:
FRP = fraction of rated thermal power (2436 Wt)
WIPD = maximm fraction of limiting power density where the liiniting power density is 13.410f/ft for 8x8, Ex8R, and P8x8R fuel.
'Ibe ratio of f'RP' to WIPD shall be set equal to 1.0 unless thd actual operating value is less than the design value of 1.0, in which case the actual operating value will be use,d.
('2) Fixed High Neutron Flux Scram Trip Setting When the Mcde Switch is in the BUN position, the APIM fixed high flux scran trip setting.
shall be:
.Si120% Power Amendment No.
64 9
8 n-,
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JAENPP 1.1 (cont'd) 2.1 (cont'd)
A.l.d.
APIN Ibd Block Trip Setting The APIM Ied block trip setting shall be:
Si0.66W+42%
where:
S = led block setting in -Weit of themal power (2436 Mit)
W = Icop recirullation flow rate in -wwst of rated (rated loop rpirculation flow rate equals (34.2 x 10 lb/hr))
In the event of operation with a =v1=_= fractibn limiting power density (MFIPD) greater than the, fraction of rated power (FRP), the setting shall be modified as follows:
S {' (0.66 W + 42%)
- FRP L NIPD,
where:
FRP = fraction of rated themal power (2436 Mit)
WIPD = mwimm. fraction of limiting power density where the limiting power density is 13.410f/ft for 8x8, 8x8R and P8x8R fuel j
+=-
The ratio of FRP tofFfrD shall be set equal to 1.0 unless the actual operati.xj value is less than the design value of 1.0, in which
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case the actual operating value will be used.
imaamit No. f, f, $> 64 y
l N
1.1 (cont'd) provided at the beginning of each fuel cycle.
B. Core 'Ibermal Power Limit (Reactor Pressure Because the boiling transition correlation
( 785 psig) is based on a large quantity of full scale data there is a very high confidence that At pressures below 785 psig the core elevation.
operation of fuel assmbly at the Safety pressure drop (0 power, O flow) is greater Limit would not produce boiling transition.
than 4.56 psi. At low powers and flows this
'Ihus, although it is not required to establish pressure differential is traintained in the the safety limit, vMitional margin exists bypass region of the core. Since the pres-between the Safety Limit and the actual sure drop in the bypass region is essentially decurrence of loss of cladding integrity.
all elevation head, tie core pressure drop at low powers and flows will always be greater However, if boiling transition were to occur, than 4.56 psi. 3 Analyses show that with a i
clad perforation would not be expected. Cladding flow of 28 x 10 Es/hr bundle flow, bundle taperatures would increase to approximately pressure drop is nea::ly independent of bundle 1100'F which is below the perforation'tm per-power and has a value of 3.5 psi. 'Ihus, the ature of the cladding material. 'Ihis has been bundle flow with a 4.56 psi driving head will 3
verified by tests in the General Electric Test be greater than 28 x 10 lbs/hr. Full scale Beactor (GE7I'R) where fuel similar in design ATIAS test data taken at pressures frcm 0 to FitzPatrick operated above the critical heat psig to 785 psig indicate that the fuel as-flux for a significant period of time (30 min-ssbly critical power at this flow is approx-utes) without clad perforation.
imately 3.35 t@ft With the design peaking factors this corr %urls to a core thermal If reactor pressure should ever exceed 1400 psia power of more than 50%. 'Ihus, a core thermal during normal power operating (the limit of power limit of 25% for reactor pressures applicability of the boiling transition corre-below 785 psig is conservative.
lation) it would be assmed that the fuel cladding integrity Safety Limit has'been violated.
In addition to the boiling transition limit (Safety Limit) operation is constrained to a maxinum IDGR = 13.4 kw/ft for 8x8, 8x8R, and P8x8R fuel. At 100% power, this limit is reached with a maximtn fraction of limiting power density (MFIPD) equal to 1.0.
In the event of opera-tion with a MFLPD greater than the fraction of rated power (FRP), the APIN scram and rod block settings shall be adjusted as required' in Specifications 2.1.A.l.c and 2.1.A.l.21.
Imadruit No.
g, M 64 9
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JAENPP BASES 2.1
_EUEL CIADDING INIEGRITY The almormal operational transients appli-The nost limiting transients have been cable.to operation of the FitzPatrick Unit analyzed to determine which result in the have been analyzed throughout the spectrun largest reduction in CRITICAL POWER RATIO.
of planned operating conditions up to the she type of transients evaluated were in-thermal power condition of 2535 bWt. She crease in pressure end power, positive analyses were based upon plant operation in reactivity in.*.rtion, and coolant temper-accordance with the operating map given in ature decrease. The limiting transient Figure 3.7-1 of the ESAR. In addition, 2436 yields the largest delta ICPR. When added is the licensed maximtzn power level of Fitz-to the Safety Limit, the required operat-Patrick, and this represents the maxinum ing limit ICPR of Specification 3.1.B is obtained.
steady-state power which shall not knowingly be exceeded.
The evaluation of a given transient begins Fuel claMing integrity is assured by the with the systen initial parameters shown in operating limit PCPR's for steady state' the current reload analysis and reference 2 that are input to a core dynamic behavior conditions given in Specification 3.1.B.
transient acuputer program described in These operating limit MCPR's are derived references 1 and 3.
The output of these fran the establisixxl fuel cladding integ-rity Safety Limit, and an analysis of abnor-programs along with the initial PCPR form mal operational transients. Ebr any abnor-the input for the further analyses of.the thermally. limited bundle with a single mal operating transient analysis evaluation channel transient thermal hydraulic. code.
with the initial condition of the reactor The principal result of the evaluation is being at the steady state operating limit, the reduction in MCPR caused by the tran-it is required that the resulting FCPR sient.
does not decrease below the Safety Limit ICPR at any time during the transient.
O Ii=Acut No. [, 64
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l JAFNPP 2.1 BASES (cont'd)
C.
References
'l.
Linford, R B.,
" Analytical fiethxis of 3
Plant Tranr.ient Evaluations for the General Electric Boiling Water Reactor",
NEDO-10802, Feb., 1973.
2.
" General Electric Fuel Application" NEDE 24011-P-A (Approved revision ntaber applicable at time that reload fuel analyses are performed).
I 3.
" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" NEDO-24154, October, 1978 Amendment No. M, 64 20 (Next page is 23) i
+
-..*4-
(
l.2 and 2.2 BASES JAENPP l
'Ihe reactor coolant pressure boundary ANSI Oxle permits pressure transients up to l
integrity is an inportant barrier in the 20 percent over the design pressure (120% x prevention of uncontrolled release of 1,150 - 1,380 psig). 'Ihe safety limit fission products.
It is essential.that pressure of 1,375 psig is refererm.ul to the the integrity of this boundary be pro-lowest elevation of the Reactor Cbolant tected by establishing a pressure limit Systs.
to be observed for all operating corrli-tions and whenever there is irradiated
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fuel in the reactor vessel.
'Ihe pressure safety limit of 1,325 psig
'1he current ~ reload analysis shows that_the l
main steam isolation valve closure transient, as measured by the vessel steam space pressure indicator is equivalent to with flux scram, is the most severe evert i
- 1,375 psig at the lowest elevation of sulting directly in a reactor coolant ftWReactor Coolant Systs, 'Ihe 1,375 syst s pressure increase. 'Ihe reactor psig value is derived frtm the design vessel pressure code limit of 1,375 psig, pressures of the reactor pressure given in ESAR Section 4.2, is above the
. vessel and reactor coolant systs peak pressure produced by.the event above.
piping. 'Ihe respective design pressures
'Ihus, the pressure safety limit (1,375 psig) are 1250 psig at 575'F for the reactor is well above the peak pressure that can vessel,1148 psig at 568*F for the re-result frcm reasonably expected overpressure circulation suction piping and 1274 psig transients. (See current reload analysis for l
at 575*F for the discharge piping. 'Ihe the curve produced by this analysis.) Reactor j
pressure safety limit was chosen as the pressure is continuously indicated in the lower of the pressure transients permitted
' control rocin during operation.
by the applicable design codes:
1965 ASME Boiler and Pressure Vessel Oxle, Section A safety limit is applied to the Residual III for pressure vessel and 1969 ANSI Heat RErnoval Systs (IERS) when it is operating B31.1 Code for the reactor coolant system in the shutdown cooling mode. When operating piping. 'Ihe ASME Boiler and Pressure in the shutdown cooling mode, the IERS is vessel Code permits pressure transients included in the reactor coolant systs.
up to 10 percent over design pressure (110% x 1,250 -1,375 psig), and the
'Ihe ntanerical distribution of safety / relief valve setpoints shown in 2.2.1.B.(2 0 1090 psi, 2 01105 psi, 7 @ 1140 psi) is justified by anal--
yses described in the General Electric report!1:EDO-24129-1, Supplement 1, and assures that the structural acceptance. criteria set forth in the Mark I Contaiment Short Term Pr@u=u are satisfied.
Isir.r.3iient No. y/, 64 29
3.1 IJMITING CONDITIONS IVR OPERATION
- 4. l' SURVEILIANCE: RBQUIRENENIS
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3.1 RE:ACIOR PIDIECTION SYSTEM 4.1 REACIOR Pivm:n0N SYSTEM Applicability:
Applicability:
Applies to the instrinentation and associated Applies to the surveillance of the instru-devices which init,iate the reactor scram.
mentation and associated devices which initiate reactor scram.
Cbjective:
Objective:
To assure the operability of the Reacto.r Protection Systesn.
'Ib specify the type of frequency of surveil-lance to be applied to the protection instnnentation.
specification:
. A.
'Ihe setpoints, miniman ntuber of trip systcsns, Specification:
minimum number of instrinnent channels that nust i
be operable for each position of the reactor A.
Instnnentation systems shall be e
mode switch shall be as shown on Table 3.1-1.
functionally tested and calibrated as
'Ibe design systen response time frcm the opening indicated in Tables 4.1-1 and 4.1-2 of the sensor contact to and including the respectively.
opening of the trip actuator contacts shall B.
Maxinun Fraction of Limiting Power not exceed 50 meec.
Density (MFLPD)
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B.
Mininun Critical Power Ratio (MTR)
'Ibe WIPD stall be determined daily during During reactor power operation at rated power reactor popu operation at > 25% rated and flow, the PCPR operating limits shall thermal power and the APBM high flux scraun not be less than those shown below:
and Rod Block trip settings adjusted if ne m aary as required by Specifications 1.
When surveillance requirenent 2.1.A.l.c and 2.1.A.l.d,'r w t.ively.
4.1.E is met (TAWE b TS)
Amendment No.
, 64 30
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3.1 (Cont'd)
JAFNPP KTR Operating Limit for Incrmental C.
K PR shall be determined daily during Cycle Core Average Exposure reactor power operation at > 25% of rated trer-l.
mal power and following any change in power Fuel Type BOC to DOC-lGWD/t to level or distribution that would cause 1GWD /t before EOC EOC operationwith a limiting control rod pattern as described in the bases for At RBM trip level setting S = 0.66 W + 39%
Specification 3.3.B.5.
8x8 1.22 1.23 D.
When it is determined that a channel has 8x8R 1.22 1.23 failed in the unsafe condition, the P8x8R 1.22 1.25 other RPS ohazmbls that monitor the same variable shall be functionally At RBM trip level setting S = 0.66W + 40%
tested imtediately before the trip systen containing the failure is tripped.
8x8 1.24 1.24
'Ihe trip system containing the unsafe 8x8R 1.24 1.24 failure may be placed in the untripped P8x8R 1.24 1.25 condition during the period in which surveillance testing is being performed At RBM trip level setting S = 0.66 W + 41%
on the other RPS channels t
8x8 1.27 1.27 E.
Verification of the limits' set forth 8x8R 1.27 1.27 in specification 3.1.B.
shall be performed P8x8R 1.27 1.27 as follows:
At RBM trip level setting S = 0.66 W + 42%
1.
'Ibe average scram time to notch 8x8 1.31 1.31 8x8R 1.31 1.31 2.
'Ibe average scram time to notch P8x8R 1.31 1.31 position 38 is determined as follows:
n n
ME =,
Ni t'i Ni' T
i=1 i=1 where n = rumber of surveillancei'teist's performed to date in the cycle, Ni =
ntmber of active rods measured in 31 Amendment Ho. 64
JmFP
'2. If requirement 4.1.E.1 is not met (i.e. TB the ith surveillance' and Ti =
( T AVE) then the Operating Limit MCPR averag am to notch,
values (as a function of t) are as given in g
g Figure 3.1-2a, 3.1-2b, 3.1-2c msured in W m mine where t' = ( tg -t )/ (tg -ty t*8t-l 3
and g = the average scram time to notch 3.
'Ibe adjusted analysis mean scram l
position 38 as defined in speci--
time is calculated as follows:
fication 4.1.E.2, 7 = the adjusted analysis mean scram 6
1 time as defined in specification 4.1.E.3, I (sec)= A, +1.65 tr g
t = the scram time to notch position
{1 4
38 as defined in specification 1,1 3.3.C.1
- Note:
Should the operating limit PCPR obtained frcm this figure be less than the operating limit where J4,= mean of the distribution PCPR found in Specification 3.1.B.1 for the average scre for the applicable RBM trip level insertion time to notch setting then specification 3.1.B.1 position 38 = 0.723 sec.
shall apply.
F = standard deviation of the distribution for average scram insertion time to notch position 38=0.054 sec.
If anytime during reactor operation greater than N,= the total ntmber of active 25% of rated power it is determined that the limit-rods measured in specifi-ing value for MCPR is being exceeded, action shall cation 4.3.c.1 then be initiated within fifteen (15) minutes to, restore operation to within the prescribed limits.
'Ihe nmf'er of rods to be scram tested If the MCPR is not returned to within the prescribed
.and the test intervals are given in limits within two (2) hours, an orderly reactor specification 4.3.C.
power reduction shall be wmera ed inmediately.
'Ibe reactor power shall te reduced to less than 25%
of rated power within the next four hours, or until the MCPR is returned to within the proscribed limits.
For core flows other than rated, the MCPR operating limit shall be nultiplied by the appropriate kg is as shown in figure 3.141.
Imendment No. )#, 64 31a v
JAENPP 3.1 BASES (cont'd)
'nirbine control valves fast clonre initia'tes a scram based on pressure switches sensing
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electro-hydraulic control (EllC) system oil pressure. 'Ihe switches are located between fast closure solenoids and the disc dtmp valves, and are set relative (500 ( P ( 850 psig) to the normal (EIIC) oil pressure of 1,600 psig so that based on the small systs voltme, they can rapidly detect valve closure or loss of hydraulic pressure.
'1he requirunent that the I1Ws be inserted in the core when the APIWs read 2.5 I
indicated on the scale in the start-up and refuel modes assures that there is proper overlap in the neutron monitoring systs functions and thus, that adequate coveraoe is provided for all ranges of reactor operation B.
'Ihe limiting transient which determines the required steady state MCPR limit depends on cycle exposure. The operating limit MCPR values as" determined frm the transient analysis in the current reload sutmittal for various core exposures are given in Specification 3.1.B.
'1he EOCS performance analysis assmed reactor operation will be limited to MCPR = 1.20, as described in NEDO-21662-2. 'the Technical 3
Specifications limit operation of the reactor to the nore conservative MCPR based on consid-eration of the limiting transient as given in Specification.3.1.B.
Amerxhent No. g, 64 35
i JAENPP TABLE 3.1-1 (cont'd)
REACIOR PfunrnCN SYSTEM (SCRAM) INSTRLNENIATION RBQUIREMENP NC7IES OF TABIE 3.1-1 (cont'd)
C.
High Flux IIN D.
Scram, Discharge Volme Iligh IcVel E.
APIM 15% Power Trip 7.
Not required to be operable when primary contalment integrity is not required.
8.
Not required'to be operable when the reactor pressure vessel head is not bolted to the vessel.
9.
'Ihe APIM downscale trip is autmatically bypassed when the IIM Instrumentation is operable and not high.
- 10. An APIM will be considered operable if there are at least 2 IPIN inputs per level and at least 11 IPIN inputs of the normal cmpiment.
- 11. See Secti6n 2.1.A.1.
- 12. 'Ihis equation will be used in the event of operation with a maxinzn fraction of limiting power density (MFLPD) greater than the fraction of rated pwer (FRP).
where:
= Fraction of rated thermal power (2436 Mit)
WLPD
= Maximsn fraction of limiting power density where the limiting power density.is 13.4 l
KW/ft for 8x8, 8x8R and P8x8R fuel.
'Ihe 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 4
W
= Icop Recirculation flow in percent of rated (rated is 34.2 x 106 g)
S
= Scram setting in percent of initial
- 13. 'Ibe Average Power Range Monitor scram function is varied (Figure 1.1-1) as a function af recirculation loop flow (W). 'Ihe trip setting of.this function nust be maintained in accordance with Specification 2.1.A.l.c.
43
__ _ _.. _. _. _ _ _hiedient No. pf, 64
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FIGURE 3.1-2a Operating Limit MCPR Versus T (Cection 4.1.E) i I
For 8x8 Fuel Types l
1.40 ~
(1, 1.35) 1.35 ~
a:
a M
a
.e4 E
(1, 1.30) ' -
.6 S
O 4
1.30 -
L 0
E 90
'n.
m W
c.
O l.25 -
A.
(0.692, 1.242) goc to (0,, 1.23) 90C
( 0~, 1. 21) ( 0.16 7, 1. 21) 1.20 -
0 0.1 0.2 0'. 3 0.4 0'. 5 0'. 6 0'. 7 d.8 d.9
[.0 t'
Option B t =0 Option A t' =l 47b Amendment No. 64 g
t-s
'l,
- I FIGURE 3.1-2b Operating Limit MCPR Versus T (Section 4.1.E)
For 8X8R Fuel Types 1.40 -
(1, 1.35) 1.35 -
a:
mOz a
4 1.30 -
o w
a
.5 M
a es -
u Z
o A
1.25 -
SOC t-SOC (0.769, 1.256)
(0, 1.23) 1.20 -
l l
l l
l l
l l
l
-]
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 y
Option B t =0 Option A t =1 47c Ammendment No. 64 4
...... -. ~.
l
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FIGURE 3.1-2c Operating Limit MCPR Versus T (Section 4.1.E)
For P8X8R Fuel Types 1.40 (1, 1.31/
1.35-(1, 1.34) h h,%
s 1.30 9
a E
1 EOU
.5 1.25 y
(0, 1.25) goc go (0.600, 1.256) u 0
i O.
O
( 0,. 1. 2 2 )
1.20 1
I I
I I
I I
I I
l 0
0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 I
Option B 'L' =0 Option A 't' = l l
Amendment No. 64 47d
JADPP 3.5 (cont'd) 4.5 (cont'd) condition, that ptmp shall be considered 2.
Following any period where the IPCI inoperable for purposes satisfying Speci-subsystems or core spray subsystems fications 3.5.A, 3.5.C, and 3.5.E.
have not been required to be operable, the discharge piping of the inoperable syst s shall be vented fr m the high l
1 H.
Average Planar Linear Heat Generation Pate point prior to the return of the (APINGR) syst s to service.
[
'Ibe APIEGR for each type of fuel as a 3.
Whenever the HPCI, RCIC, or Core d
function of average planar exposure shall Spray System is lined up to take not exceed the limiting value shown in suction frm the uudensate storage l'
Figures 3.5.3 through 3.5.10.If anytime tank, the discharge piping of the during reactor power operation greater HPCI, RCIC, and Core Spray shall
(
than 25% of rated power it is determined be vented frm the high point of i
that the limiting value for APINGR is the system, and water flow observed 8
being exceeded, action shall then be on a monthly basis.
initiated within 15 minutes to restore operation to within the prescribed limits.
4.
'Ihe level switches located on the If the APIBGR is not returned to within Core Spray and PHR Syste discharge the prescribed limits within two (2) lxxirs, piping high points which monitor an orderly reactor power reduction shall be these lines to insure they are full u,ma ced inmediately. 'Ihe reactor power shall be functionally tested each shall be r d M to less than 25% of rated month.
T
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power within the next four hours, or until the APIRGR is returned to within the pre-H. Average Planar Linear Heat Generation Rate scribed lind.ts.
(APIRGR)
'Ihe APIBGR for each type of fuel as a function of average planar exposure shall be deterirdned daily during reactor operation at > 25% rated thermal power.
Amerrhent No. 36 64 123
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9 JMNPP 3.5 (cont'd) 4.5 (cont'd)
I.
Linear Heat Generation Rate (UlGR)
'Ihe linear heat generation rate (UIGR) of any I. Linear Heat Generation Rate (U1GR) rod in any fuel assably at any axial location shall not exceed the maxinun allowable UlGR of
'lhe U1GR shall be checked daily during l
13.4 KW/ft for 8x8, 8x8R and P8x8R burdles.
reactor operation at h 25% thermal power.
l If anytime during reactor power operation greater than 25% of rated power it is determined that the limiting value for UlGR is being exceeded, action shall then be initiated within 15 minutes to re-store operation to within the prescribed limits.
If the UIGR is not returned to within the pre-scribed limits within two (2) hours, an orderly reactor power reduction shall be wme.uced imne-diately.- '1he reactor power shall be reduced to less than 25% of rated power within the next four hours, or until the U1GR is returned to within the prescribed limits.
O G
im-x&mrt No. yf, 64 j
124
i JAENPP 3.5 BASES (cont'd) requirments for the emergency diesel generators, are within the 10 CFR 50 Appenli.x K limit.
'Ibe limiting value for APIIIGR is shown in G.
Maintenance of Filled Discharge Pipe Figure 3.5.3 through 3.5-10.
l If the discharge piping of the core spray, LPCI, I.
Linear Heat Generation Rate (IJ1GR)
ICIC, and HPCI are not filled, a water hamer can develop in this piping when the ptmp(s) are
'Ihis specification assures that the linear started. To minimize damage to the discharge heat generation rate in any red is less than piping and to ensure added margin in the operation the design linear heat generation.
of these systas, this technical specification requires the discharge lines to be filled when-
'Ihe IllGR shall be checked daily during reactor l ever the system is required to be operable. If operation at),25% cower to determine if fuel burn-a discharge pipe is not filled, the pmps that up, or control rod novment has caused changes in supply that line nust be assumed to be inoperable power distribution. For IliGR to be a limit-for technical specification purposes. However, ing value below 25% rated thermal power, if a water haromer were to occur, the system the ratio of local Il1GR to average IllGR would would still perform its design function.
have to be greater than 10 which is precluded by a considerable margin when enploying any H.
Average Planar Linear Heat Generation Rate (APIJIGR) permissible control rod pattern.
'Ihis specification assures that the peak cladding tsperature following the-postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50 Appendix K.
'Ihe peak clarirling tenperature following a postu-lated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assmbly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assmbly. Since expected local variations in power distribution within a fuel assmbly affect i
the calculated peak clad tenperature by less than + 20*F relative to the peak tsperature for a typical fuel design, the limit on the average linear heat generation rate is suf-ficient to assure that calculated tenperatures Tvrerdrent No. 48f o.64 130
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Amendment No. [ [ [, 64 l
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Reference:
NEDO-21662-2 (As Ammended August 1981) 135a Amendment No. )6',64
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Reference:
NEDO-21662-2 (As Armnended a
August 1981)
Amendment No. Jd, 64 135b
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NEDO-21662-2 (As Ammended 1
August 1981) l Amendment No.
, 64 135c 4
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NEDO-21662-2 (As Ammended August 1981) a 135f Amendment No. 73, 64
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NEDO-21662-2 (As Ammended l
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Amendment No. 64 135h 1
1
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M 5.0 DESIGN FEA'IURES B.
'Ihe reactor oore contains 137 cruciform-shaped control rods
~
5.1 SI'IE as described in Section 3.4 of thi FSAR.
A. 'Ihe Jees A. FitzPatrick Nuclear Power Plant is located on the PISNY 5.~3 REACIOR PRESSURE VESSEL portion of the Nine Mile Point site, approximately 3,000 ft. cast of the
'Ihe reactor pressure vessel is as l
Nine Mile Point Nuclear Station, Unit 1.
described in Table 4.2-1 and 4.2-2
'Ihe MiP-JAF site is on Iake Ontario of the FSAR. 'Ihe applicable design in Oswego Cbuntry, New York, approxi-CXJdes are described in Section 4.2 of the FSAR.
mately 7 miles northeast of. Oswego.
'Ihe plant is located at coordinates north 4,819, 545.012 m,. east 386, 968.945 m, 5.4 CONTAIM1ENT on the Universal Transverse Mercator A.
'Ihe principal design parameters System.
and characteristics for the B. 'Ihe nearest point on the property primary contalment are given in' i
line frcan the reactor building and Table 5.2-1 of the ESAR.
any points of potential gaseous B.
'Ihe secondary mntairunent is as l
effluents, with the exception of the lake shoreline, is located at the described in Section 5.3 and the northeast corner of the property.
applicable codes are as described i
'Ihis distance is approximately in Section 12.4 of the FSAR.
3,200 ft. and is the radius of the exclusion areas as defined in 10 CFR C.
Penetrations of the primary oon-taiment and piping passing through i
100.3.
such =WuExtiona are designed in E
h.widiuch with Utandards set forth 5.2 REACR)R in Section 5.2 c'l the ESAR.
5.5 EVEL S'IORAGE A. 'Ihe reactor oore consists of not more than 560 fuel asamblies. Ebr A.
'Ihe new fuel abKage facility design l
the current cycle three fuel types criteria are to maintain a K,ff dry are present in the core:
8x8, 8x8R and P8x8R. 'Ihese fuel 40.90 and flooded < 0.95.
I types are described in NEDO-24011.
Ompliance shall be verified prior to' i
'Ihe 8x8 fuel has 63 fuel rods and introduction of any new fuel design 1 water rod, and the 8x8R and to this facility.
P8x8R fuel have 62 fuel rods and 2 water rods.
Imadimit No. Jif,
- fif,
, 64 245
-