ML20205S136
| ML20205S136 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/07/1988 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20205S142 | List: |
| References | |
| NUDOCS 8811100266 | |
| Download: ML20205S136 (17) | |
Text
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UNITsD STATES l
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I NUCLEAR REGULATORY COMMIS$10N l
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POWER AtlTHORITY OF THE STATE OF NEW YORK t
DOCKET NO. 50-333 i
JAMES A. FIT 7PATPICK NUCLEAR POWER PLANT AMENDPENT TO FAC!l.ITY OPERATING LICENSE s
Amendment No. !17 License No. OPR-59 1.
The Nuclear Reculatory Comission (the Coesnission1 has found that:
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j A.
The application for amendment by Power Authority of the State j
of New York (the licensee) dated July 29, 1988, complies with the standards and requirevnents of the Atomic Energy Act of 1954, as amended (ti Act) and the Comission's rules and regulations set i
j forth in If,CFR Chapter 11 l
B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of j
l the Comission, i
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There is reasonable assurance (1) that the activities authorized I
by this amendment can he conducted wfthout endangering the health i
and safety of the public, and (11) that such activities will be
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I conducted in compliance with the Cemission's regulations; D.
The issuance of this amendment will not be inimical to the corron defense and security or to the health and safety of the publict j
a and
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E.
The issuance of this emendment is in accordance with 10 CFR Part
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51 of the Comission's regulations and all applicable requirements j
have been satisfied.
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2.
Accordingly, the license is amnded by changes to the Technical
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Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.f?) of Facility Operating License
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No. OPR-59 is hereby amended to read as follows:
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8811100266 831107 t
FDR ADCCK 05000333 I
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(2) Technical Specifications e
I The Technical Specifications contained in Apperdices A and P, as revised throu'5 Arendment No.117, are f
hereby incorporated in the license. The licensee shall t
operate the fa:ility in accordance with the Technical l
Specifications, j
3.
This Itcense amendment is effective as of the date of its issuance, f
FOR THE NUCLEAR REGULATORY COMMIS$!ON l
+L/a. Cw Robert A. Capra, Otreetor Pro.iect Directorate I-1 Division of Reactor Pro,iects, !/!!
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Attachment:
Changes to the Technical t
Specifications l
I Date of Issuance: November 7 1988 1
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l ATTACHMENT TO LICENSE AMENOMENT NO.117 l
FACILITY OPERATING LICENSE No, OPR-59 DOCKET NO. 50-333 Revise Appendix A as follows:
Remove Pages Insert Pages vil vii 7
7 12 12 13 13 31 31 43a 43a 17a 47a 47b 47b 123 123 130 130 135h 135h 135k 135k 13F1 1351 245 245 L
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JAFWPP LIST or FImutta lilDLK1 1111A lasA 3.1-1 Manual Flow Control 47a l
3.1-2 Operating Limit MCPR versus 4?b 4.1-1 Graphic Aid f a the Selection of an Adequate Interval Between Tests 48 4.2 1 Test Interval vs. Probability of System Usavailability 47 3.4 1 Sodium Pentaborate Solutlos of Gystem Volume-Concentration Requirements 110 7.4-2 seturatica Temperature of Sodium Pentaborate Solution 111
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3.5-1 Thermal Power and Core Flow Limits of Specifications 3.5.J.1 and 3.5.J.2.
134 3.5-6 (Deleted) 1354 3.5-7 (Deleted 135e 3.5-8 (Deleted) 135f 3.5-9 MAPLNGR Versas Planar Average Raposure Reload 4, F8DRR284L 1359
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3.5-10 (Deleted) 135h i
3.5-11 MAPLNGR Versus Planar Average Erposure Reload 6, e
RP4DR3299 1351 3.5-13 MAPLNGR Versus Planar Average Esposure Reload 4, RD136A
.45k 3.5-14 AAPLNGR Versus Planar Average Erposure Reload 8. RD339A 1351 3.6 1 Reactor Vessel Pressure Temeerature Limits 163 l
1 4.6 1 Chloride Stress Corrosion Test Results at 500'T 164 i
6.1 1 Mausgement Organisation Chart 259 l
l 4.2 1 Plant Staf f organisation 260 i
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Ame n dme nt No. J4, >f, &S, 64, 74, 14, DS, 96, (15 117 l
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JArtePP 1.1 gDEL_fLAppleeGJNTEGRJII 2.1 (Ugh _C %IG15G INTEGRITI Applicability:
Apsilcability:
The Safety Limits er*4blished to preserve the f uel h Limiting safety System settlegs apply to trip cladding latogrity apply to those variables which settings of the imetroments and "Jwices which are monitor the fuel thermal behavler.
provided to prowest the feel cladelag integrity Safety Limits from helag escoeded.
Didac11YS s IM 9cL129:
The objective of the safety Limits is to establish h objective of the Limitief E*f*tY System Settlage limits below which the lategrity of the fuel cle6 ding la te deflee the level of the process verlables at as preserved, which aetematic protective acties is laitiated to proveet the feel cladding integrity Safety Limits from helag escoeded.
G99clllCBLiOngs SeecktkcaLlaans I
A.
RfDCL9f_.l'It&Eule } 783_JPalgWige Fjew )* 1R%
A.
Trly Settlaga ci_Reted m limiting safety system trip settings shall be The esistence of a minimum critical powe. eatio as specified heless l
(DeCPR) ters than 1.04 shall constitute violatfor.
of the twel cladding integrity safety limit.
1.
MERLL93 Flum Trie Settlaga hereattar called the safety Limit. As peCPF l
safety limit of 1.05 shall apply during single-a.
less - h IEBE fles scram setting shall i
I loop operettom.
be.et at f_129/125 of tr.11 scale.
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+5. p8 117 7
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JAFDPP l
1.1 BASES 1
1.1 TVEL CLADDING I)fTEGRITI l
j The fuel cladding integrity limit is set such that no elevated clad temperature and the possibility of j
calculated fuel damage would occur as a result of as clad failure. Bovever, the aulatence of critical abnormal operational transient. Because fuel damage power, or bolllag transitloa, is not a directly is not directly observable, a step-back approach is observable parameter la an operating reactor.
used to establish a Safety Limit such that the miel-Therefore, the margia to boiling transition is num critical power ratio (MCPR) is no loss than 1.04.
calculated from plar.t operatlag parameters such j
MCPR pl.04 representss a conservative margin relative as core power, core flow, fee hater temperature, to the conditions required to maintain fuel cladding and core power distributloa. The margia for each j
integrity. The fuel cladding is one of the physical fuel assembly is characterised by the critical i
barriers which a.eparate radicactive materials f rom power ratio (CPR) which is the ratio of the the environs. The integrity of this cladding barrier bundle power which would produce onset of transi-is related to its relative freedom from perforations tion boilleg divided by the actual bundle power.
or cracking. Although some corrosion or use related The minimum value of this ratio for any bundle la cracking may occur during the life of the cladding, the core is the slaimum critical power ratio j
fission product migration from this source is incre-(MCPR). It is assumed that the plaat operatior=
l mentally cumulative and continuously measurable.
la controlled to the aosiaal protective setpoints j
Fuel cladding perfcrations, however, can result from via the instrumented variable, i.e.,
the oper-
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thermal stresses which occur from reactor operation ating domain. The current load llan limit significantly above design conditions and the protec-analysis contains the current operating domain tion system safety settings. Mhile fission product map.
The Safety Limit (MCPR of 1.04) has migration from cladding perforation is just as sufficient conservatism to assure that la the
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meesurable as that from use related cracking, the event of an abnormal operational transient j
thermally caused cladding perforations signal a initiated from the MCPR operatlag conditions la threshold, beyond which still greater thermal specification 3.1.5, more thm= 99.9% of the fuel stresses may cause gross rather than incremental rods la the core are espected to avoid boiling cladding deterioration. Therefore, the fuel cladding transition. The MCPR fuel claddlag safety limit Safety Limit is defined with margin to the conditions is increased by 0.01 for slagle-loop operation as which would produce onset of transition boiling, (MCPR discussed la Reference 2.
The margia between i
of 1.0).
These conditions represent a significant MCPR of 1.0 (osset of transitica boiling) and the departure from the condition intended by design for Safety Limit is derived from a detailed statisti-l P.nned operation.
cal analysis considerlag all of the uncertainties j
in monitoring the core operatlag state includlag A.
Rv&19I_ffen ure >785_p d g_and_Cgre Flow > 101 the uncertalaty la the boillag transition corre-1 QLRated lation as described in Reference 1.
The uncer-
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tanties employed in derivine the Safety Limit are Onset of transition boiling results in a decrease j
in heat transfer from the clad and, therefore, l
f" Aaenam:nt No. 14, }e, 24, 30, 43, M, 94117 l
12
JAFNPP 1.1 (cont'd) provided at the beginning of each fuel cycle.
At 100% power, this limit is reached with maximum Secause the boiling transition correlation is f raction of limiting power density (MFi.PD) equel based on a large quantity of full scale data to 1.0.
In the event of operation with MFLPD there is a very high confidence that operation of greater than the fraction of rated power (FRP),
fuel assembly at the Safety Limit woul.1 not the APRM scram and rod block settings shall be produce boiling transition. Thus, although it is adjusted e4 required in specifications 2.1.A.I.c not res-;1 red to establish the safety limit, and 2.1.A.1.d.
additional margin exists bitween the Safety Limit and the actual occurrence of loss of cladding B.
Core Thermal Power Limit (Reactor Pressure T785 integrity.
psigl However, if boiling transition were to occur, clad At pressures below 785 psig, the core elevation perforation would not be expected. Cladding pressura drop la greater than 4.56 poi for no temperatures would increase to approximately boiling in the bypass region. At low powers and 1100*F which is below the perforation temperature flows, this pressure drop is due to the elevation of the cladding naterial. This has been verified pressure of the bypass region of the core.
by tests in the General Electric Test Reactor Analysis shows that for bundle power la the range (CETR) where fuel similar in design to FitzPatrick of 1-5 Mut, the channel flow will never go below operated above the critical heat flux for a 28 x 103 lb/hr. This flow results from the significent period of time (30 minutes) without pressure differential between the bypass region clad perforation.
and the fuel channel. The pressure differential is primarily a result of changes la the elevation If reactor pressure should ever exceed 1400 psia pressure drop due to the density difference during normal power operation (the limit of between the bolllag water la the fuel ch===m1 and applicability of the boiling transition correla-the non-boiling water in the bypass region. Full tion) it would de assumed that the fuel cladding scale ATLAS test data taken at pressures from 0 integrity Safety Limit has been violated.
to 785 psig indicate that the fuel assembly 3
critical power at 28 x 10 lb/hr is approxi-In addition to the boiling transition limit mately 3.35 MWt.
With the dealga peaking (Safety Limit), operation is constrained to a factors, this corresponds to a core thermal power maximum LHCR of 14.4 KW/ft for GE878EB fuel and of more than 50%.
Thus, a core thermal power 13.4 KW/ft for the remainder.
limit of 25% for reactor pressures below 785 psig is conservative.
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Amendment No. M 21, 30, a3, M, M, Idb117 13 s--.-
JAFNPP 3.1 (CONTIWED)
MCPR Operating Limit for Incremental C.
MCPR shall be determined daily during reactor Cycle Core Average Eggesure power operation at 2_ 25% of rated thermal power and following any change in power level or d.is-
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At kBM Hi-trip BOC to EOC-2CWD/t to EOC-1GWD/t tribution that would cause operation with a
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level _scitint EOC-2GWD/t 59'*-lGWD/t to EOC limiting control rod patterm as described in the l
bases for Specificatloa 3.3.B.5.
J S =.66W + 39%
1.25 1.27 1.30 D.
When it is determined that a cb====1 has failed S =.66W + 40%
1.25 1.27 1.30 in the unsafe condition, the other RPS channels l
that monitor the same variable shall be function-S =.66W + 41%
1.25 1.27 1.30 ally tested immediately before the trip system containing the failure is tripped. The trip S =. 6 6W + 4 2%
1.28 1.28 1.30 system containing the unsafe fall?re may be j
placed in the untripped condition during the S =.66W + 43%
1.33 1.33 1.33 period in which surveillance testing is being performed on the other RPS channels.
S =.66W + 44%
1.33 1.33 1.33 E.
Verification of the limits set forth in speci-2.
If requirement 4.1.E.1 is not met (i.e. T 4 TAyg) fication 3.1.B shall be performed as follows:
g then the Operating Limit MCPR values (as a func-l tion of T ) is as given in Figure 3.1-2.
1.
The average scram time to notch position 38 I
shall be Where T=
(T AVE - BI/I A-B)
AVE B
and T AVE =
the average scram time to notch 2.
The average scram time to notch position 38 position 38 as defined in speci-is determined as follows:
fication 4.1.E.2,
'E the adjusted analysis mean scram
=
B a
a time as defined in specification N1 1 i Mi 4.1.E.3, 7
=
the scram time to notch position AVE
=
A 38 as defined in specification i=1 I=1 3.3.C.1 where: a = number of surveillance tests performed to date in the cycle, Ni = number of active rods measured in Amendment. No. p'4, 74, H, $6, 96, lob i 17 31
JAFMPP IABT.E 3.1-1 (c~ t*dl REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMNT HQIES OF TABLE 3.1-1 (cont *d) 14.
The APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds l
directly to inst ~taneous neutron flux.
15.
This Average Power Range Monitor scraa4 function is fixed pokat and is increased when the reactor mode switch is place in the Run position.
16.* During the proposed Hydrogen Additton Test, the normal background radiation level will increase by approximately a factor of 5 for peak hydrogen concentration. Therefore, prior to performance of the test, the Main Steam Line Radiar.lon Monit or Trip Level Setpoint will be raised to fthree times the lacreased radiation levels. The test will be conducted at power levels >80% of normal rated power. During controlled power reduction, the setpoint will be readjusted prior to going below 20% rated power without the setpoint change, control rod withdrawal will be prohibited until the necessary trip setpoint adjustment is made.
17.
This APRM Flow Referenced Scram setting is applicable to two loop operation. For one loop operation this setting becomes S i (C,66W.54%-0.66dW)(FRP/MFLPD) where dW = Difference between two-loop and single-loop effective drive flow at the same core flow.
- This specification is in effect only during Operating Cycle 7.
sf,<c, M. I17 Amend.aent No. y3, 43a
e.
JAFNFP Figure 3.3-1 K
VAmin g
5,4--
. _. = - -
1.3 1.2 AUTOMATIC FLOW CONTROL N
1."1 MANUAL FLOW CONTMOL SCOOP TUGE SETPOtNT CAUSAATION PCSITION'D SUCH THAT FLOWMAX = 102.5%
l 107.0%
I 112.0 %
j 1.0 117.0 %
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30 40 50 60 70 80 90 100 CORT FLOW (%)
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AmOndment No. g.117 l
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JAFNPP Figure 3.1-2 oneratina Limit Mc71 versus T (defined in Section 3.1.B.2)
FOR ALL FtTEL TYPES 1.35*
~1.35
- 1.34 Operating Limit EOC
-1.32 MCPR 1.30-
-1.30 EOC-1 1.27 -
1.25-EOC-2
-1.25 1.23 -
-1.23 I
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O 0.2 0.4 U.6 0.8 1
5 Am:ndment No. 64, 74, 74, S(, W 117 47b
JAFNPP 3.5 (cont *d) 4.5 (cont'd) condition, that pump shall be considered inoper-2.
Following any period where the LPCI subsys-able for purposes satisfying Specifications tems or core spray subsystems have not been 3.5.A, 3.5.C, and 3.5.E.
required to be operable, the discharge piping of the inoperable system shall be H.
hygfage Planar Llegar Hs t Generation Rate vented from the high polat prior to the (APLHGR) return of the system to service.
During power operation, the APLHGR for each type 3.
Whenever the NPCI, RCIC, or Core Spray System of fuel as a functica of axial location and is lined up to take auction from the condom-average planar exposure shall be within limits sate storage tank, the discharge piping of based on applicable APLHG3 limit values which the HPCI, RCIC, and Core Spray shall be have been approved for the respective fuel and vented from the high point of the system, lattice types. When hand calculations are and water flow observed on a monthly basis, required, the APLHGR for each type of fuel as a function of average planar exposure shall not 4.
The level switches located on the Core Spray exceed the limiting value for the most limiting and RHR System discharge piping high polats lattice (excluding natural uranium) shown in which monitor these lines to insure they are l
Figures 3.5-11 through 3.5-14 during two full shall be functionally tested each month.
recirculation loop operation. During single loop operation, the APLHGR for each fuel type shall H.
Averaae Planar Linear Heat Generation Rate not exceed the above values multiplied by 0.84 (APLHGR)
(see Bases 3.5.K, Reference 1).
If anytime during reactor power operation greater than 25%
The APLHGR for each type of fuel as a function of of rated power it is determined that the limiting average planar exposure shall be determlaed asily value for APLHGR la being exceeded, action shall during reactor operation at 1 25% rated thermal then be initiated within 15 minutes to restore power.
operation to within the prescribed limits.
If the APLHGR is not returned to within the prescribed limits within two (2) hours, an orderly reactor power reduction shall be commenced immediately. The reactor power shall be reduced to less than 25% of rated power within the next four hours, or until the APLHGR is returaed to within the prescribed limits.
Amendment No. M, M, 14, SK, SK, 189.117 123
f JAFMPP 3.5 BASES (cont'd) requirements for the emergency diesel generators.
are within the 10 CFR 50 Appendix K limit. The limiting values for APLMGR are given la Figures G.
Maintenanct_.gf Filled Dissharge Pipe 3.5-11 through 3.5-14.
Approved limitlag values
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of APLHGR as a function of fuel type are given in If the dischstge piping of the core spray, LPCI, NEDO-21662-2 (as amended) for Reload 6 fuel.
l RCIC, and HPCI are not filled, a water haarAr can Approved limiting values of APLEGE'as a function develop in this piping when the pump (s) are of fuel and lattice types are gives la NEDC-started. To minimize damage to the discharge 31317P (as amended) for Reload 7 and 8 feel.
l piping and to ensure added margin in the opera-These values are multiplied by 0.54 during Single tion of these systems, this technical specifica-Loop Operation. The derivation of this multi-tion requires the discharge lines to be filled plier can be found in Bases 3.5.K, Reference 1.
whenever the system is required to be operable.
If a discharge pipe is not filled, the pumps the I.
Linear Heat Generation Rate (LEGR) supply that line must be assumed to be inoperable for technical specification purposes. However, This specification assures that the linear heat if a water hm==*r were to occur, the system would generation rate la any rod is less than the still perform its design function.
design linear heat generatloa.
1 H.
hyttoge PlanaI_Lincar Heat Geogration Rata The LHGR shall be checked daily during reactor IAELHGR1 operatloc at 25% rated thermal power to deter-mine if fuel burnup, or control rod movement, haa This sp9cification assures that the peak cladding caused changes la power distributloa. For LEGR temperature following the postulated design basis to be a limitiLg value below 25% rated thermal loss-of-coolant accident will not exceed the limit power, the ratio of local IRGE to average IJIGR specified in 10 CFR 50 Appendix K.
would have to be greater th==
10 which la pre-cluded by a considerable margia when employing The peak cladding temperature following a postu-any permissible control rod pattera.
lated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.
Since espected local variations in power distri-bution within a fuel assembly affect the calcu-lated peak clad temperature by less than 1 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated terceratures Amembaent No. % 7f, 8f, SK, J99. II7 4
130 4
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JAFNPP Figure 3.5-13 Maximum Averaae Planar Linear Heat Generation Rate (Mapre n)
Versus Averaae Planar Rwnosure 14-Reload 8, BD336A 13-
Reference:
NEDC-31317P o27 x o 12-u Mk E U 11-O
$3 10-c a.
- 2 9-
.e O' #
s-
.>c a t 7-au XC 6-J l
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5 10 15 20 25 30 35 40 45 50 Planar Average Exposure (GWd/St)
For single-loop operation, these This curve represents the limiting MAPIJIGR values are multiplied by 0.84.
exposure dependent MAPIJIGR values..
Amendment No.117 135k a
JAFNPP Figure 3.5-14 Maximum Averaae Planar Linear Heat Generation Rate (MAPLMcD)
Versus Averaae Planar Rrnosure 14-ReloaC 8, BD339A 13-
Reference:
NEDC-31317P oem 241-u N 35 E U 11-
- ix 0
g3 10-cm NI A
9-ee av ee hM 8-
$c do g y 7-au Y$
2 3 l
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5 10 15 20 25 30 35 40 45 50 Planar Average Exposure (GWd/St) l For single-loop operation, these This curve represents the limiting MAPLHGR values are multiplied by 0.84.
exposure dependent MAPLHGR values.
Amendment No.Il7
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l 1351 l
JAFMPP 5.0 DESIGN FEATURES B. The reactor core contains 137 cruciform-shaped control rods as described la Section 3.4 of 5.1 SIIg
- he FSAR.
A. The' James A. FitzPatrick Nuclear Power Plant 5.3 REACTOR PEESSURE VESSEL is located ca the PASNY portion of the Nine Mile Point rite, approximately 3,000 ft. east The reactor pressure vessel is as described la of the Nine Mile Polat Nuclear Station, Unit Table 4.2-1 and 4.2-2 of the FSAR. The applicable 1.
W NPP-JAF site is on Lake Ontario la design codes are described la Section 4.2 of the Oswego Country, New York, approximately 7 FSAR.
miles northeast of Oswego. The plant is located at coordinates morth 4,819, 543.012 m, 5.4 CONTAIMENT east 386, 968.945 m, on the Universal Transverse Mercator System.
I The principal des Qa parameters and charac-teristics for the primary containment are B. The nearest point on the property line from given la Table 5.2-1 of the FSAR.
the reactor building and any points of poten-tial gaseous of fluents, with the exception of B. The secondary containment is as described la the lake shoreline, is located at the north-Section 5.3 and the applicable codes are as east corner of the property. This distance is described in Section 12.4 of the FSAR.
approximately 3,200 ft. and is the radius of
- the exclusion areas as defined in 10 CFR 100.3.
C. Penetrations of the primary conta1 ament and piping passing through such penetrations are 5.2 EEACIOR designed la accordance with standards set forth la Section 5.2 of the FSAR.
A. The reactor core consists of not more than 560 l
tuel assemblies. For the current cycle, three 5.5 FUEL STORAGE fuel types are present la the core: BPSX8R, GESX8EB, and QUADe. The GE fuel types are A. The new fuel storage facility design criteria described la NEDO-24011. The BP8X8R fuel type are to malatala a I,gg dry 40.90 and has 62 fuel rods and 2 water rods and the flooded
<0.95.
Compliance shall be verified GESISEB fuel type has 60 fuel rods and 4 water prior to introduction of any new fuel design rods. The QUAD + fuel type is described in to this facility.
WCAP-11159 and has 64 fuel rods.
Amendment No. M, 4 7, M, 6(, 66, 74, IDT II7 245 4