JPN-88-037, Proposed Tech Specs,Deleting Cycle 8 Re Discharged Fuel & Adding Cycle 9 to Reduce Fuel Cladding Integrity Safety Limit

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Proposed Tech Specs,Deleting Cycle 8 Re Discharged Fuel & Adding Cycle 9 to Reduce Fuel Cladding Integrity Safety Limit
ML20151M881
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 07/29/1988
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML19292J200 List:
References
JPN-88-037, JPN-88-37, NUDOCS 8808080028
Download: ML20151M881 (25)
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ATTACHMENT I TO JPN-88-037 l

l PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING RELOAD 8/ CYCLE 9 (JPTS-88-016) l i

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i New York Power Authority i i

! JAMES A, FITZPATRICK NUCLEAR POWER PLANT I i Docket No. 50-333 DPR-59 l

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l JAFNPP i

LIST OF FIGURES j

Flours Title Eagg 3.1-1 Manual Flow Control 47a 3.1-2 Operating Limit MCPR versus 47b 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 l 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 and 3.5.J.2. 134 3.5-6 (Deleted) 135d 4 3.5-7 (Deleted) 135e 3.5-8 (Deleted) 135f 3.5-9 MAPLHGR Versus Planar Average Exposure Reload 4, P8DRB284L 1359 P

3.5-10 (Deleted) 135h 3.5-11 MAPLHGR Versus Planar Average Exposure Reload 6, BP8DRB299 1351 9

3.5-13 MAPLHGR Versus Planar Average Exposure Reload 8, BD336A 135k l

j 3.5-14 MAPLHGR Versus Planar Average Exposure Reload 8, BD339A 1351

( 3.6-1 Reactor Vessel Pressure Temperature Limits 163 1

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 1 i l

1 1

Amendment No. J4, Af, +5, 64, 74, 74, P8, 96, 136 1

vii

JAFNPP ,

1.1 BTEL CIADJJyG INTEGRITY 2.1 ELEMLADDING INTEGRITY .

Applicability: Applicability:

The Safety Limits established to preserve the fuel The Limiting Safety System Settings apply to trip cladding integrity apply to those variables which settings of the instruments and devices which are monitor the fuel thermal behavior. provided to prevent the fuel cladding integrity Safety Limits from being exceeded.

Objective: Qbiective:

The objective of the Safety Limits is to establish The objective of the Limiting Safety System Settings limits below which the integrity of the fuel cladding is to define the level of the process variables at is preserved. which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.

Specifications: Specifications:

A. Peactor Pressure > 785 psig and Core Flow 2 10% A. Trip Settinos of Ratttd The limiting safety system trip settings shall be

< The existence of a minimum critical ex)wer ratio as specified below:

l (MCPR) less than 1.04 shall constitute violation of the fuel cladding intog11ty safety limit, 1. Eeutron Flux Trio Settir.gs hereafter called the Safety Limit. An MCPE l safety limit of 1.05 shall apply during single- a. IRM - The IRM flux scram setting shall loop operation. be set at f_120/125 of full scale.

Amendment No. M , 74, 3d, 95, p6 7

JAFNPP 1.1 BASES 1.1 FUEL CLADDINJ_INTEGRJIX The fuel cladding integrity limit is set such that no elevated clad temperature and the possibility of calculated fuel damage would occur as a result of an clad failure. However, the existence of critical abnormal operational transient. Because fuel damage power, or boiling transition, is not a directly is not directly observable, a step-back approach is observable parameter in an operating reactor.

rsed to establish a Safety Limit such that the mini- Therefore, the margin to boiling transition is mum critical power ratio (MCPR) is no less than 1.04. calculated from plant operating parameters such MCPR pl.04 represents a conservative margin relative as core power, core flow, feedwater temperature, to the conditions required to maintain fuel cladding and core power distribution. The margin for each integrity. The fuel cladding is one of the physical fuel assembly is characterized by the critical barriers which separate radioactive materials from .ower 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 boiling divided by the actual bundle power.

cr cracking. Although some corrosion er use related The minimum value of this ratio for any bundle in cracking may occur during the life of the cladding, the core is the minimum critical power ratio fission product migration from this source is incre- (MCPR). It is assuined that the plant operation mentally cumulative and continuously measurable. is controlled to the nominal protective setpoints Fuel cladding perforations, however, can result from via the instrumented variable, i.e., the oper-thermal stresses ~*;ch occur from reactor operation ating domain. The current load line limit cignificantly above design conditions and the protec- analysis contains the current operating domain tion system safety settings. While fission proouct map. The Safety Limit '.MCPR of 1.04) has migration from cladding perforation is just as sufficient conservatism to assure that in the measurable as that from use related cracking, the event of an abnormal operational transient thermally caused cladding perforations signal a initiated from the MCPR operating conditions in threshold, beyond which still greater thermal specification 3.1.B, more than 99.9% of the fuel ctresses may cause gross rather than incremental rods in the core are expected to avoid boiling cladding deteriorstion. Therefore, the fuel cladding transition. The MCPR fuel cladding safety limit Safety Limit is defined with margin to the conditions is increased by 0.01 for single-loop operation as which would produce onset of transition boiling, (MCPR discussed in Reference 2. The margin between of 1.0). These conditions represent a significant MCPR of 1.0 (onset of transition boiling) and the departure from the condition intended by design for Safety Limit is derived from a detailed statisti-planned operation. cal analysis considering all of the uncertainties in monitoring the core operating state including A. EgActor Pressure > 785 psio and_ Core Flow > 113 the uncertainty in the boiling transition corre-nC_ Rated lation as described in Reference 1. The uncer-tanties employed in deriving the Safety Limit are Onset of transition boiling results in a decrease in neat transfer from the clad and, therefore, Amendment No. 14, K , 24, 30, e , M , 96 12

JAFNPP 1.1 (cont'd) .

Provided at the beginning of each fuel cycle. At 100% power, this limit is reached with maximum Because the boiling transition correlation is fraction of limiting power density (MFLPD) equal Daced 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 (FRF),

fuel assembly at the Safety Limit would not the APRM scram and rod block settings shall be Produce boiling transition. Thus, although it is adjusted as required in specifications 2.1.A.1.c not required to establish the safety limit, and 2.1.A.I.d.

additional margin exists between the Safety Limit and the actual occurrence of loss of cladding B. Core Thermal Power Limit (Reactor Pressure T785 integrity. psial However, if boiling transition were to occur, clad At pressures below 785 psig, the cora elevation perforation would not be expected. Cladding pressure drop is greater than 4.56 psi for no temperatures would increase to approximately boiling in the bypass region. At low powers and l 1100'F which is below the perforation temparature flows, this pressure drop is due to the elevation of the cladding material. 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 in the range (CETR) where fuel similar in design to FitzPatrick of 1-5 MNt, 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 significant 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 in the elevation If reactor press ce should ever exceed 1400 psia pressure drop due to the density difference during normal power operation (the limit of between the boiling water in the fuel channel and applicability of the boiling transition correla- the non-boiling water in the bypass region. Full tion) It would be 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 critical power at 2P, x 103 lb/hr is approxi-In addition to the boiling transition limit mately 3.35 MNt. With the design peaking (Safety Limit), operation is constrained to a factors, this corresponds to a core thermal power maximum LHGR of 14.4 KW/ft for GE8X8EB 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.

Amendment No. 14, 21, M , #3, (4, ~J4, 18 13

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JAFNPP ".

3.1 (CONTINUED)

MCPR Operating Limit for Incremental C. MCPR shall be determined daily during reactor Cycle Core Averaoe Exposure power operation at 2 25% of rated thermal power and following an;r change in power level or dis-At RBM Hi-trip BOC to EOC-2GWD/t to EOC-1GWD/t tribution that would cause operation with a 1stvel settino EOC-2GWD/t E_OC-1GWD/t__ to EOC limiting control rod pattern as described in the bases for Specification 3.3.B.S.

S = .66W + 39% 1.25 1.27 1.30 D. When it is determined that a channel has failed S = .66W + 40% 1.25 1.27 1.30 in the unsafe condition, the other RPS channels that monitor the same variable shall be function-S = .66W + 41% 1.2S 1.27 1.30 ally tested inmediately before the trip system containing the failure is tripped. The trip S = .66W + 42% 3.23 1.28 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 surveillance testing is being performed on the other RPS channels.

S = .SoM + 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 f gDAVE) 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 Where T= (? AVE f BMI A- B) AVE I B and iAVE = the average scram time to notch 2. The averaga scram time to notch position 38 position 38 as defined in speci- is determined as follows:

fication 4.1.E.2, TB = the adjusted analysis mean scram n n

. time as defined in specification

4.1.E.3, 7AVE = Ni T i Ni 7A = the scram time to notch position 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 l

of active rods measured in Amendment No. g4, 74, % , 86, 98, 169 31

_ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - _ .-- _--n - - - - - . - - - -- - - ,

JAENPP TABL_E 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION _REOUIREMENT NOTES OF TABLE 3.1-1 (cont'd)

14. The APRM flow biased high neutron flux signal is fed through a time constant circuit of approrimately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds l directly to instantaneous neutron flux.
15. This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is place in the Run position.

16.* During the proposed Hydrogen Addition 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 Radiation Monitor Trip Level Setpoint will be raised to fthree times the increased 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 (0.66W+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.

Amendment No. fG, 84, 90, /

43a (

i JAFNPP Figure 3.1-1 1 i

K g FACTOR 1.4 i

1.3 -

+

1.2 -

AUTOMATIC FLOW CONTROL 1.'1 -

MANUAL FLOW CONTROL SCOOP TUBE SET-FOINT CAUBRATION POSITIONED SUCH THAT FLOWMAX = 102.5%

107.0%

1.0 - 112.0%

117.0 %

0.e i I ' ' L ' '

40 40 50 60 70 80 90 100 CORE FLOW t%) l l

l i Amendment No. J4 1

47a

JAFNPP )

i Figure 3.1-2 l

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ODeratinc Linli MCPR Versus T '

(defined in Section 3.1.B.2) 1 LOR ALL FUEL TYPES 1.35' ~1.35

- 1.34 Operating Limit EOC -1.32 MCPR 1.30- -1.30 EOC-1 W

1.27-1.25- EOC-2 -1.25 i

i 1.23 - -1.23 i I I I I I I I I I O 0.2 0.4 0.6 0.8 1 l

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' Amendment No. (4, 74, 74 , 84, W 47b ,

1

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. Averace Planar Linear Heat Generation Rate vented from the high point prior to the (APLUGR1 return of the system to service.

During power operation, the APLHGR for each type 3. Whenever the HPCI, RCIC, or Core Spray System of fuel as a function of axial location and is lined up to take suction from the conden-average planar exposure shall be within limits sate storage tank, the discharge piping of based on applicable APLHGR 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 excced the limiting value for the most limiting and RHR System discharge piping high points 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. Averace 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 determined daily value for APLHGR is 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 returned to within the prescribed limits.

1 Amendment No. 4ff, 6(, 74, SE, 98, 189 123

JAFNPP 3.5 BASES (cont'd) requirements for the emergency diesel generators. are within the 10 CFR 50 Appendix K limit. The limiting values for APLHGR are given in Figures G. Maintenance of Filled Discharoe Pipe 3.5-11 through 3.5-14. Approved limiting values I of APLHGR as a function of fuel type are given in If the discharge piping of the core spray, LPCI, NEDO-21662-2 (as amended) for Reload 6 fuel. I RCIC, and HPCI are not filled, a water hammer can Approved limiting values of APLHGR as a function develop in this piping when the pump (s) are of fuel and lattice types are given in NEDC-started. To minimize damage to the discharge 31317P (as amended) for Reload 7 and 8 fuel. f piping and to ensure added margin in the opera- These values are multiplied by 0.84 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 (LHGR) supply that line must be assumed to be inoperable for technical specification purposes. However, This specification assares that the linear heat if a water hammer were to occur, the system would generation rate in any rod is less than the still perform its design function. design linear heat generation.

H. Averace Planar Linear Heat Generation Rate The LHGR shall be checked daily during reactor (APLHGR) operation at 25% rated thermal power to deter-mine if fuel burnup, or control rod movement, has This specification assures that the peak cladding caused changes in power distribution. For LHGR temperature following the postulated design basis to be a limiting value below 25% rated thermal loss-of-coolant accident will not exceed the limit power, the ratio of local LHGR to average LHGR specified in 10 CFR 50 Appendix K. would have to be greater than 10 which is pre-cluded by a considerable margin when employing The peak cladding temperature following a postu- any permissible control rod pattern.

lated loss-of-coolant accident is primarily a l function of the average heat generation rate of - l all the rods of a fuel assembly at any axial location and is only dependent secondarily on the l rod to rod power distribution within an assembly.

l Since expected local variations in power distri-bution within a fuel assembly affect the calcu-lated peak clad temperature by less than 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. pg, 7f, 8f, 98, 399 130

JAFNPP Figure 3.5-10 -

(This page is intentionally blank.)

1 Amendment No. 6(, 74, pff, 135h

JAFNPP Figure 3.5-13 -

Maximum Averace Planar Linear Heat Generation Rate (MAPLHGR)

Versus Averace Planar Exposure 14-Reload 8, BD336A 13-

Reference:

NEDC-31317P a

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$ U 11-3x0

$3 cm 10-Sb

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$$ 7-Eu we xc

$o 6-1 I I I I I I I I I O 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 MAPLHGR values are multiplied by 0.84. exposure dependent MAPLHGR values.

Amendment No.

135k

JAFNPP Figure 3.5-14 .

Maximum Averace Planar Linear Heat Generation Rate (MAPLHGR)

Versus Average Planar Exposure 14-Reload 8, BD339A 13-

Reference:

NEDC-31317P e

en j 12-u 4N c 3:

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A 9-ee tnv cc gZ 8- _

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-r4 m xc e8 6-1 I I I I I I I I I O 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 MAPLHGR values are multiplied by 0.84. exposure dependent MAPLHGR values.

Amendment No.

1351

______._...______..____________._______________--.______.-________m_ _ _

JAFNPP 5.0 RESIGN FEATURES B. The reactor core contains 137 cruciform-shaped control rods as described in Section 3.4 of 5.1 SITS 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 of the Nine Mile Point Nuclear Station, Unit Table 4.2-1 and 4.2-2 of the FSAR. The applicable

1. The NPP-JAF site is on Lake Ontario in design codes are described in Section 4.2 of the Oswego Country, New York, approximately 7 FSAR.

miles northeast of Oswego. The plant is 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.

tne reactor building and any points of poten-tial gaseous effluents, with the exception of B. The secondary containment is as described id

< 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 containment and piping passing through such penetrations are 5

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 l fuel assemblies. For the current cycle, three 5.5 FUEL STORAGE fuel types are present in the core: BP8X8R, GE8X8EB, and QUAD +. The GE fuel types are A. The new fuel storage facility design criteria described in NEDO-24011. The BP8X8R fuel type are to maintain a Ke gg dry (0.90 and has 62 fuel rods and 2 water rods and the flooded <0.95. Compliance shall be verified GE8X8EB 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-ll159 and has 64 fuel rods.

Amendment No. 3d, 42', 49, 64', 96,"Hf, IDT 245

ATTACHMENT II TO JPN-88-037 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING RELOAD 8/ CYCLE 9 (JPTS-88-016) s New York Power Authority

. JAMES A. FITZPATRICK NUCLEAR FOWER PLANT .,

Docket No. 50-333 DPR-59

Attachment 11 to JPN-88-037 i SAFETY EVALUATION

  • ge 1 of 7 I. DESCRIPTION OF TIIE PROPOSED CIIANGES l

l 'Iae proposed changes to the James A. FitzPatrick Technical Specifications revise pages vii,7, l 12,13, 31, 43a, 47a, 47b,123,130,135h, and 245, and adds two new Figures, 3.5-13 and 14, l on new pages 135k and 135L.

l l

Page vii, List of Figure s

[a] Replace entry 3.5-10 with "(Deleted)"

1

[b] Add two new entries 3.5-13 and 3.5-14 The figures will reside on new pages 135k and 135L. Entries 3.5-13(14) will read as follows:

h1APLHGR Versus Planar Average Exposure Reload 8, BD336A h1APLHGR Versus Planar Average Exposure Reload 8, BD339A Page 7, 61.1.A. Fuel Cladding Integrity

[c] In the first sentence, replace "1.07" with "1.04."

[d] In the second sentence, replace "An h1CPR Limit of 1.08" with "Ar. htCPR safety limit of 1.05."

Page 12, Bases for $1.1. Fuel Cladding Integrity

[e] In two places, replace "1.07" with 1.04."

Bases for 61.1. A.

[f] Replace "1.07" with 1.04."

Page 13, Bases for 61.1.B. Core Thermal Power Limit

[g] Replace this section in its entirety with:

At pressures below 785 psig, the core elevation pressure drop is greater than 4.56 psi for no boiling in the bypass region. At low powers and flows, this pressure drop is due to the elevation pressure of the bypass region of the core. Analysis shows that for bundle power in the range of 1-5 htWt, the channel flow will never go below 28 x 108 lb/hr. This flow results from the pressure differential between the bypass region and the fuel channel. The pressure differential is primarily a result of changes in the elevation pressure drop due to the density difference between the boiling water in the fuel channel and the non-boiling water in the bypass region. Full scale ATLAS test data taken at pressures from 0 to 785 psig indicate that the fuel assembly critical power at 28 x 108 lb/hr is approximately 3.35 h1Wt. With the desir,n peaking factors, this corresponds to a core thermal power of more thaa 50%. Thus, a core thermal power limit of 25% for reactor pressures below 785 psig is conservative.

s v , , _ _- , _ .

Attachment II to JPN-88-037 SAFETY EVALUATION Page 2 of 7 Page 31,63.1.B.l. hiCPR Operating Limit for incremental Cycle Core Average Exposure

[h] Replace the values in this table with values specific for FitzPatrick Cycle 9. The new values can be found in Attachment I.

Page 43a, Notes for Table 3.1-1

[i] In note 14, correct the spelling of "signal."

Page 47a, Figure 3.1-1 K, Factor

[j] Replace the existing fis,ure with one approved for use with the GEXL-PLUS correhtion. See Attachment I for the new figure.

Page 47b, Figure 3.1-2 Operating Limit hiCPR Versus U for All Fuel Types

[k] Replace the existing figure with one specific for Cycle 9 operation. See Attachment I for the new figure.

Page 123,63.5.H. Average Planar Lituar Heat Generation Rate (APLHGR)

[1] Replace "Figures 3.5-10 th:ough 3.5-12" with "Figures 3.5-11 through 3.5-14" 1

Page 130, Bases for 63.5.H. Average Planar Linear Heat Generation Rate (APLHGR)  !

[m] In the top paragraph of the right column, make the following changes:

Rep! ace "Figures 3.5-10 through 3.5-12" with "Figures 3.5-11 through 3.5-14;"

replace "Reload 5 and 6 fuel" with "Reload 6 fuel;"

and replace "NEDC-31317P for Reload 7 fuel" with "NEDC-31317P (as amended) for Reload 7 and 8 fuel."

Page 135h, Figure 3.5-10

[n] Delete this figure. Insert "(This page is intentionally blank.)"

Pages 135k and 135L, Figures 3.5-13 and 3.5-14 hiaximum Average Planar Linear Heat eneration Rate (hiAPLHGR) Versus P;anar Average Exposure - Fuel Types BD336A and BD339A

[o] Add new figures prescribing the hiAPLHGR limit vs. Average Planar Exposure for fuel types BD336A (page 135k) and BD339A (page 135L), added as Reload 8.

Page 245, 45.2.A. Reactor

[p] This section is revised to reflect the fuel types in the FitzPatrick Cycle 9 core. See Attachment I for the text. l l

l

L Attachment II to JPN-88-037 SAFETY EVALUATION Page 3 of 7 II. PURPOSE OF TIIE PROPOSED CIIANGES The purpose of the proposed changes is to support plant start-up and operation after the Reload 8/ Cycle 9 refueling outage. During this outage,184 fuel bundles are to be removed from the reactor core and replaced by new fuel. The changes to the Technical Specifications involve deleting specifications associated with the discharged fuel and with Cycle 8 specific analyses, and replacing them with ones which are appropriate for the new fuel and are based on Cycle 9 specific analyses.

To simplify the discussion of the proposed changes, the 16 individual changes on 14 technical specification pages are grouped into three categories:

A) Changes to reflect the removal of fuel type P8DRB299 from the reactor core and its replacement with fuel types BD336A and BD339A; B) Changes to reflect the Cycle 9 specific transient and accident analyses, use of the GEXL-PLUS critical power correlation, and the new fuel cladding integrity safety limit; C) Miscellaneous changes (e.g., correction of a typographical error and clarification of I text). I CATEGORY A 184 fuel bundles are to be removed from the reactor core and placed in the spent fuel pool for storage. Replacing these fuel bundles in the core will be 152 bundles of fuel type BD336A and 32 bundles of fuel type BD339A. The two new fuel types are General Electric's GE8X8EB design and are mechanically identical to the Reload 7 fuel. The U 2ss enrichment and gadolinium content are varied to support cycle specific needs and to improve fuel economy.

The changes to the Technical Specifications involve changes to the List of Figures, MAPLHGR figures and reactor design sections. The following changes fall into this category:

a, b,1, m, n, o, and p.

CATEGORY B Cycle 9 specific transient analyses, performed by General Electric Co. (GE), determine the operating limits for Cycle 9. The analyses were performed with GE's GEXL-PLUS thermal correlation and the revised safety limit MCPR of 1.04. The results of these analyses are contained in Reference 3 and are included in this application as Attachment II.

The changes to the Technical Specifications involve changes to the MCPR sections and Base;,

MCPR operatin8 limit ta'Ule, the MCPR figure, and the K g curves. The following changes fall into this category: c, d, e, f, h, j, and k.

CATEGORY C One typographical error (the misspelling of the word "signal") is corrected. Change [i] to the Technical Specifications reflects this correction.

In NRC Inspution Report 50-333/M-05 (Reference 4), an NRC inspector considered the Technical Specification Bases for ll.l.B to be unclear. This section discusses the reactor thermal power limit when the reactor pressure is below 785 psig. The Authority committed to

Attachment II to JPN-88-037 SAFETY EVALUATION Page 4 of 7 submit a revision to this bases section in this reload application submittal (Reference 5).

Change [g] to the Technical Specifications is that revision.

Cleed pages vii and 123, contained in Attachment I to this application, carry over errors wL. currently exist in the Technical Specifications. The Authority has previously submitted an application for amendment to the Technical Specifications to correct these and other administrative and typographical errors. (Reference 6)

III. IMPACT OF TIIE PROPOSED CIIANGES The overall impact of the proposed changes would be to allow start-up and operation of the FitzPatrick Nuclear Power Plant following the upcoming Reload 8/ Cycle 9 refueling outage.

This outage is entrently scheduled to begin in late August 1988 and last until early November.

To simplify the discussion of the impact of each of the individual changes, this evaluation will address the three categories of changes that were previously defined in Section II above.

CATEGORY A The 184 fuel bundle of Reload 8 are of fuel bundle type GE8X8EB. 152 of these bundles are designated BD336A, and the 32 others are designated BD339A. These fuel bundles incorporate the design features described in Reference 7.

The fuel to be added in Reload 8 is similar to that added in Reload 7 in that it contains several lattice types of varying gadolinium content. To determine the proper h1APLIIGR value for a particular axial location in a fuel bundle, ,he h1APLIIGR tables contained in Reference 8 will be programmed into the plant process computer and backup computer system. When hand calculations are necessary, the most limiting enriched uranium lattice hfAPLIIGR value is applicable. The exposure dependent limiting hfAPLIIGR values are shown in Figures 3.5-13 and 3.5-14 on pages 135k and 1351 for fuel types BD336A and BD339A respectively. The tables used to generate these figures are included in Attachment 111.

The hfAPLilGR curve for fuel type P8DRB299 (Figure 3.5-10 on page 135h) is deleted since all fuel bundles of this tyra are being removed from the core.

CATEGORY B The changes in this category reflect the Cycle 9 specific transient and accident analyses performed by General Electric Co. for FitzPatrick. The results of these analyses are contained in Reference 3 and are included in this application as Attachment III.

The analyses were performed with GE's GEXL-PLUS critical heat flux correlation. This correlation provides a more accurate determination of the critical heat flux. Implementation of the GEXL-PLUS correlation increases plant fuel cycle efficiency and operating flexibility without reducing the current safety margins (Reference 9). Use of the GEXL-PLUS correlation was approved generically for GE fueled BWRs by the NRC in Amendment 15 to Reference 7.

The analyses required for Cycle 9 were performed with the revised safety limit minimum critical power ratio (hfCPR) of 1.04, instead of the previous safety limit hfCPR of 1.07. The safety limit hiCPR of 1.04 is a result of the statistical analyses performed by GE for reactor cores which are operated with a second successive reload of high enrichment, high R-factor,

Attachment 11 to JPN-88-037 l SAFETY EVALUATION  ;

Page 5 of 7 l 8x8 fuel. The analyses associated with the new safety limit predict that 99.9% of tne fuel rods in the core would avoid boiling transition. This is the same criteria that was previously applied to the safety limit of 1.07 and therefore, the change does not reduce any margin of safety. Use of the revised MCPR safety limit of 1.04 was generically approved by the NRC in Amendment 14 to Reference 7.

CATEGORY C Correction of the misspelling of "signal" has no impact on plant operation and is being made to correct the error.

Clarification of the Bases for Ql.l.B. has no impact on the plant since no change is being made to the safety limit power level for the Core Thermal ?ower Limit when the reactor pressure is below 785 psig. The Authority concurs with the NRC that the existing Bases section is unclear. The proposed replacement to this section was developed with the assistance of General Electric Co. It attempts to more clearly describe the theoretical and experimental bases for the existing power level limit than the existing paragraph.

IV. EVALUATION OF SIGNIFICANT IIAZARDS CONSIDERATION Operation of the FitzPatrick Plant in accordance with the proposed 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 probabliity or consequences of an accident previously evaluated. NRC approved methodologies and codes have been used to perform all analyses concerning the General Electric Co. fuel to be loaded at this refueling (Reference 7). The fuel design has been reviewed and approved for use at FitzPatrick under the constraints and methodologies detailed in Reference 7.

There are no unique aspects of this fuel or its application which have not i undergone prior NRC review and approval. The refueling of the FitzPatrick l reactor and Cycle 9 operation does not increase the probability or consequences of any accident previously evaluated.

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 8 are mechanically identical to those inserted in Reload 7 and will not create the possibility of a new or different type of accident. The nuclear characteristics of the individual fuel bundles and the core loading pattern have been fully analyzed by the General Electric Co. and do not 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 (Reference 7).
3. involve a significant reduction in a margin of safety. The analyses performed in support of this reload assure maintenance of all existing margins of safety. These analyses have resulted in core wide (MCPR) and bundle specific (MAPLilGR) limits for General Electric Co. fuel which, when applied to the reloaded core, assure operation within the design criteria previously approved in Reference 7.

The revised MCPR safety limit provides the same margin of safety as the previous

Attachment II to JPN-88-037 SAFETY EVALUATION Page 6 of 7 safety limit in preventing boiling transition. This change was previously approved by the NRC for use in GE fueled BWR reactors (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 chance 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 those found previously acceptable to the NRC for a previous core at the facility in question are involved.

V. Ih1PLEh1ENTATION OF TIIE PROPOSED CIIANGE Implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the chan8es impact the environment.

VI. CONCLUSION The change, as proposed, does not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, it:

A. will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; B. will not increase the possibility of an accident or malfunction of a different type from any previously evaluated 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 defined in 10 CFR 50.92.

VII. REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report.
2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20, 1972, and Supplements.
3. General Electric Co. Report, "Supplemental Reload Licensing Report for James A.

FitzPatrick Nuclear Power Plant - Reload 8 (Cycle 9)," 23A5898, Revision 0, June, 1988. (Included as Attachment Ill)

4. NRC letter, E. J. Brunner to J. D. Leonard, Jr. (PASNY), dated March 14, 1978, transmitting NRC Inspection Report 50-333/78-05: Page 9, Item 6, Technical Specifications Review - Open Item 78-05-07.

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Attachment II to JPN-88-037 )

SAFETY EVALUATION Page 7 of 7

5. NRC letter, E. C. Wenzinger to R. J. Converse (NYPA), dated March 29, 1988, transmitting NRC Inspection Report 50-333/88-01: Enclosure 2, Item 2, Previous Inspection Findings, - Inspection Followup Item 78-05-07.
6. NYPA letter, J. C. Brons to the NRC, JPN-88-023, dated May 27, 1988, containing an application for amendment to the FitzPatrick Technical I Specifications regarding Administrative Changes. I
7. General Electric Licensing Topical Report, "GESTAR-II General Electric Standard Application for Reactor Fuel," NEDE 240ll-P-A-8, May 1986. l l
8. Genersl Electric Co. Report, "James A. FitzPatrick Nuclear Power Plant l SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NRDC-31317P, l Errata and Addenda No. 2, June 1988. (Included as Attachment IV)
9. General Electric Co. letter, J. S. Charnley to C. O. Thomas (NRC), dated January I 25, 1986, concerning Amendment 15 to GESTAR-II (Implementation of GEXL- 4 PLUS critical power correlation).

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ATTACHMENT III TO JPN-88-037 SUPPLEMENTAL RELOAD LICENSING REPORT FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT I RELOAD 8 (CYCLE 9) i (JPTS-88-016)  !

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 i

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  • 6 ATTACHMENT IV TO JPN-88-037 JAMES A. FITZPATRICK NUCLEAR POWER PLANT SAFER /GESTR-LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS ERRATA AND ADDENDA No. 2 (J PTS-88-016) s New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 h

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