ML20212E221
| ML20212E221 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/23/1986 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML19292G521 | List: |
| References | |
| NUDOCS 8701050277 | |
| Download: ML20212E221 (29) | |
Text
...
ATTACHMENT I TO JPN-86 63 PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING RELOAD 7/ CYCLE 8 (JPTS-86-023)
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 Bh10gjh h P P
_ _ . _ - _ _ _~_ _ . _ - _ - - -_
JAF!PP LIST OF FIGURES e
Fiaure Title a Lage_
3.1-1 Manual Flow Control 47a 3.1-2 Operating Limit MCPR versus 1[' 47b l
4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Tests 48 4.2-1 Test Interval vs. Probability of System Unavailability 87 3.4-1 Sodium Pentaborate Solution of System Volume-Concentration Requirements 110 3.4-2 Saturation Temperature of Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Limits of Specifications 3.5.J.1, 3.5.J.2 and 3.5.J.3 134 l
3.5-6 (Deleted) 135d 3.5-7 (Deleted) 135e 3.5-8 (Deleted) 135f 3.5-9 (Deleted) 135g 3.5-10 (Deleted) 135h 3.5-11 (Deleted) 1351 3.6-1 Reactor Vessel Thermal Pressurization Limitations 163 4.6-1 Chloride Stress Corrosion Test Results at 500*F 164 6.1-1 Management Organization Chart 259 6.2-1 Plant Staff Organization 260 Amendment No. )4, / 3 , g, 4(, Pl J4,JHI, 98 vil a
JAFMPP surveillance tests, checks, calibrations, and V. Electrically Disarmed Control Rod examinations shall be performed within the '
specified surveillance intervals. These inter- To disarm a rod drive electrically, the four j vals may be adjusted + 25 percent. The interval amphenol type plus connectors are removed from as pertaining to instrument and electric surveil- the drive insert and withdrawal solenoids ren-l lance shall never exceed one operating cycle. In dering the rod incapable of withdrawal. This
- cases where the elapsed interval has exceeded 100 procedure is equivalent to valving out the drive j percent of the specified interval, the next sur- and is preferred. Electrical disarming does not veillance interval shall conunence at the end of eliminate position indication. i
! the original specified interval.
j W. Hith Pressure Water Fire Protection System j U. Thermal Parameters The High Pressure Water Fire Protection System j 1 1. Minimum critical power ratio (MCPR)-Ratio of consists of: a water source and pumps; and that power in a fuel assembly which is distribution system piping with associated post j calculated to cause some point in that fuel indicator valves (isolation valves). Such valves assembly to experience boiling transition to include the yard hydrant curb valves and the
- the actual assembly operating power as first valve ahead of the water flow alarm device calculated by application of the GEIL on each sprinkler or water spray subsystem,
- correlation (Reference NEDE-10958).
X. Stammered Test Basis
- 2. Fraction of Limiting Power Density - The
] ratio of the linear heat generation rate A Staggered Test Basis shall consist of:
i (LHGR) existing at a given location to the design LHGR. The design LHGR is 14.4 KW/ft a. A test schedule for a systems, subsystems, f trains or other designated components ob-l for GE8x8EB fuel and 13.4 KW/ft for the a remainder. tained by dividing the specified test
! interval into n equal subintervals.
- 3. Maximum Fraction of Limiting Power Density - ~
j The Maximum Fraction of Limiting Power b. The testing of one system, subsystem, train
! Density (MFLPD) is the highest value or other designated component at the begin-3 existing in the core of the Fraction of ning of each subinterval.
j Limiting Power Density (FLPD).
1 Y. Rated Recirculation Flow l 4. Transition Boiling -
Transition boiling means the boiling region between nucleate That drive flow which produces a core flow of I and film boiling. Transition boiling is the 77.0 x 106 lb/hr.
f region in which both nucleate and film i boiling occur intermittently with neither l type being completely stable.
l Amendment No. 36', 64, Pf, J(
' JAFMPP ,
i 1.1 (cont'd) 2.1 (cont'd)
Reactor Water Level (Hot or Cold Shutdown In the event of operation with a maximum fraction I D.
Conditions) of limiting power density (NFLPD) greater than the fraction of rated power (FRP), the setting Whenever the reactor is in the shutdown condition shall be modified as follows:
i with irradiated fuel in the reactor vessel, the water level shall not be less than that corres- S 4 (0.66 W + 54%) x (FRP/MFLPD) ponding to 18 inches above the Top of Active Fuel when it is seated in the core. for two loop operation or, i
1 S d (0.66 W + 54% - 0.66 W)(FRP/MFLPD) for single loop operation Where:
FRP = fraction of rated thermal power (2436 MWt) 1 MFLPD = maximum fraction of limiting power j density where the limiting power l density is 14.4 KW/ft for GE8X8EB fuel and 13.4 KW/ft for the f remainder.
f 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 j
actual operating value will be used.
2 (2) Fixed fligh Neutron Flux Scram Trip Setting When the Mode Switch is in the RUN position, i
' the APRM fixed high flux scram trip setting shall be:
l
) S # 120% Power 4
4 I
Amendment No. }/, 36,46,$4,34,%
9 i
JAFNPP 2.1 (cont'd)
NFLPD = maximum fraction of limiting power density where the limiting power density is 14.4 KW/ft for GE8X8EB fuel and 13.4 KW/ft for the remainder.
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.
Amendment No. [
10a
i . .
1 ,o l JAFNPP ^
1.1 (cont'd) provided at the beginning of each fuel cycle. At 100% power, this limit is reached with a maxi-Because the boiling transition correlation is mum fraction of limiting power density (NFLPD) based on a large quantity of full scale data equal to 1.0. In the event of operation with a there is a very high confidence that operation of MFLPD greater than the fraction of rated power
- fuel assembly at the Safety Limit would not (FRP), the APRM scram and rod block settings 1 produce boiling transition. Thus, although it is shall be adjusted as required in specifications
! not required to establish the safety limit, 2.1.A.1.c and 2.1.A.1.d.
i additional margin exists between the Safety Limit and the actual occurrence of loss of cladding B. Core Thermal Power Limit (Reactor Pressure < 785 integrity. psis)
However, if boiling transition were to occur, clad At pressures below 785 psig the core elevation 4 perforation would not be expected. Cladding pressure drop (0 power, O flow) is greater than temperatures would increase to approximately 4.56 psi. At low powers and flows this pressure 1100*F which is below the perforation temperature differential is maintained in the bypass region of the cladding material. This has been verified of the core. Since the pressure drop in the by-by tests in the General Electric Test Reactor pass region is essentially all elevation head.
(CETR) where fuel similar in design to FitzPatrick the core pressure drop at low powers and flows operated above the critical heat flux for a will always by greater than 4.56 psi. Analyses show that with a flow of 28 x 103 lbs/hr bundle
- significant period of time (30 minutes) without i clad perforation. flow, bundle pressure drop is nearly independent i of bundle power and has a value of 3.5 psi. Thus, If reactor pressure should ever exceed 1400 psia the bundle flow with a 4.56 psi driving head will
! during normal power operation (the limit of be greater than 28 x 103 lbs/hr. Full scall
- applicability of the boiling transition correla- ATLAS test data taken at pressures from 0 psig to tion) it would be assumed that the fuel cladding 785 psig indicate that the fuel assembly critical
- integrity Safety Limit has been violated. power at this flow is approximately 3.35 MWt.
] With the design peaking factors this corresponds
- In addition to the boiling transition limit to a core thermal power of more than 50%. Thus, a l (Safety Limit), operation is constrained to a core thermal power limit of 25% for reactor pres-maximum LHGR of 14.4 KW/ft for GE8X8EB fuel and sures below 785 psig is conservative.
13.4 KW/ft for the remainder.
4 i
l l
i i
i 4 Amendment No. 14,fi, pd, W, y, 74
- 13 l
a i
i JAFNPP 4.1 SrRVEILLANCE REQUIREMENTS 3.1 LIMITING CONDITIONS FOR OPERATION 4.1 REACTOR PROTECTION SYSTEM 3.1 REACTOR PROTECTION SYSTEM Applicability: .
Applicability:
Applies to the instrumentation and associated devices Applies to the surveillance of the instrumentation which initiate the reactor scram, and associated devices which initiate reactor scram.
Objective:
Ob3cetive:
To essure the operability of the Reactor Protection To specify the type of frequency of surveillance to be applied to the protection instrumentation.
System.
Specification:
Specification:
A. Instrumentation systems shall be functionally A. The setpoints, minimum number of trip systems, tested and calibrated as indicated in Tables minimum number of instrument channels that must 4.1-1 and 4.1-2 respectively, be operable for each position of the reactor mode switch shall be as shown on Table 3.1-1. The -
design system response time from the opening of i the sensor contact to and including the opening of the trip actuator contacts shall not exceed 50 msec.
1 B. Maximum Fraction of Limitina Power Density (MFLPD)
. B. Minimum Critical Power Ratio (MCPR)
The MFLPD shall be determined daily during power operation, the MCPR l l During reactor reactor power operation at },_ 25% rated thermai l operating limit shall not be less than that shown power and the APRM high flux scram and Rod Block 1
below: trip settings adjusted if necessary as2.1.A.1.d, required by Specifications 2.1.A.1.c and 1
respectively.
- 1. When surveillance requirement 4.1.E is met.
(T AVE I IB) 1
! Amendment F3. #4, M, FI 3Of
JAFNPP 3.1 (CONTINUED)
MCPR Operating Limit for Incremental C. MCPR shall be determined daily during reactor Cycle core Average Exposure power operation at 2_ 25% of rated thermal power and following any change in power level or dis-4t RBM Hi-trip BOC to EOC-2GWD/t to EOC-1GWD/t tribution that would cause operation with a level setting EOC-2GWD/t EOC-1GWD/t to EOC limiting control rod pattern as described in the bases for Specification 3.3.B.S.
S= .66W + 39% 1.23 1.29 1.30 D. When it is determined that a channel has failed S= .66W + 40% 1.23 1.29 1.30 in the unsafe condition, the other RPS channels that monitor the same variable shall be function-S= .66W + 41% 1.23 1.29 1.30 ally tested inanediately before the trip system containing the failure is tripped. The trip S= .66W + 42% 1.27 1.29 1.30 system containing the unsafe failure may be placed in the untripped condition during the S= .66W + 43% 1.33 1.33 1.33 period in which surveilla..re 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.Tg47 AVE) fication 3.1.B shall be performed as follows:
then the Operating Limit MCPR values (as a func-tion of T ) is as given in Figure 3.1-2. 1. The average scram time to notch position 38 shall be:
Where T= (TAVE - b )/C A- B) AVE I B and TAVE = 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, B = the adjusted analysis mean scram time as defined in specification 4.1.E.3, y = Ni Ui Ni A = the scram time to notch position AVE 38 as defined in specification i=1 i=1 3.3.C.1 where: n = number of surveillance tests performed to date in the cycle, Ni = number of active rods measured in Amendment No. / , g, g , [, j[
31
- Note: Should the operating limit MCPR the ith surveillance and i = sverage obtained from this figure be less scram time to notch position 38 of all rods i than the operating limit MCPR found measured in the ith surveillance test.
l in Specification 3.1.B.1 for the applicable RBM trip level setting 3. The adjusted analysis mean scra's time is l then Specification 3.1.B.1 shall calculated as follows:
I apply.
l N1
- 3. During single loop operation, the operating limit MCPR shall be increased by 0.01 as specified in T
]
Specification 3.1.B.1 or 3.1.B.2 to reflect the B (sec) = [ +1.65 7 n
- increase in safety lim"it MCPR. (See Ni Specification 1.1.A).
i=1
- 4. During Reactor power operation with core flow less than 100% of rated, the MCPR operating limit shall be multiplied by the appropriate Kg as whereg = mean of the distribution for the j shown in Figure 3.1.1. average scram insertion time to
- the pickup of notch position 38 =
l 5. If anytime during reactor operation at greater 0.706 sec.
'I than 25% of rated power it is determined that the ,
limiting value for MCPR is being exceeded, action r = standard deviation of the distri-shall then be initiated within fifteen (15) bution for average scram j minutes to restore operation to within the insertion time to the pickup of l prescribed limits. If the MCPR is not returned notch position 38 = 0.016 sec.
to within the prescribed limits within two (2) hours, an orderly reactor power reduction shall Ni= the total number of active rods i begin immediately. The reactor power shall be measured in specification 4.3.C.1 l reduced to less than 25% of rated power within I the next four hours, or until the MCPR is The number of rods to be scram tested and the l returned to within the prescribed limits. test intervals are given in Specification j 4.3.C.
i ,
1 i
l I
1 Amendment No. M , M , J4
{ 31a
JAFNPP 3.1 BASES (cont'd)
Turbine control valves fast closure initiates a scram based on pressure switches sensing electro-hydraulic control (EHC) system oil pressure. The switches are located between fast closure solenoids and the disc dump valves, and are set relative (500 < P( 850 psig) to the normal (EHC) oil pressure of 1.600 psig so that based on the small system volume, they can rapidly detect valve closure or loss of hydraulic pressure.
The requirement that the IRM's be inserted in the core when the APRM's read 2.5 indicated on the scale in the start-up and refuel modes assures that there is proper overlap in the neutron monitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation. .
B. .The limiting transient which determines the required steady state MCPR limit depends on cycle exposure. The operating limit MCPR values as determined from the transient analysis in the current reload submittal for various core exposures are given in Specification 3.1.B.
The ECCS performance analyses assumed reactor operation will be limited to MCPR = 1.20, as described in NEDO-21662 and NEDC-31317P. The l Technical Specifications limit operation of the reactor to the more conservative MCPR based on consideration of the limiting transient as given in Specification 3.1.B.
Amendment No. W, f(
35
m JAFNPP TABLE 3.1-1 (cant'd) d IEACIDR PRCTfffrIm SYSIB4 (SCRAM) INSTRLMNTATIN REQUIRIMNr Ybtes of "able 3.1-1 (mnt'd)
C. Iligh Flux IIN D.
Scram Discharge Volume Ifigh Iavel when any control rod in a control cell containing fuel is not fully inserted.
E. APIM 15% Ibwer Trip
- 7. Not required to bu gerable when primary cantainment integrity is not required.
- 8. Tbt required to be operable when the reactor pressure vessel head is not bolted to the vessel.
- 9. %e APIN (kwnscale trip is automatically bypassed when the IRM Instrumentation is operable and not hich.
- 10. normal An APIN will be considered operable if there are at least 2 LPIN inputs per level and at least 11 LPIN inputs of the complement.
- 11. See Section 2.1.A.1. '
- 12. %is equation will be used in the event of operation with a maximum fraction of limiting power density (MELPD) greater than the fraction of rated power (FRP).
There: FRP =
Fraction of Rated hermal Power (2436 Ptit)
PIFLPD =
Maximum Fraction of Limiting Pbwer Density where the limiting power density is 14.4 IGi/f t for GEBX8EB fuel and 13.410f/f t for the remainder.
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= Icop Recirculation Flow in per nt of rated.
S= Scran Setting in percent of rated thermal power.
- 13. The Average Ibwer Range Pbnitor scram function is varied as a function of recirculation flow (W). The trip setting of this function must be maintained in accordance with Specification 2.1.A.1.c.
Amendment Ib. f9, f#, K F/, 68f, pf, K
r 1 l
Figure 3.1-2 Operatina Limit MCPR Versus T (Defined in Section 3.1.B.2)
FOR ALL FUEL TYPES 1,40 -
- 1.36 1.35 -
Operating E0C Limit MCPR 1.30 -
1.29 - L E0C-1 - 1*30 1.27 1.25 -
- '
- 7) 1.23 -
1.20 -
0 0l2 0l4 0!6 0!8 I
Amendment No. d f , g, g, y ,
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. Averar.n Planar Linear Heat Generation Rate vented from the high point prior to the (APLHCR) return of the system to service.
The APLHCR for each type of fuel as a function of 3. Whenever the HPCI, RCIC, or Core Spray System average planar exposure shall not exceed the is lined up to take suction from the conden-limiting values given in NEDO-21662-2 and sate storage tank, the discharge piping of NEDC-31317P for two loop operation. For single the NPCI, RCIC, and Core Spray shall be loop operation these values are reduced by vented from the high point of the system.
l multiplying by 0.84 (see Bases 3.5.K, Reference and water flow observed on a monthly basis.
1). If anytime during reactor power operation greater than 25% of rated power it is determined 4. The level switches located on the Core Spray that the limiting value for APLHGR is being and RHR System discharge piping high points exceeded, action shall then be initiated within which monitor these lines to insure they are 15 minutes to restore operation to within the full shall be functionally tested each month, prescribed limits. If the APLHGR is not returned -
to within the prescribed limits within two (2) H. Average Planar Linear Heat Generation Rate hours, an orderly reactor power reduction shall (APLHCR) be commenced immediately. The reactor power shall be reduced to less than 25% or rated power The APLHCR for each type of fuel as a function of within the next four hours, or until the APLHGR average planar exposure shall be determined daily is returned to within the prescribed limits, during reactor operation at 225% rated thermal power.
Amendment No.d3 , k, eke N 191
f.: * . :- y
'y .- ..
- E- .._
+
f j JAFNPP 4.5 (cont'd) -
3.5 (cont'd)
I. Linear Heat Generation Rate (LHCR) [
- 1. Linear Heat Generation Rate (LHGR)
The LHGR shall be checked daily during reactcr j The linear heat generation rate (LHGR) of any rod operation at 2,25% rated thermal power.
in any fuel assembly at any axial location shall ,
not exceed the maximum allowable LHGR of 14.4KW/ft for GE8x8EB fuel and 13.4 kW/ft for the remainder of the fuel.
It anyt.ime during reactor power op$ ration greater than 25% of rated power it is determined that the limiting value tor LHGR is being exceeded, action shall then be initiated within 15 minutes to re-store operation to within the prescribed limits.
If the LHGR is not returned to within the pre-scribed limits within two (2) hours, an ordarly reactor power reduction shall be conunenced immed- -
iately. The reactor power shall be reduced to ,
less than 25% of rated power within the next four hours, or until the LHGR is returned to within the prescribed limits.
Amendment No. 4g, pf, g 124
JAFNPP 3.5 BASES (cont'd) requirements for the emergency diesel generators. are w! thin the 10 CFR 50 Appendix K limit. The limiting values for APLHCR are given in NEDO G. Maintenance of Filled Discharge Pipe 21662-2 and NEDC-31317P. These values are multiplied by 0.84 during Single Loop Operation.
If the discharge piping of the core spray, LPCI, The derivation of this multiplier can be found in RCIC, and PPCI are not filled, a water hanumer can Bases 3.5.K Reference 1.
develop in this piping when the pump (s) are started. To minimize damage to the discharge I. Linear Heat Generation Rate (LHGR) piping and to ensure added margin in the opera-tion of these systems, this technical specifica- This specification assures that the linear heat tion requires the discharge lines to be filled generation rate in any rod is less than the whenever the nystem is required to be operable. design linear heat generation.
If a discharge pipe is not filled, the pumps the supply that line must be assumed to be inoperable The LHCR shall be checked daily during reactor for technical specification purposes. However, operation at 25% rated thermal power to deter-if a water hammer were to occur, the system would mine if fuel burnup, or control rod movement, has still perform its design function. caused changes in power distribution. For LHGR to be a limiting value below 25% rated thermal i H. Averar.e Planar Linear Heat Generation Rate power, the ratio of local LHGR to average LHCR '
(APLHGR) would have to be greater than 10 which is pre-cluded by a considerable margin when employing This specification assures that the peak cladding any permissible control rod pattern.
temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50 Appendix K.
The peak cladding temperature following a postulated 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 expected local variations in power distri-bution within a fuel assembly affect the calcu-lated peak clad temperature by less than i 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 temperatures Amendment No. (4, M,JHI, JT 130 '
JAFNPP Figure 3.5-9 (This page is intentionally blank.)
Amendment No. p4, 96 135g
1
, 9
)
k n
a l
? b y
l l
0 a 1 n
- o 5 i
. t P 3 n h P e 5 N e t 3 F r n 1 A u i J g i s F i e
g a
p s
i h
T
(
g
/7
/ .
o N
t n
a m
d n
e m
A
< l l1f
)
k n
a
, l b
y l
l 1 a 1 n
- o 5 i
. t 1 P 3 n 5 P e 3 N e t 1 F r n A u i J g i s F i e
g a
p s
i h
T
(
M ,
o N
t n
e m
d n
e m
A f 1 lll l l
- 0. The rametne cere contains 137 crucifors-shaped 5.0 DESIGN FEATURES control rods as 4 ascribed in Section 3.4 of 5.1 SITE the FSAR.
A. The James A. FitzPatrick Nuclear Power Plant 5.3 REACTOR PRESSURE VESSEL is located on the PASNY portion of the Nine Mile Point site, approximately 3,000 ft. east The reactor pressure vessel is as described in Table 4.2-1 and 4.2-2 of the FSAR. The applicable of the Nine Mile Point Nuclear Station. Unit design codes are described in Section 4.2 of the
- 1. The NPP-JAF site is on Lake Ontario in Oswego Country. New York, approximately 7 FSAR.
miles northeast of Oswego. The plant is I located at coordinates north 4.819, 545.012 m, 5.4 CONTAINMENT
! east 386, 968.945 m, on the Universal Transverse Mercator System. A. The principal design parameters and charac-teristics for the primary containment are B. The nearest point on the property line from given in Table 5.2-1 of the FSAR.
the reactor building and any points of poten-i tial gaseous effluents, with the exception of B. The secondary containment is as described in the lake shoreline, is located at the north- Section 5.3 and the applicable codes are as j
east corner of the property. This distance is described in Section 12.4 of the FSAR.
q approximately 3,200 ft. and is the radius of C. Penetrations of the primary containment and-the exclusion areas as detined in 10 CFR 100.3.
j piping passing through such penetrations are j 5.2 REACTOR designed in accordance with standards set forth in Section 5.2 of the FSAR.
A. The reactor core consists of not more than 560 fuel assemblies. For the current cycle, four 5.5 FUEL STORAGE l fuel types are present in the core: P8X8R, BP8X8R, GE8X8EB, and QUAD +. The GE fuel types A. The new fuel storage facility design criteria j are described in NEDO-24011. Both P8X8R and are to maintain a K,gg dry ( 0.90 and
' BP8X8R fuel types have 62 fuel rods and 2 flooded ( 0.95. Compliance shall be verified water rods and GE8X8EB fuel type has 60 fuel prior to introduction of any new fuel design l
rods and 4 water rods. The QUAD + fuel type is to this facility, described in WCAP-11159 and has 64 fuel rods.
t i
i l
I I
l Amendment No. Jd, 4d, $s, N , $(, /
245
ATTACHMENT II TO JPN-86-63 SAFETY EVALUATION FOR PROPOSED TECHNICAL EEECIFICATION CHANGES REGARDING RELOAD 7/ CYCLE 8 (JPTS-86-023) s NEW YORK POWER AUTHORITY JAMEU A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
. , , , , , . -y__ _,,- - - - - -- - . . _ _ _ , - , , _ , - - -, . . , , . _ _ . , - _ . . - . _ . , , _ _ - . . . _ , . . _, _m.-..-.. --- -.._ --._ ,.
i l
, I. DESCRIPTION OF THE PROPOSED CHANGES On page vii, references to figures 3.5.9 through 3.5.11 have been deleted.
Also on page vii, changes are made to the entries for Figures 3.1-2 and 3.5-1 to add text inadvertantly missing from the Authority amendment request submittal, approved and issued by the NRC as Amendment 98. These changes involve inserting the Greek letter " 7;", and the reference to specification 3.5.J.3.
On page 6, Section 1.0.U.2 is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB fuel loaded as Reload 7, and that the LHGR limit remains 13.4 KW/ft for the remainder of the
. fuel.
, On page 9, Section 2.1.A.1.c, the definition for MFLPD is
. changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
On page loa, Section 2.1.A.1.d, the definition for MFLPD is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.-
On page 13, the BASES for Section 1.1.A is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
On pages 30f through 31a, Section 3.1.B has been revised and reorganized. Minor changes are made to the text to correct typographical errors and for clarification.
l On page 31, Section 3.1.B.1, the "MCPR Operating Limit for
! Incremental Cycle Core Average Exposure" table has been revised to reflect the transient analyses performed for the Reload 7/ Cycle 8 core (Reference 3).
On page 31a, Section 4.1.E.3, the values of,g/ and r are revised from 0.723 and 0.054 to 0.706 and 0.016 repectively in accordance with the accepted procedure as documented in Reference 5.
On page 35, Bases 3.1.B, the reference for the Cycle 8 Loss-Of-Coolant-Accident analyses is added.
On page 43, Notes for Table 3.1-1, Item 12, the definition for MFLPD is changed to reflect the 14.4 Kw/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel. Also in this note, the definition of "S" is changed from " percent of initial" to " Percent of rated thermal power.
- ,,.-.-e- -----..v.-..m--- - - , - . - - , , . . , , , - . , . - - . . . - , _ , , , - . . , . - - - . , - , - , - , - , , , , . . - - - - - - . - . . - - - - - - - - ,-- --
On page 47b, the " Operating Limit MCPR Versus l$ (Defined in Section 3.1.B.2) for All Fuel Types" graph is revised to reflect the transient analyses performed for the Reload 7/ Cycle 8 core (Reference 3).
On pages 123, and 130 references to figures 3.5-9 through 3.5-11 have been deleted and are replaced with the appropriate General Electric Loss-of-Coolant-Accident analysis reports.
On page 124, Section 3.5.I is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
Pages 135g, 135h, 135i have been deleted.
On page 245, Section 5.2 is changed to reflect the current a fuel designs used in the FitzPatrick core.
II. PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed changes is to support plant start-up and operation after the Reload 7/ Cycle 8 refueling outage. During this outage, 188 fuel bundles are to be removed from the reactor core and replaced with new fuel. As part of a program to qualify Westinghouse Nuclear Energy Systems as a vendor of nuclear fuel for FitzPatrick, four (4) demonstration lead test assemblies of their QUAD + design are included in Reload 7.
The changes to the Technical Specifications involve deleting specifications associated with the discharged fuel and with
- Cycle 7 specific analyses. These are replaced with ones appropriate for the new fuel and based on cycle 8 specific analyses.
The changes to page vii are purely administrative. The " List of Figures" is updated to reflect changes made as part of this proposed change. Other changes correct existing administrative errors on this page.
The changes to pages 6, 9, 10a, 13, 43, and 124 reflect a new LHGR limit of 14.4 KW/ft for the fuel type GE8X8EB added as Reload 7 and as accepted and documented in Reference 5.
The reorganization and other administrative changes to pages 30f through 31a have been made to facilitate use of Section 3.1.B.
The changes to the "MCPR Operating Limit for Incremental Cycle Core Average Exposure" table in Section 3.1 on page 31 and the " Operating Limit MCPR Verses 'd.(Defined in Section 3.1.B.2)
For All Fuel Types" graph on page 47b reflect cycle specific transient analyses performed by General Electric for the Reload 7/ Cycle 8 core (Reference 3). The results of these analyses are included in this application as Attachment III.
Changes to the values of// and cP on page 31a are made to reflect the use of the GEMINI transient methods for calculating the transient 21CPR. This General Electric methodology has been approved by the NRC and is documented in Reference 5. General Electric used this methodology in their Cycle 8 specific transient analyses for FitzPatrick. (Reference 3).
i The changes to page 35 add the reference to the Cycle 8 specific Loss-of-Coolant-Accident analysis.
The changes to pages 123 and 130 relect the deletion of references to the MAPLHGR curves, Figures 3.5-9 through 3.5-11, and the addition of references to LOCA analysis reports. These reports provide the MAPLHGR limits for fuel in the FitzPatrick Core.
MAPLHGR limits for the QUAD + demonstration aassemblies will be those applicable to the Reload 6 BP8DRB299 bundles following the practice approved by the NRC in Reference 7 in which the licensee applied the MAPLEGR limits for the General Electric assembly to the matched QUAD + assemblies.
MAPLHGR curves on pages 135g through 135i have been deleted to reflect the practice of using the original source material, specifically the LOCA analysis reports NEDO-21662-2 and NEDC-31317P, to obtain limiting MAPLHGR values for surveillance.
The introduction of GE8X8EB fuel, which contains several lattice types of varying gadolinium content, would further complicate the use of the Technical Specification curves. Deleting these curves in conjunction with adding a reference to the source material will avoid a potential source of error.
To determine the proper MAPLHGR value for a particular axial location in a fuel bundle, the MAPLHGR tables in the LOCA analysis t references will be programmed into the plant process computer.
j Other acceptable methods will be used if the computer is I unavailable.
Section 5.2 on page 245 describes the physical design of fuel bundles in the FitzPatrick core. This section is revised to reflect the addition of fuel types GE8X8EB and QUAD + inserted as Reload 7. Fuel type 8X8R is no longer used in the FitzPatrick core and is deleted from this section. Fuel type BP8X8R, added as Reload 6, was inadvertantly ommitted from this section in the Technical Specification amendment submittal for Reload 6/ Cycle 7.
III. IMPACT OF THE PROPOSED CHANGES The impact of the proposed changes would be to allow startup and operation of FitzPatrick following the upcoming Reload 7/ Cycle 8 refueling outage. This outage is currently scheduled to begin in January 1987.
l l
. Of the 188 new fuel bundles to be inserted into the l FitzPatrick core for Reload 7, four (4) are supplied by l Westinghouse Nuclear Energy Systems and are designated as fuel type QUAD +. These bundles are demonstration lead test assemblies to be used as part of the program to qualify Westinghouse as a vendor of nuclear fuel for FitzPatrick. A descrf.ption of the I nuclear, thermal, and mechanical characteristics of this bundle type has previously been transmitted to the NRC for review in Reference 6.
The QUAD + fuel bundles were designed to be compatible with the General Electric fuel used in the FitzPatrick core. The nuclear and hydraulic properties of the QUAD + bundles have been ,
designed to be as nearly identical to the BP8DRB299 bundles inserted as Relo'ad 6 as practical. For the purpose of the General Electric Cycle 8 specific analyses, fuel bundles of type BP8DRB299
, were assumed to reside in the core locations designated for the QUAD + bundles.
These bundles are very similar to the QUAD + bundles approved for use in the Brown's Ferry Unit 2 core (Reference 7). The core loading constraints applied by Westinghouse and the TVA for the Brown's Ferry QUAD + lead test assemblies also apply at FitzPatrick. These six constraints are given in Reference 6. The calculations done to support the conclusion that the QUAD +
assemblies will operate with a margin of at least 20% in bundle power to the lead assembly during Cycle 8 are contained in Reference 8. 5.'he application of the six constraints and the calculations demonstrating the required 20% power margin support the conclusion as given in the NRC safety evaluation report that the QUAD + asse:nblies will have no adverse safety impact on Cycle 8 operation at FitzPatrick.
The remaining 184 bundles of Reload 7 are of fuel bundle type GE8X8EB and are designated BD319A. This fuel type incorporates the design fee.tures described in Reference 5 for GE8X8EB fuel.
Analyses on the Cycle 8 core at FitzPatrick have resulted in small changes in the MCPR values over the Cycle 7 values. Use of these new MCPR values, as documented in this application as Attachment III, will assure no adverse safety impact in the reloaded core.
A new loss of coolant accident (LOCA) analysis has been i performed for FitzPatrick in order to support operation of the GE8X8EB fuel to its design power levels (Attachment IV). The analysis has been performed using codes and methods approved by the NRC. The full break spectrum has been analyzed with appropriate failure criteria applied and nominal input values assumed. Peak clad temperatures (PCT) based on Appendix K criteria were calculated at discrete points in the break spectrum, as provided in the approved application metlodology. These peak clad temperatures show considerable margin to PCT limits and demonstrate no adverse safety impact associated with the use of the GE8X8EB HAPLHGR limits as given in this appli. cation.
The maximum linear heat generation rate (MLHGR) limit of 14.4 KW/ft for GE8X8EB fuel as applied for in this application is the MLHGR considered in the NRC review of this fuel type. Generic NRC acceptance of this fuel design is documented as Amendment 10 to Reference 5.
Thej'y and 9' parameters used for scram speed surveillance for MCPR limits are changed to be consistent with the requirements of the GEMINI transient methods approved as Amendment 11 to Reference 5.
Another change to the Technical Specifications necessary to support Cycle 8 operation will be a change to the Standby Liquid Control System Bases. This change will specify a minimum of 660 ppm boron concentration in the reactor core following a Standby Liquid Control System injection. This change and its associated o
safety evaluation will be submitted to the NRC in a separate Technical Specification amendment application.
IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the porposed Amendment would not involve a significant hazards consideration as stated in 10 CFR 50.92 since it would not:
- 1. involve a significant increase in the probability or consequences of an accident previously evaluated.
Approved methodologies and codes have been used to perform all analyses concerning the General Electric fuel to be loaded at this refueling (Reference 5). The fuel design has been reviewed and approved for use at FitzPatrick under the constraints and methodologies detailed in Reference 5. There are no unique aspects of this fuel or its application which have not undergone prior NRC review and approval.
The Westinghouse fuel to be introduced has been designed and will be operated within the constraints detailed in a previous NRC safety evaluation report approving this type of demonstration program generically for BWRs 3 through 6 (Reference 7). The four QUAD + assemblies were designed to be compatible with currently used bundles and possess, as closely as possible, the operational characteristics of a previously used and accepted General Electric bundle. Constraints are applied to the use of the QUAD + bundles to assure that none of them will become a lead assembly during operation, nor a limiting bundle under transient conditions. Therefore, the probability or consequences of the accidents
! previously evaluated and described in the FSAR have not been increased.
r
- 2. create the possibility of a new or different kind of accident from any accident previously evaluated.
Refueling the FitzPatrick reactor is a periodic evolution performed in accordance with appropriate procedures and controlled by the Technical Specifications. The fuel bundles inserted as Reload 7 are not sufficiently different from previously used )
bundles as to create the possibility of a new or different type of accident. The assemblies have been fully reviewed and approved for use in power reactors by the NRC (References 5 and 7).
- 3. involve a significant reduction in a margin of safety.
The analyses performed in support of this reload assure maintenance of existing margins of safety. These analyses have resulted in core wide (MCPR) and bundle specific (MLHGR and MAPLHGR) limits for General Electric
, fuel which, when applied to the reloaded core, assure operation within the design criteria previously approved in Reference 5.
Through core positioning constraints and inherent design features, use of appropriate existing bundle specific limits and core-wide MCPR limits for the QUAD +
assemblies is conservative as concluded in Reference 7. ,
In the April 6, 1983 FEDERAL REGISTER (48FR14870), the NRC published examples of license amendments that are not likely to involve significant hazards considerations. Example number (iii) of that list is applicable to this proposed change and states in part:
"For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from these found previously acceptable to the NRC for a previous core at the facility in question are involved."
V. IMPLEMENTATION OF THE PROPOSED CHANGE Implementation of these changes, as proposed, will not impact the_ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.
VI. CONCLUSION The change, as proposed, does not constitute an unreviewed i safety question as defines in 10 CFR 50.59, that is it:
- a. will not change the probability nor the consequences of i an accident or malfunction of equipment important to I safety as previously evaluated in the Safety Analysis )
Report;
- b. will not increase the possibility of an accident or malfunction of a different type than any evaluated i previously in the Safety Analysis Report;
- c. will not reduce the margin of safety as defined in the basis for any technical specification;
- d. does not constitute an unreviewed safety question; and
- e. involves no significant hazards consideration, as O
defined in 10 CFR 50.92.
r .
VII. REFERENCES
- 1. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR) as updated.
- 2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).
- 3. General Electric Report, " Supplemental Reload Licensing Submittal for James A. FitzPatrick Nuclear Power Plant Reload 7," 23A4825, November, 1986. (Included as Attachment III).
- 4. General Electric Report, " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis,"
NEDC-31317P, October 1986. (Included as Attachment IV).
- 5. General Electric Licensing Topical Report, "GESTAR II General Electric Standard Application for Reactor Fuel," NEDE 240ll-P-A-8, May 1986.
- 6. Westinghouse letter, E.F Rahe to R.M. Bernero, dated i
December 10, 1986, (NS-KnC-86-3187), submitting Westinghouse
! report WCAP-ll159 for NRC review.
! 7. NRC letter, M. Grotenhuis to S.A. White (TVA), dated August
! 19, 1986, providing the Safety Evaluation supporting l
Amendment No. 125 to TVA's Bwowns Ferry Unit 2.
- 8. Westinghouse letter, J.P. Ducru.- to G.L. Rorke, dated December 15, 1986, providing a report entitled, "FitzPatrick QUAD + Demonstration Assembly Power Margin and Monitoring."
l
f.
ATTACHMENT III TO JPN-86-63 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT RELOAD 7 (JPTS-86-023)
NL'W YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l
r 4
ATTACHMENT IV TO JPN 63 4
JAMES A. FITZPATRICK NUCLEAR POWER PLTJT SAFER /GESTR - LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS NEDC-31317P (JPTS-86-023)
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59