ML20006A408

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Proposed Tech Spec Changes Re Removal of cycle-specific Parameter Limits from Tech Specs & Relocating Limits to Core Operating Limits Rept
ML20006A408
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 01/12/1990
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20006A403 List:
References
JPTS-88-020, JPTS-88-20, NUDOCS 9001260228
Download: ML20006A408 (41)
v • d • e


Text

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ATTACHMENT I q l

PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING REMOVAL OF CYCLE SPECIFIC PARAMETER LIMITS-JPTS-88-020 I l

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

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' JAFNPP LIST OF FIGURES i Figure Title - Py '

3.1 1 (Deleted) 47a b and-3.12 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 34.7 B 10 Atom % Enriched Volume- 110 Concentration Requirements 3.4 2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Umits of Specifications 3.5.J.1,3.5.J.2 and 134 3.5.J.3 >

3.5-3 (Deleted) 135al through 3.5-14 3.6 1 Reactor Vessel Pressure - Temperature Umits 163 4.6-1 Chloride Stress Corrosion Test Results at 500 F 164 6.1 1 (Deleted) 259 6.2 1 (Deleted) 260 Amendment No. }4, Of,46,64,74,74,66,96,129,14,126, ttf,13'f vii

d JAFNPP 1.0 (cont'd) V. Electrically Disarmed Control Rod surveillance tests, checks, calibrabons, and examinations shall J psann a rom MM,Mour . pieW type M un n muors are removed from the drive insert and withdrawal -

be performed within the specified surveillancein. tervals. These i intervals may be adjusted 25 percent. The interval as ds W@ rod @ d WW. M pertaining to instrument and electric surveillance shall never g e is equivMo v% M Mr.1ve andis WM n menate p,ui sh.

! exceed one operating cycle. In cases where the elapsed interval has exceeded 100 percent of the specified interval, the next W. High Pressure Water Fire Protechon System surveillance interval shall commence at the end of the original The High Pressure Water Fire Protechon System consists of: a

specified interval.

water source and pumps; and distnbubon system peping with U. Thermal Parameters associated post indicator valves (isolabon valves). Such valves

include the yard hydr i a A valves and the first valve ahead of
1. Minimum critical power ratio (MCPR)- Minimum value of the water flow alarm ot,.a on each sprinkler or water spray .

< the ratio of that power in a fuel assembly which is subsystem.

calculated to cause some point in that fuel assembly to X Staggerd Test Basis experience boiling transition to the actual assembly l operating power for all fuel assernblies in the core. A Staggered Test Basis shall consist et

2. Fraction of Umiting Power Dansity - The ratio of the linear a. A test schedule for "n* systems, subsystems, trains heat generation rate (LHGR) existing at a given location to '

or other designWed wipiMainM by l the design LHGR.

dividing the specif;od test interval into "n" equal

3. Maximum Fraction of Umiting Power Density - The subintervals.

Maximum Fraction cf Umiting Power Density (MFLPD) is b. The testing of one system, subsystem, train or other .

the highest value existing in the core of the Fraction of designMM sw wit g the beginning d each Limiting Power Density (FLPD).

subintervi l

4. Transition Boiling - Transition boiling means the boiling j region between nucleate and film boiling. Transition Y. Rated Recirculation Flow boiling is the region in which both nucleate and film boiling

]- That drive flow which produces a core flow of 77.0 x 1d6lb/hr.

occur intermittently with neither type being completely stable.

Amendment No. 46,64, M,74,106,134 6

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AD. Core Operating Umits Report (COLR)

Z. Top of Active Fuel '

This report is the plant-specific document that prowdes the core .

- The Top of Active Fuel, correspondicg to the top of the ennched opa ath g Imts for the current opwati g cycle. These cycle- _

fuel column of each fuel bundle, is located 352.5 inches above specific operating Imts shall be determined for each reload vessel zero, wtich is the lowest point in the inside bottom of the - cycle in accordance with Specification 6.9.A.4. Plant operation reactor vessel. (See General Electric drawing No. 919D690BD.) wittin these opaaik g limits is addressed in individual Technical Specifications. '

R Rod M ~

Rod density is the number of control rod notches inserted ' i expressed as a fraction of the total number of control rod notches. All rods fully inserted is a condition representing 100 percent rod density.

AB. Purge-Purging Purge or Purging is the controlled process of discharging air or gas from a conficsinent in such a manner that replacement air or

, gas is required to purify the wi= =ia n.

AC. Venting

, Venting is the controlled process of releasing air or gas from a

! confinement in such a manner that replacement air or gas is not 'J provided or required.

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Amendment No. J d,5)1f 6a

_ _ _ _ _ . - _ = _ . - _ . . _ _ _ _ = _ _ - _ _ - _ . .

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g JAFNPP 1.1 (cont'd) 2.1 (cont'd)

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1.

[ b. ' APRM Flux Scram Trip Setting (Refuel or Start & Hot B. Core Thermal Power Umit (Reactor Pressure - <785 psig) - Standby Mode)

When the reactor pressure is <785 psig or core flow is less than APRM - The APRM a.ux scram settmg shall be <15 or equal to 10% of rated, the core thermal power shall not - percent of rated neutron flux with the Reactor Mode exceed 25 percent of rated thermal power. Switch in Startup/ Hot Standby or Refuel.

C. Power Transient c. APRM Flux Scram Trip Settogs (Run Mode)

To ensure that the Safety Umit established in Specification 1.1.A and 1.1.B i,s not exceeded, each required scram shall be initiated (1) Flow Referenced Neutron Flux Scram Trip by its expected scram signal. The Safety Umit shall be assumed Sening to be exceeded when scram is accomplished by a means other When the Mode Switch is in the RUN por:,ition,

, than the expected scram signal. the APRM flow referenced flux scram trip settmg shall be less than or equal to the limit established in the Cc*e Gim din g Umsts Report (COLR). This limit must be adjusted for single loop operation as specified in the 1 COLB.

! For no combination of recirculabon flow rate

! and core thermal power shall the APRM flux scram trip setting be allowed to exceed 117%

# of rated thermal power.

1 Amendment No. If,30',46,7tf,98, pr4 8

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JAFNPP 1.1 (cont'd) 2.1 (cont'd)

D. Reactor Water Level (Hot or Cold Shutdown Condit'ons) (2). Fixed High Neutron Flux Scram Trip Settog Whenever the reactor is in the shutdown condition with irradiated When the Mode Switch is in the RUN posdion, the fuel in the reactor vessel, the water level shall not be less than APRM fixed high flux scram trip settog shall be:

that corresponding to 18 inches above the Top of Active Fuel 3 g3g %

when it is seated in the core. -

d. APRM Rod Block Setting The APRM Rod block trip setting shall be less than or equal to the limit specEM in the Core OpereGy Urruts Report (COLR). The setting must be adjusted for sogle loop operation as specified in the COLR.

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Amendment No.1,g 4 46,6f;74 96,1p 9

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TH!S PAGE;S INTENTIONALLY BLANK 4

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i Amendment No.1/ 3g, M, S(,7f,74,p6 10

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JAFNPP 1.1 BASES 1.1 FUEL CLADDING INTEGRITY A. Reactor Pressure >785 psig and Core Flow >10% of Rated ~

The fuel cladding integrity limit is set such that no calculated Onset of transition boiling results in a decrease in heat transfer fuel damage would occur as a result of an abre mai from the clad and, therefore, elevated clad temperature and the operational transient. Because fuel damage is not directly possibility of clad failure. However, the existence of critical observable, a step-back approach is used to establish a Safety power, or boiling transition, is not a directly observable

, Umit such that the minimum critical power ratio (MCPR) is no parameter in an operating reactor. Therefore, the margin to less than 1.04. MCPR > 1.04 represents a conservative margin boiEng transition is calculated from plant operating parameters relative to the conditions required to maintain fuel cladding such as core power, core flow, feedwater temperature, and integrity. The fuel cladding is one of the physical barriers which core power distribution. The margin for each fuel a65minbiy is separate radioactive materials from the environs. The integrity characterized by the critical power ratio (CPR) which is the

! of this cladding barrier is related to its relative freedom from ratio of the bundle power which would produce onset of 5

perforations or cracking. Although some corrosion or use transition boiling dmded by the actual bundle power. The related cracking may occur 'during the life of the cladding, minimum value of this ratio for any bundle in the core is the fission product migration from this source is incrementally minimum critical power ratio (MCPR). It is assumed that the cumulative and continuously measurable. Fuel cladding plant operation is controlled to the nominal protective setpvir,Ls -

4 perforations, however, can result from thermal stresses which via the instrumented variable, i.e., the operating domain. The

, occur from reactor operation significantly above design current load line limit analysis contains the current operating  !

j conditions and the protection system safety settings. While domain map. The Safety Umit (MCPR of 1.04) has sufficient fission product migration from cladding perforation is just as conservatism to assure that in the event of an ebnvimal

measurable as that from use related cracking, the thermally operational transient initiated from the MCPR operating limit in caused cladding perforations signal a threshold, beyond which the Core Operating Umits Report, more than 99.9% of the fuel l

! still greater thermal stresses may cause gross rather than rods in the core are expected to avoed boiling transition. The incremental cladding deterioration. Therefore, the fuel cladding MCPR fuel cladding safety limit is increased by 0.01 for single-Safety Umit is defined with margin to the cerxiitions which loop operation as discussed in Reference 2. The margin would produce onset of transition boiling, (MCPR of 1.0). between MCPR of 1.0 (onset of transition boiling) and the These conditions represent a significant departure from the Safety Umit is derived from a detailed statistical analysis ,

condition intended by design for planned operation. considering all of the uncertainties in monit6 ring the core operating state including the uncertainty in the boiling transition correlation as desuibed in Referenca 1. The uncertanties employed in deriving the Safety Umit are Amendment No.14,18, /r,30,46,7f,96, td 12

s 1 JAFNPP. $

1.1 (cont'd) l l provided in Reference 1. Because the boiling transition At 100% power, this limit is reached with a maximum frachon of

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correlation is based on a large quantity of full scale data there is Iimiting power density (MFLPD) equal to 1.00. In the event of I a very high confida ice that operation of fuel assembly at the operation with a MFLPD greater than the frachon of rated power Safety Umit would not produce boiling transition. Thus, although (FRP), the APRM scram and rod block settogs shail be adjusted - ,

, it is not required to establish the safety limit, additional margin as required in the Co.a Operating Umits Report. I exists between the Safety Umit and the actual occurrence of loss B. Core Thermal Power Umit (Reactor Pressure <785 psig) of cladding integnty. .

However, if boiling transition were to occur, clad perforation A paswa Wm @ h mWm paswdop b greater than 4.56 psi for no boilmg in the bypass region. At low would not be expected. Cladding temperatures would increase powers and flows, this pressure drop is due to the elevation -

to approximately 1100 F which is below the perforation g e d tM W @ of h me.' W h M j temperature of the cladding matenal. This has been verified by for bundle power in the range of 1-5 MWt, the chis nial flow will tests in the General Electric Test Reactor (GETR) where fuel never go below 28 x 163tb/hr. This flow results from the similar ,n i design to FitzPatrick operated above the critical heat flux for a significant period of time (30 minutes) without clad gesswe Mwdid h tM by regon and the W j

h sei. The pesswe Mwentid is Wily a M of h p

! #mation. in the elevation pressure drop due to the densit/ difference If reactor pressure should ever exceed 1400 psia during normal between the boiling water in the fuel channel and the non-boiling power operation (the limit of applicability of the boiling transition water in the bypass regon. Full scale ATLAS test data taken at correlation) it would be assumed that the fuel cladding integrity pressures from 0 to 785 pJg indicate that the fuel assan eif Safety Umit has been violated.

citical power at 28 x 17 lb/hr is approximately 3.35 MWt. With l

In addition to the boiling transition limit (Safety Umit), operation the W peaMng factas,2 mrW to a me M i is constrained by the maximum LHGR identified in the Core m d mme .

,ame y d m W

! Operating Umits Report. 25Uw reacta pesswes W M psig is consavh i

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Amendment No. 14,24,30,42C 6(, F,199, tyf 13 r . .- -

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1.1 BASES (Cont'd) E. References C. Power Transient 1. General Electric Standard Application for Reactor Fuel,? -

Plant safety analyses have shown that the scrams caused by N 4 , latest amW m WWWn exceeding any safety system setting will assure that the Safety 2.. FitzPatrick Nucbar Fower Plant Singte. Loop Operation,- ,*

Umit of 1.1.A or 1.1.B will not be exceeded. Scram times are NEDO 24281, Aajust 1980.

checked periodically to assure the insertion times are adequate.

The thermal power trensient resulting when a scram is accomplished other than by.the expected scram signal (e.g.,

scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause feel damage. However, for this specification a Safety Umit violation will be assumed when a scram is only accomplished by means of a backup '

feature of the plant design. The concept of not approach ng a Safety Umit provided scram signals are operable is suppo.-ted by the extensive plant safety analysis.

D. Reactor Water Level (Hot or Cold Shutdown Condition)

During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of '

the active fuel during this time, the ability to cool the core is l reduced. This reduc' ion in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the ~

water level be reduced to two-thirds the core height.

Establishment of the Safety Umit at 18 in. above the top of the fuel provides adequate margin. This level will be continuously monitored whenever the recirculation pumps are not operating.

9

! Amendment No.1/,96 l

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JAFNPP ,

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BASES

.x 2.1 FUEL CLADDING INTEGRITY.

The most lirruting transients have been analyzed to determine The abnormal operational transients applicable to operation of which result in the largest reduction in CRITICAL POWER RATIO. .

the FitzPatrick Unit have been analyzed throughout the spectrum . The type of transeents evaluated were irbrease in pressure and of planned operating conditions up to the its6ial power power, positive rea::fvity msertion, and a colant temperature l condition of 2436 MWt. The analyses were based upon plant decrease. The limiting trantierit y% Ids tr.e largest delta MCPR.

operation in acccrdance with the operating map given in the When added to the Safety Umst, the required cpoid g limit l current load line limit analysis. In addition,2436 MWt is the MCPR in the Core Opoi& g Umsts Report is obtained.

licensed madmum power level of FitzPatrick, and this represents

, g gg , y gg ..g the mum steady-state power which shall not kr6-irgiy be

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as h ~m W M @ W hm 2

. that are input to the core dynamec behavior transeent computer The transient analyses performed for eaci1 reload are gven in programs described in Reference 2. The output of these Reference 2. Models and model conservatism are also programs along with the initial MCPR form the input for the described in this reference. As discussed in Reference 4, the further analyses of the thaineTy limited bundle with a sogle core wide transient analysis for one recirculation pump operation eteiiioi transient thoiuwd hydraulic code. The princspal result of is conservatively bounded by two. loop operation analysis, and the evaluation is the reduction in MCPR caused by the transsent.

the flow-dependent rod block and scram setpoint equations are y

, adjusted for one-pump operation.

Fuel cladding integnty is assured by the applicable operating limit MCPR for steady state conditions given in the Core -

Operating Umits Report (GOLR). These operating limit MCPR's are derived from the established fuel cladding integnty Safety Uinit, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady siste operating limit, it is required that the resulting MCPR does not decrease below the Safety Umit MCPR at any time during the transient. I i

Amendment No. $$,Q4,74,p6 15

- - - ~ . . . _ _ _ __ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ - -_ ._

JAFNPP

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2.1 BASES (Cont'd) A. Trip Settings l The MCPR operating limits in the COLR are conservatively assumed to The bases for indnndual trip settings are discussed in the exist prior to initiation of the transients. following paragraphs.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces rnore pessimistic

1. n W ngs answers than would result by using expected values of control
a. IRM Flux Scram Trip Setting parameters and analyzing at higher power levels.

Steady-state operation without forced recirculation is not pemwtted. The sys analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps. IRM is a 5-decade instrument which covers the range of power level between that covered by the in summary: SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5

. The abnormal operational transients were analyzed to the decades are broken down into 10 ranges, each licensed maximum power level. being one-half of a decade in size. The IRM scram

. The licensed maximum power level is 2436 MWt. trip setting of 120 divisions is actrve in each range of the IRM. For example,if the instrument were on

. Analyses of transients employ adequately conservative values of Range 1, the scram setting would be a 120 divisions

, the controlling reactor parameters. for that range; likewise, if the instrument were on l . The analytical procedures now used result in a more logical range 5,the saam w@ M 120 h on M l range. Thus, as the IRMis ranged up to I answer than the attemative method of assuming a higher starting amin @e theincrease b m y,y scram I power in conjunction with the expected values for the parameters. tnp setting is also ranged up. The most sig miceiit sources of reactivity change dunng the power i

increase are due to control rod withdrawal. For

insequence control rod withdrawal, the rate of I

change of power is slow enough due to the physical l limitation of withdrawing control rods, that heat flux .I is in equilibrium with the neutron flux and an IRM i scram would result in a reactor shutdown well before any Safety Umit is exceeded.

l l

i l Amendment No.1/, if,2f,36 16 4

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a JAFNPP ..

2.1 BASES (Cont'd) from the MCPR operating limits provided in the Core Operating Umits Report.

c. APRM Flux Scram Trip Setting (Run Mode) (cont'd) '
d. APRM Rod BlockTrip Setting -

J rated power. This reduced flow referenced trip setpoint will result in an eariier scram during slow thermal # E0*" *8

  • W a W Wng thWh hata De 6 l transients, such as the loss of 80*F feedwater heating event, than would result with the 129% fixed high neutron system pre a cadml rod Mto pm W awa%W a given point at constant eh flux scram trip. The lower flow referenced scram setpoint ,

l therefore decreases the seventy (AOPR) of a skw thermal ate,W h ph an W W @h me Saan Ms md M@ W M h transient and allows lower Operating Umits if such a transient is the limiting shuird operational transient M h W h ate, I atdhg pen s anW @W reactw poww M to enaease during a certain exposure interval in the cycle.

excessive values due to control withdrawal. The flow The APRM fixed high neutron flux signal does not variable trip setting parallels that of the APRM Scram and incorporate the time constant, but responds directly to provides margin to scram, assuming a steady-state instantaneous neutron flux. ' This scram setpoint scrams operation at the trip setting, over the entire recirculation the reactor during fast power increase transients if credit is flow range. The actual power distribution in the core is  !

not taken for a direct (position) scram, and also serves to established by specified control rod sequences and is scram the reactor if credit is not taken for the flow monitored continuously by the in-core LPRM system. As referenced scram. with the APRM scram trip setting, the APRM rod block trip

' setting is adjusted downward if the maximum fractKm of The scram trip setting must be adjusted to ensure that the limiting power densdy exceeds the fract,on  : of rated power, LHGR transient peak is not increased for any combination of ; wimum fraction of limiting power density (MFLPD) thus preserving the APRM rod block margin. As with the i and reactor core thermal power. The scram setting is scam setting, Ws ma# aWished W a@ng the i adjusted in accordance with the formula in the Core APR Q n.

) Operating Umits Report, when the MFLPD is greater than

2. Reactor Water low Level Scram Trip Setting

, the fraction of rated power (FRP). This adjustment may be accomplished by either (1) reducing the APRM scram and The reactor low water level scram is set at a point which will l rod block settings or (2) adjusting the indicated APRM assure that the water level used in the Bases for the Safety Umit -

signal to reflect the high peaking condition. is maintained. The scram setpoint is based on normal operating Analyses of the limiting transients show that no scram }empaatum and pmsswe Mons h theIwel adjustment is required to assure that the MCPR will be instmmadah,on is hy canpensatn greater than the Safety Umit when the transient is initiated 4

, Amendment No. 49,119 18

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2.1 BASES (Cont'd)

C. References

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1. (Deleted) l
2. " General Electric Standard Application for Reactor Fuel",

. NEDE 24011-P-A (Approved revision number applicable at time that reload fuel analyses are performed).

3. (Deleted) l
4. FitzPatrick Nuclear Power Plant Single-Loop Operation, NEDO-24281, August,1980.

s T

Amendment No. 44, Q4,S6 S3 (Next pageis 23)

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3.1 UMITING CONDITIONS FOR OPERATION 4.1 SURVEllt.ANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Applicability:

Applies to the instrumentation and associated devices which initiate the Applies to the surveillance of the instrumerf.ation and associated reactor scram. devices which initiate reactor scram.

Objective: Objective:

To assure the operability of the Reactor Protection System. To specrfy the type of frequency of surveil!ance to be applied to the protection instrumentation.

Spacification: Specification:

A. The setpoints, minimum number of trip systems, minimum A. Instrumentation systems shall be functionally tested and number of instrument channels that must be operable for each calibrated as indicated in Tables 4.1-1 and 4.1-2 respectively.

position of the reactor modo switch shall be as shown on Table 3.1-1. The design system response time from the opening of the sensor contact to and including the opening of the trip actuator contacts shall not exceed 50 msec.

B. Maximum Fraction of Umiting Power Density (MFLrT.h '

B. Minimum Critical Power Ratio (MCPR) The MFLPD shall be determined daily during reactor power Du-ing reactor power operation, the MCPR operating limit shall operation at >25% rated thermal power and the APRM high flux l not be less than that shown in the Core Operating Umits Report. scram and Rod Block trip settings adj.ssted if necessary as required by Specifications 2.1.A.1.c and 2.1.A.1.d, respectively.

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Amendment No. 4ff,64,86,1p9 30f

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-JAFNPP 3.1 (cont'd) 4.1 (cont'd) -t

1. If anytime during reactor operation at greater than 25% of C. MCPR shall be determoed daily dunng reactor power operation .

rated power it is deta n s md that the operating limit MCPR at >25% of rated thermal power and following any change in is being exceeded, action shall then be initiated within power level or distribubon that would cause operation with a fifteen (15) minutes to restore operation to within the hrrwtog control rod pattem as desenbed in the bases for prescribed limits. If the MCPR is not retumed to within the Spa.irication 3.3.B.S.

prescribed limits within two (2) hours, an orderly reactor D. When it is determoed that a channel has failed in the unsafe power reduction shall begin immediately. The reactor
  1. h h RPS N M eh m -

i power shall be reduced to less than 25% of rated pows vanable shall be ft, n fdvi .6Hy tested immediately before the trip within the next four hours or until the MCPR is retumed t F -

a f is . h@W contammg the unsafe failure may be placed in the untnpped condition dunng the penod in which survesitance testmg is bemg patcn6=d on the other RPS chan es.

I E. WiTic,tJcni of the MCPR operating limits shall be performed in ..

accordarre with the Core Operating Umrts Report.

l i

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.s j Amendment No. 94,74,7585,96,109,1/7 31 i _

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3.1 BASES (cont'd)

Turbine control valves fast closures initiates a scram based on pressure switches sensirxj e;ectivhydraulic 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.

l 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 sutmttel for vanous core exposures are l given in the Core Operating Umits Report (COLR).

. The ECCS performance analyses assumed reactor operation

' will be limited to MCPR = 1.20, as described in NEDO-21682 and NEDC-31317P. The Technical Specificnivr6 limit i operation of the reactor to the more conservative MCPR based l on consideration of the limiting transient as given in the COLR.

i 6

Amendment No. 4Ef, C/,1ps 35

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TABLE 3.1-1 .. .

2i REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT-Minimum No. Mode in Wtuch Functon of Operable Must be Operable Total Number of instrument instrument Channels

Chamels Per Refuel Startup RurF Pronded by Desen Achon
Trip System (1) Trip Function Trip Level Setting (6)(16) for Both Trip Systems (1)'

1 Mode Switch in X X- X 1 Mode Switch .A Shutdown (4 Selections)

Manual Scram X X X A 1 2 Instrument Channels 3 IRM High Rux < 120/125 of X X 8 Instrument Creres A 4 full scale 3 IRM inoperative X X 8 Instrument Cieres A 2 APRM Neutron Flux- < 15% Power X X 6 Instrument Cieres A Startup (15) l 2 APRM Row Referenced (12) X 6 Instrument Chamels A or B Neutron Rux (Not to exceed l 117 %) (13)(14)

+

2 APRM Fixed High < 120% Power X 6 Instrument Char.nels A or B -

Neutron Rux (14) ,

2 APRM inoperative (10) X X X 6 Instrument Channels A or B Amendment No. 1)( GO,43,72, Ef, SEf,1;WI

. 41

, -, . ._, s - . -. . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _____ ____ _. .__.

+ ,

n.

JAFNPP ,.'..

TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REGNREMENT '

NOTES OF TABLE 3.1 (cont'd)

C. High Flux IRM.

. D. Scram Discharge Volume High Level when any control rod in a control cell contairwng fuel is not fully inserted.

I E. APRM 15% Power Trip.

l 7. Not required to be operable when primary containment integnty is not required.

8. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
9. The APRM downscale trip is automatically bypassed when the IRM Instrumentation is operable and not high.
10. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 11 LPRM inputs of the normal cun$ ni=1.
11. See Section 2.1.A.1.
12. The APRM Flow Referenced Neutron Rux Scram setting shall be less than or equal to the limit established in the Core Operating Lirruts Report.
13. The Average Power Range Monitor scram funcin6 varied as a functKm of recirculation flow (W). The trip settog of this funchon must be maintained in accordance with Specrfication 2.1.A.1.c.
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 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. i
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 Une Radiation Monitor Trip Level Setpoent will be raised to < three times the increased radiation levels. The test will be conducted at power levels > 80% of normal rated power. Dunng controlled

. power reduction, the setpoint wi!! 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.

  • This specification is in effect only during Operating Cycle 7.

Amendment No. [,% h,6)f,7/,7/,1p

s- a .O JAFNPP d1 .

e Figures 3.1 1 and 3.12 have boon doloted i.

b, b

Amendment No.14,1pf 47a-b

}

JAFNPP ,

TABLE 32-3

  • INSTRUMENTATION THAT INMATES CONTROL ROO BLOCKS - -

Minimum Nd.

of Operable Total Number of Instrument Instrument Ctewes Channels Per _ Provided by Design Trip System Instrument Trip LevelSetting for Both Cta es Achon i

l 2 APRM Upscale (Flow Biased) (8) 6 inst. Channels (1) 2 APRM Upscale (Start-up Mode) < 12% 6 Inst Cte es -

(1) 2 APRM Downscale 12.5 indicated on scale 6 Inst. Channels (1) l 1(6) Rod Block Monitor (Flow Biased) (8) 2 Inst. Ct= = es (1) 1(6) Rod Block Monitor (Downscale) 12.5 indicated on scale 2 Inst.Cie -es (1) 3 IRM Downscale (2) 12% of fullscale 8 Inst. Channels (1) 3 IRM Detector not in Start-up Position (7) 8 Inst.Cteaes -

(1) 3 IRM Upscale < 86.4% of full scale 8 Inst. Cta n es (1) 2(4) SRM Detector not in Start-up Position (3) 4 Inst. Channels (1)

SRM Upscale 5 2(4)(5) < 10 counts /sec 4 Inst. Channels (1) 1 Scram DischargeInstrument < 26.0 gallons per 2 Inst. Channe!s (9)(10)

Volume High Water Level instrument volume a NOTES FOR TABLE 3.2-3

1. For the Start-up and Run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function.

The SRM and IRM block need not be operable in run mode, and t

Amendment No. ff,62,7/,jMI 72

JAFNPP . .

TABLE 32-3 (Cont'd)

INSTRUMENTATION THAT INITIATES CONTROL ROD Bt.OCKS NOTES FOR TABLE 32.-3 the APRM and RBM rod blocks need not be operable in start-up mode. From and after the tirne it is found that the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped. From and after the time it is found that the first column cannot be met for both trip systems, the systems shall be tripped.

2. IRM downscale is bypassed when it is on its lowest range.
3. This function is bypassed when the count rate is > 100 cps.

4 One of the four SRM inputs may be bypassed.

5. This SRM Function is bypassed when the IRM range switches are on range 8 or above.
6. The trip is bypassed when the reactor power is < 30%.
7. This function is bypassed when the Mode Switch is placed in Run.
8. The Flow Biased APRM Upscale and Rod Block Monitor trip level setpoint shall be less than or equal to the limit established in the Core I Operating Umits Report.
9. When the reactor is subcritical and the reactor water temperature is less than 2127, the control rod block is required to be operable orff if any control rod in a control cell containing fuel is not fully inserted.
10. When one of the instruments associated with scram discharge instrument volume high water rod blocks is not operabic, the trip system shall be tripped.

Amendment No. 46,92,7E,76 73

, .a.: - ., ,.- .--,u -- - '+-'+ 'ew- ' ' "

']

O 4

9 9

4 JAFNPP . .

1 THIS PAGE INTENTIONALLY Bl.ANK g.

Amendment No. 36,96,96,y6 74

JAFNPP 3.3 and 4.3 BASES (cont'c')

rods have been withdrawn (e.g., groups A12 and 3A ,,it is This system backs up #m operata who witidraws cortol mds demonstrated that the Group Notch made for the control drives according to written sequences. The specified restnchons with is enforced. This daiTiciistration is made by paduirei6g the one chas ciel out of service conservatwely assure that fuel hardware functivnal test sequence. The Group Notch restraints desvinga will not occur due to rod withdrawal errors when this i are automatically removed above 20% power. condition exists.

Dunng reactor shutdown, similar surveillance checks shall be A limrting control rod pattem is a pattem which results in the made with regard to rod group availability as soon as core being on a thermal hydraulic limit (e.g., MCPR limit). l 7

automatic initiation of the RSCS occurs and subsequently at Dunrig use of such pin:,ii6,it is judged that teshng of tre RBM appropnate stages of the control rod insertron. System prior to withdrawal of such rods to assure its operability

4. The Source Range Monitor (SRM) System performs no e assue Mempmpw Wawh nd occw. Mis #m i automatic safety system furschon; i.e., it has no scram funchon.

r d the h W to % % W lt does provide me operator with a visual indication of neutron Ws W h 6 mds &h #m W are level. The cv6 sequences of reactmty accidents are functions W w as #my W Mo N occurence d of the initial neutron flux. The requirement of at least 3 counts '" P -

per sec. assures that any transient, should it occur, begins at EU' "'e.'

i #may'" W M h or above the initialvalue of 10-8of rated power usedin the C. Scram insertion Times analyses of transient cold conditions. One operable SRM cu is ciai would be adequate to monitor the iipgvacu to h Md Rod @ mis Wned to % N re i^ a rWe fast W to FM M ATiage,i.e., to enticality using noTic seous pattems of squattered control prevent Um MCPR from baevi ig less than tie SafMy Umst.

rodwithdrawal. Aminimumof twooperableSRM'sare Scram insertion time test criteria of Section 3.3.C.1 were used pmvided as an added consenatism. '

to generate the genene scram reactmty curve shown in

5. The Rod Block Monitor (RBM) is designed to auturnatically NEDE-24011-P-A. This genenc curve was used in analysis of
prevent fuel damage in the event of erroneous rod withdrawal non-pressunzation transeents to determine MCPR limits.

from locations of high power density during high power level Thorefore, the required protechon is provided.

4 operation. Two chdieis are provided, and one of these may be bypassed from the console for maintenance and/or testing 2

Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.

4 1

Amendment No. 4ff,56,f 102 i

l

JAFNPP 3.5 (cont'd) 4.5 (cont'd) condition, that pump shall be evihed inoperable for 2. Followng any period where the LPCI subsystems or core purposes of satisfying Specifications 3.5.A. 3.5.C, and spray subsystems have not been mamtam3d in a filled 35.E. condition; the discharge pipeg of the affected subsystem shall be vented from the high pomt of the system and H. Average Planar Unear Heat Generation Rate (APLHGR) water flow observed.

During power operation, the APLHGR for each type of fuel as a 3. Whenever the HPCI or RCIC System is lined up to take function of axiallocation and average planar exposure shall be suction from the condensate storage tank, the discharge within limits based on applicable APLHGR limit values wtuch piping of the HPC: or RCIC shall be vented from the high have been approved for the respective fuel and lattice types. point of the system, and water flow observed on a monthly l These values are provided in the Core Orm aG.g Cunits Report. bases.

If at anytime during reactor power operation greater than 25% of

4. The level switches located on the Core Spray and RHR rated power it is determined that the limitog value for APLHGR is g gy 3  %

being exceeded, action shall then be initiated withm 15 minutes Med M g e se M M M to restore operation to within the presenbed limits. If the APLHGR is not retumed to within the prescribed limits within two 1 (2) hours, an orderly reactor pcwer reduction shall be H. Average Planar Unear Heat Generation Rate (APLHGR) commenced immediately. The reactor power shall be reduced to less than 25% of rated power withm the next four hours, or The APLH3R for each type of fuel as a function of average until the APLHGR is retumed to within the prescribed limits. planar exposure shall be determoed daily during reactor

operation at >25% rated thermal power.

1 4

I Amendment No. 46,64,74,86,96,10s, lyt,1)f,184' 123 L

~ _ . _

JAFNPP 3.5 (cont'd) 4.5 (cont'd)

1. Unear Heat Generation Rate (LHGR) 1. Unear Heat Generation Rate (U4GR)

The knear heat generation rate (LHGR) of any rod in any fuel The LHGR shall be determined daily during reactor operation at esimisiliy at any axial locahon shall not exceed the maximum >25% rated thermal power.

l allowable LHGR given in the Core Operating Umrts Report i If anytime dunng reactor power operation grsater than 25% of rated power it is determined that the limiting value for LHGR is being exceeded, action shall then be initiated withm 15 rrunutes .

l to resto e operation to within the prescribed lirruts. If the LHGR is not retumed to withm the presuibed hrruts within two (2) i hours, an orderly reactor power reduchon shal: be cuiivis >ced i

immediately. The reactor powc shall be reduced to less than  !

25% of rated power within the next four hours, or until the LHGR is retumed to within the presenbed limits.

i 2

1 i

4 i Amendment No. 4ti,Q4,7j(,1p6

, 124 ir ec.-- or ... c ., . .: w e .#-- _____,______._.._m_______m

- -m_. . _ m_-_m.__m_ . _.u_m_.._m _ _ - _ _ _ __m __.m_______- .__m___.

JAFNPP 3.5 BASES (cont'df requirements for the emergency diesel generators. the calculated peak clad temperature by less than + 2(TF G. Maintenance of Filled Discharge Pipe r to the Wanpaabe fw a tyW W W,h limit on the average linear heat generation rate is sufficient to if the discharge piping of the core spray, LPCI, RCIC, and HPCI assure that calculated temperatures are within the 10 CFR 50 are not filled, a water hammer can develop in this piping when Appendix K limit. The limiting values for APLHGR are given in the pump (s) are started. To minimize damage to the discharge the Core Operating Limits Report. During Single Loop piping and to ensure added margin in the operation of these Operation a multipreer is applied to these values. The derivation systems, this technical specification requires the discharge of this multiplier can be found in Bases 3.5.K, Reference 1.

lines to be filled whenever the system is required to be operable. If a discharge pipe is not filled, the pumps the supply

1. Linear Heat Generation Rate (LHGR) that line must be assumed to be inoperable for technical This specification assures that the linear heat generation rate in specification purposes. However,if a water hammer were to any rod is less than the design linear heat generation.

occur, the system would still perform its design function.

The LHGR shall be checked daily during reactor operation at H. Average Planar Unear Heat Generation Rate (APLHGR) 25% rated thermal power to determine if fuel bumup, or control This specification assures that the peak cladding temperature md m np w strh. 6 following the postulated design basis loss-of-coolant accident R to M aIMng value W 25% raw W poww, will not exceed the limit specified in 10 CFR 50 Appendix K. rat,o oM y to amage NR w@ Mo M i

greater than 10 which is precluded by a considerable margin The peak cladding temperature following a postulated loss-of- when employing any permissible control rod pattem.

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 distribution within a fuel assembly affect Amendment No. jid,-d, Slf,96,1p5,14 130

JAFNPP - .

Figures 353 through 3.5-14 have been deleted Amendment No. 3[,[4,7/

135a - 1351

JAFNPP 5.0 DESIGN FEATURES 5.3 REACTOR PRESSUREVESSEL 5.1 S E The reactor pressure vessel is as described in Table 42-1 and A. The James A. FitzPatrick Nuclear Power Plant is located on the 4 - .

eW gn are hMm PASNY portion of the Nine Mile Point site, approximately 3,000 ection 42 d tM GR.

ft. east of the Nine Mile Point Nucicar Station, Unit 1. The NPP-JAF site is on Lake Ontario in Oswego Country, New York, 5.4 NAINMENT approximately 7 miles northeast of Oswego. The plant is located at coordinates north 4,813,545.012 m, east 386,968.945 m, on A W i Wh ist W m the Un,iversal Transverse Mercator System. rimary containment are given in Table 52-1 of the FSAR.

B. B. e nmentis as MMm W SM The nearest point on the property line from the reactor building and any points of potential gaseous effluents, with the exception of the lake shoreline, is located at the nort.*ast comer of the A~

property. This distance is approximately 3,200 ft. and is the C. Penetrations of the primary containment and piping passing radius of the exclusion areas as defined in 10 CFR 100.3. through such penetrations are designed in accordance 'with standards set forth in Section 52 of the FSAR.

52 REACTOR 5.5 FUEL STORAGE A. The reactor core consists of not more than 560 fuel assembiios.

The fuel types present in the core are listed in the Core A. The new fuel storage facility design criteria are to maintain a Kg Operating Umits Report. dry <0.90 and flooded <0.95. Compliance shall be verifimi prior B. The reactor core contains 137 cruciform-shaped control rods as to WrNe of any new W W to tWW described in Section 3.4 of the FSAR.

Amendment No. 36,4( 46,94,66,74,18),1 4 245

l 1

.. + .-

mpp i

I i

(A) ROUTINE REPORTS (Continued) i

4. CORE OPERATING LIMITS REPORT The core operating limits shall be established and documented in the Core l Operating Umits Report (COLR) before each reload cycle or any remaining part of I a reload cycle. The analytical bases used to determine the core operating limits  !

shall be those previously reviewed and approved by the NRC as described in:

a. " General Electric Standard Application for Reactor Fuel,* NEDE 24011 P, latest approved version and amendments,
b. ' James A. FitzPatrick Nuclear Power Plant SAFER /GESTR LOCA Loss-of Coolant Accident Analysis,' NEDC-31317P, October,1986 including latest errata and addenda.
c. " Loss-of Coolant Accident Analysis for James A. FitzPatrick Nuclear Power i Plant," NEDO-21662 2, July,1977 including latest errata and addenda.

)

l The core operating limits shall be determined so that all applicable Ilmits (e.g., fuel l thermal-mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits l such as shutdown margin, and transient and accident analysis limits) of the safety I analysis are met. The COLR, including any mid-cycle revisions and supplements  ;

thereto, shall be provided upon issuance, for each reload cycle, to the NRC 1 Document Control Desk with copies to the Regional Administrator and Resident l Inspector.

]

. i 1

l j

l L

I l

l l \

L l

l l

l Amendment No.

254-c

JAFNPP p- ,

P I

o I

?

?

y l-(THESE PAGES INTENTIONALLY BLANK) t

.t i

i r-1-

l L

Amendment No. 32,110 254 d thru 254 f h

1 ,

,. t -

C i

ATTACHMENT ll SAFETY EVALUATION FOR PP,0 POSED TECHNICAL SPECIFICATION CH ANGES REGARDING

~

REMOVAL OF CYCLE SPECIFIC PARAMETER LIMITS l JPTS 88-020 ,

F i

x l

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50 333 l

l 1 .

o - 4 ,_, ,,r~- - ~ys---

Attachment il

, . SAFETY EVALUATION Page 1 of 9

1. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A. FitzPatrick Technical Specifications are contained in Attachment I and are described below. In addition to these changed pages, all text from the following pages has been either relocated or deleted, and these pages should be removed from the Technical Specifications: Ba, 10a, 31 a, 43a, 47b, 135a, 135b, 135c, 135d,135e,135f,135g,135h,1351,135),135k, and 135!.

Page vil, Ust of Figures Figures 3.1 1 and 3.12 are deleted and the pages combined.

Figures 3.5-3 through 3.514 are deleted and the pages combined.

Page 6, Specifications 1.0.U.1 and 2 ,

insert " Minimum value of the* to the definition of Minimum Critical Power Ratio.

Replace *as calculated by application of the GEXL correlation (Reference NEDE 10958)*

- with, "for all fuel assemblies in the core."

Delete "The design LHGR is 14.4 KW/ft for GE8x8EB fuel and 13.4 KW/ft for the remaindet."

Page 6a, Specifications 1.0.AD ,

A new definitionis added to read:

l: AD. Core Operating Umits Report (COLR)

This report is the plant specific document that provides the core operating limits for the current operating cycle. These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification -

6.9.A.4. Plant operation within these operating limits is addressed in individual Technical Specifications.

Page 8, Specification 2.1.A.1.c.(1)

L The scram trip setting formula and remainder of the column are replaced with, *less than or equal to the limit established in the Core Operating Umits Report (COLR). This limit must be adjusted for single loop operation as specified in the COLR.* -

l' Page 8a. Specification 2.1.A.1.c.(1) (cont'd) l Relocate this specification to page 8 and remove this page.

Page 9, Specification 2.1.A.1.c.(1) (cont'd)

Delete this specification in its entirety.

Y k

i

- =

Attachment il SAFETY EVALUATION Pap 2 d e  ;

PaDe 10, Speelfication 2.1.A.1.d l The Rod Block trip setting formula and remainder of the column are replaced with, *less than or equal to the limit established in the Core Operating Umits Report (COLR). This limit must be ad!usted for single loop operation as specified in the COLR.* ,

This specification is relocated to page 9 and page 10 is now intentionally blank. .

Page 12, Bases 1.1.A Replace *MCPR operating conditions in specification 3.1.B* with, *MCPR operating limit in the Core Operating Umits Report."

Page 13, Bases 1.1.A FIRST PARAGRAPH Replace *at the beginning of each fuel cycle

  • with, "in Reference 1.*

FOURTH PARAGRAPH '

I Replace "to a maximum LHGR of 14.4 KW/ft for GE8x8EB fuel and 13.4 KW/ft for the remainder

  • with, 'by the maximum LHGR Identified in the Core Operating Umits Report."

FIFTH PARAGRAPH Insert an additional "0" into *1.0* to make the number of significant figures consistent with the specifications.

Replace ' specifications 2.1.A.1.c and 2.1.A.1.d* with, "the Core Operating Umits Report.' .

Page 14, Bases 1.1.E Replace Reierence 1. with, " General Electric Standard Application for Reactor Fuel, NEDE 24011 P, latest approved revision and amendments.*

Delete Reference 3.

Page 15, Bases 2.1 '

FIRST PARAGRAPH Replace *2535' with *of 2436.* .

I Insert *MWt* after *2436.*

THIRD PARAGRAPH insert " applicable" before " operating Umit." >

Replace *MCPR's" with *MCPR.* ,

Replace " Specification 3.1.B* with, "the Core Operating Umits Report COLR.* ,

FOURTH PARAGRAPH l

Replace " Specification 3.1.B* with, "the Core Operating Limits Report.'

s

Attachment il  !

SAFETY EVALUATION Page 3 of 9 FIFTH PARAGRAPH Replace 'a' with "the," replace ' program

  • with
  • programs," and replace
  • References 1 end ,

3' with " Reference 2.*

Page 16, Bases 2.1 (cont'd)

FIRST PARAGRAPH Replace "of specification 3.1.B* with, "in the COLR.*

  • THIRD PARAGRAPH I Replace *will not be* with *is not."

In subparagraph (I), replace 'a power level of 2535 MWt* with, 'the licensed maximum power level.'

Replace the roman numerals of subparagraphs i through IV with bullets (*).

Page 18, Bases 2.1.A.1.c

, THIRD PARAGRAPH Replace

  • Specification 2.1.A.1.c' with, 'the Core Operating Umits Report.' '

FOURTH PARAGRAPH Replace

  • Spec!fication 3.1.B* with, "the Core Operating Umits Report."

Page 20, Bases 2.1.C l Delete References 1. and 3.

In Reference 2, replace

  • Fuel Application
  • with, " Standard Application for Reactor Fuel.*

I Page 30f, Specification 3.1.B and 3.1.B.1 Replace *Below:* and the subsequent specification 3.1.B.1 with, "in the Core Operating l Umits Report.'

l ,

Page 31, Specification 3.1.B.1 (cont *d) and,5.1.B.2 Delete these specifications in their entirety.

4.1.E Replace this specification and all subparagraphs with, " Verification of the MCPR operating limits shall be performed in accordance with the Core Operating Umits Report."

l Page 31a, Specifications 3.1.B.2 (cont'd),3.1.B.3,4 and 5, and 4.1.E.3 Deleted the note associated with Specification 3.1.B.2.

l 3.1.B.3 and 4 Delete these specifications in their entirety.

Attachment 11

. SAFETY EVALUATION  !

Pap 4 d 9 3.1.B.5 '

]

Replace

  • limiting value for* with, ' operating Umit." l Renumber this specification 3.1.B.1 and relocate to page 31. l 4.1.E.3 Delete this specification in its entirety. I Ramove page 31a from the specifications.

Page 35, Bases 3.1.B Replace " Specification 3.1.B' with, "the Core Operating Umits Report (COLR)."

Replace " Specification 3.1.B' with, 'the COLR.*  !

1 Page 41, Table 3.1 1  ;

Replace the Trip Level Setting formula for the APRM Flow Referenced Neutron Flux Scram I with *(12)."

]

Delete the reference to notes (12) and (17) from the Trip Function for the APRM Flow Referenced Neutron Flux Scram, i

Page 43, Notes of Table 3.1 1 (cont'd)

Replace note 12 with the following:  !

s 12. The APRM Flow Referenced Neutron Flux Scram setting shall be less than or equal to the limit established in the Core Operating Umits Report.  ;

Page 43a, Notes of Table 3.1 1 (cont'd)

Delete note 17 in its entirety.

Relocate notes 14 through 16 to page 43 and remove page 43a from the Technical Specifications.

Pages 47a and 47b., Figures 3.1 1 and 3.12 ,

Delete both figures and combine the pages into a single page 47a-b.

i Page 72, Table 3.2-3 Replace the Trip Level Setting formulas for both Flow Blased APRM Upscale and Rod Block Monitor Control Rod Blocks with, *(8)."

Page 73, Notes for Table 3.2 3 Replace note 8 with the following:

i

( )

' ~

Attachment ll l

. SAFETY EVALUATION l Pap 5o 9 i

8. The Flow Blased APRM Upscale and Rod Block Monitor trip level setpoint shall be less than or equal to the limit established in the Core Operating Umits Report.

Page 74, Notes for Table 3.2 3 (cont'd)  ;

Delete notes 11 and 12 in their entirety.  :

Remove the headings from this page and insert *This Page intentionally Blank."

Page 102, 3.3 and 4.3 Bases,6B.5 THIRD PARAGRAPH Replace 'i.e., MCPR limits as shown in Specification 3.1.B" with, "e.g., MCPR limit."  ;

Page 123, Specification 3.5.H Replace the second and third sentences with, "These values are provided in the Core Operating Umits Report."

in specifications 4.5.G.2 and 3 on this page, restore the Amendment 132 changes inadvertantly deleted by Amendment 134.

Page 124, Specification 3.5.1 Replace "of 14.4 KW/ft for GE8x8EB fuel and 13.4 KW/ft for the remainder of the fuel"  ;

with, "given in the Core Operating Umits Report."

Specification 4.5.1 Replace

  • checked
  • with " determined" to accurately reflect that the LHGR is a calculated value, not an instrument reading.

, Page 130, Bases 3.5.H SECOND PARAGRAPH In the third sentence, replace ' Figures 3.5-11 through 3.514" with, "the Core Operating Umits Report.'

Delete the fourth and fifth sentences in their entirety.

Replace the sixth sentence through the word 'during" with, 'A multiplier is applied to these values during."

L Page 135a through 1351, Figures 3.5-3 through 3.5-14 These figures have been deleted and the pages combined into a single page 135a-1351.

l

l. Page 245, Specification 5.2.A Delete the second sentence through the end of this Specification. In their place insert, l "The fuel types present in the core are listed in the Core Operating Umits Report."

7'o-Attachment ll

- SAFETY EVALUATION Page 6 of 9 j i

Page 254 c, Specification 6.9.A.4 '

Insert a new Specification 6.9.A.4 on a new page 254 c. The text of this specification is given in Attachment 1. ,

Page 254 c thru 254 f Renumber this page *254-d thru 254 f' to support the change described above.

i ll.

PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed Technical Specification changes is to remove cycle specific ,

parameter limits in accordance with the guidance provided by the NRC in Generic Letter 8816 (Reference 1). Use of the Generic Letter 88-16 alternative consists of three separate actions to modify the Technical Specifications: l

1) The addition of a definition of a formal report that includes the values of cycle-specific parameter limits that have been estab!!shed using an NRC approved i methodology and consistent with all applicable limits of the safety analysis. At FitzPatrick, the report will be titled, ' Core Operating Umits Report.*
2) The addition of an administrative reporting requirement to submit the Core Operating Umit Report to the NRC for information.
3) The modification of individual Technical Specifications to note that cycle specific parameters shall be maintained within the limits provided in the Core Operating '

Umits Report.

,c ' The proposed Technical Specification changes are responsive to industry and NRC efforts to improve Technical Specifications, reduce the administrative burden on the NRC and the l

New York Power Authority, and permit future reloads to be accomplished without license amendments. The proposed changes are consistent with those discussed previously between the NRC and General Electric Co. as described in Reference 2.

The following Technical Specification parameters have been identified as cycle specific limits that can be relocated to the Core Operating Umits Report:

1) Operating Umit Minimum Critical Power Ratio (MCPR);
2) Flow Dependent MCPR Umits;
3) Maximum Average Planar Unear Heat Generation Rate (MAPLHGR);
4) Unear Heat Generation Rate (LHGR);
5) Flow-blased Average Power Range Monitor (APRM) and Rod Block Mor.itor (RBM) settings; and
6) Fuel design features.

l C

'. Attachment ll sarmavatuaTion i Page 7 of 9 In addition, discussions contained in the Techn! cal Specification Bases associated with the above parameters which are cycle-specific are modified in accordance with the guidance of Generic Letter 88-16.

The Authority is implementing those Generic Letter 8816 changes during the Reload  !

9/ Cycle 10 refueling outage. The Authority will prepare a Core Operating Limits Report i (COLR) to support the reloaded core. The Cycle 10 COLR will be provided to the NRC l upon issuance, but no later than at the startup of Cycle 10 as required by the proposed Technical Specifications. The core will be operated for the remainder of the current operating oycle with the Cycle 9 specific limits contained in the Technical Specifications i and will commence Cycle 10 with the Cycle 10 specific limits in a PORC and SRC j reviewed COLR. At no time will the core be operated without the cycle-specific limits in )

either the Technical Specifications or the COLR.

As part of this Technical Specification amendment, an additional change is also proposed.

The bases for Specification 2.1 on pages 15 and 16 state that the abnormal operational .

transients were analyzed at a power of 2535 MWt, correspondin0 to 104 percent of the licensed maximum power level of 2436 MWt. However, the NRC has approved the GE ,

transient analysis methods designated GEMINI methods, which use the nominal (100%) I power level in transient analyses. Consequently, the Bases to Specification 2.1 are i modified to state that transient analyses are performed at 100 percent power (the maximum licensed power level) consistent with the NRC approval given in Reference 3. )

This rnethod of transient analysis was approved for FitzPatrick Cycle 8 operation in j l Amendment 109 to the Technical Specifications (Reference 4). l 1

l 111. IMPACT OF THE PROPOSED CHANGES I A. Generic Letter 88-16 Changes l

l The current method of controlling reactor physics parameters to assure conformance with 10 CFR 50.36 is to specify the values determined to be within specified acceptance criteria, usually the limits of the safety analysis, using an approved calev5 tion i '

methodology. The proposed Technical Specification changes maintain control of the values of cycle specific parameters and assure conformance to 10 CFR 50.36 by

,. specifying the approved calculation methodology and approved acceptance criteria. The ,

L Core Operating Umits Report documents the specific values of parameter limits that are determined using these methods and that meet the acceptance criteria. The Technical I. Specifications continue to require that operation will remain within limits, and that required I

remedial actions are taken if the limits are not met.

The Core Operating Umits Report for each cycle, and any necessary mid cycle revisions, will be provided to the NRC for information. This report will be reviewed by both PORC and SRC to provide a similar level of quality assurance and document control for the Core Operating Umits Report as for the Technical Specifications. This will ensure that the proper operating limits are being enforced.

l B. Technical Specification Bases 2.1 Change This change updates the Bases to reflect the power level used in the FitzPatrick tranent I

analyses. With the introduction of the approved GEMINI methods, transient analyses are performed at the 100 percent power level. Previously, analyses were performed at a power level in excess of 100 percent to account for uncertainties in power level

F

      • Attachment 11 SAFETY EVALUATION Page 8 of 9 measurement as required by Regulatory Guide 1.49. However, with GEMINI methods, power level measurement uncertainty is accounted for instead by increasing the MCPR calculated with the GEMINI meth@ instead of the power level as used previously. The NRC has generically approved this method of accounting for power level measurement uncertainty in Reference 3 and has approved its use at FitzPatrick in Reference 4.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance v*%ne proposed Amendment would not involve a significant hazards consideration as defi.4d in 10 CFR 50.92 since it would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

A. Generic Letter 8816 Changes; The proposed amendment merely moves cycle specific parameter limits from the Technical Specifications to the Core Operating Umits Report. NRC approved methodologies will continue to be used as the basis for establishing those limits. The establishment of these limits in accordance with NRC-approved methodology and the incorporation of these results into the Core Opraating Umits Report will ensure that proper steps have been taken to establish the values of these limits. Furthermore, the submittal of the Gore Operating Umits Report to the NRC will allow the staff to continue to trend the values of these limits.

B. Technical Specification Bases 2.1 Change: j The use of 100 percent power in the analysis of abnormal operational I transients using GEMINI methods has been reviewed and approved previously i by the NRC for both generic and FitzPatrick specific application (see j References 3 and 4). Power level measurement uncertaintles are accounted for adequately in the MCPR Operating Umit, and the level of confidence that 1 the MCPR Safety Umit will not be violated as a result of a transient is not reduced.

i

2. create the poasibility of a new or different kind of accident from any i accident previously evaluated. j No safety related equipment, function, or plant operation will be altered as a  !

result of the proposed changes. The changes do not create any new accident mode. The level of document control and quality assurance applied to the  !

preparation and use of the Core Operating Umits Report will be equivalent to l that applied to Technical Specifications.

3. Involve a significant reduction in a margin of safety.

A. Generic Letter 8816 Changes:

l The proposed changes are administrative in nature and do not impact the I operation of the plant in a manner that will reduce the margin of safety. The proposed amendment still requires operation within the limits determined ,

',' Attachment ll g SAFETY EVALUATION Page 9 of 9 usirg NRC-approved methods, and that appropriate remedial actions be taken if the limits are violated.

B. Technical Specification Bases 2.1 Change:

The MCPR Operating Umit continues to be determined using an approved methodology that conservatively accounts for power level measurement uncertainties. The same criterion for acceptable operation is maintained; that is,99.9 percent of all fuel rods will not enter boiling transition in the event of the limiting transient. Therefore, the margin of safety is not reduced.

V. IMPLEMENTATION OF THE PROPOSED CHANGE Implementation of the proposed changes will not impact the Al. ARA or Fire Protection Programs at the FitzPatrick plant, nor will the changes 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:

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a. will not change the probability nor the conscquences 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; and
d. involves no significant hazards consideration, as defined in 10 CFR 50.92.

Vll. REFERENCES

1. NRC Generic Letter 88-16, " Removal of Cycle Specific Parameter Umits froni Technical Specifications," dated October 4,1988.
2. GE letter, J. S. Charnley to M. W. Hodges (NRC),
3. NRC letter, G. C. Lainas to J. S. Charnley (GE), " Acceptance for Referencing of Ucensing Topical Report NEDE 24011 P A, 'GE Generic Ucensing Reload Report,' Supplement to Amendment 11,* dated March 22,1986.

l 4. NRC letter, H. l. Abe! son to J. C. Brons (NYPA),

  • Amendment 109 to Technical Specifications," dated April 3,1987.

l l 5. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis L Report.

! 6. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), '

dated November 20,1972, and Supplements.

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