ML20154D252

Proposed Tech Specs Increasing Authorized Max Power Level by Approx 4.1% to 2536 Mwt from Current Limit of 2436 Mwt

ML20154D252
Person / Time
Site:	FitzPatrick
Issue date:	06/12/1992
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20154D245	List: JPN-92-028, Forwards Application for Amend to License DPR-59,increasing Authorized Max Power Level by Approx 4.1% to 2536 Mwt from Current Limit of 2436 Mwt.Proprietary TRs NEDC-32016P & NEDC-31317P-1,encl.Without Proprietary TRs ML20154D246 ML20154D252
References
NUDOCS 9810070176
Download: ML20154D252 (115)

Text

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ATTACHMENT l to JPN 92-028 4

PROPOSED TECHNICAL SPECIFICATION CHANGES POWER UPRATE (JPTS 91025)

NewYork Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No.54333 DPR-59 9610070176 920612?

PDR ADOCK 05000333 P PDR;

i i

, , JAFNPP  !

1.0 (cont'd)  !

C. Cold Condition - Reactor coolant temperature <212*F. 3. Instrument Channel - An instrument channel means an D. arranguinent of a sensor and auxiliary equiprnent required i Hot Standby Condition - Hot Stan@y condition means operation with coolant temperature >212"F, the Mode Switch in to generate and transmit to a trip system a sangle trip  !

signal related to the plant parameter monitored by that l Startup/ Hot Standby and reactor pressure <1,040 psig. Instrument channel.

i E. Immediate - Immeciate means that the required action will be 4. Instrument Check - An instrument check is a qualitative i irutiated as soon as practicable considering the safe operation d determination of acceptable operatWity by observation of the unit and the importance of the required action. instrument behavior dunng operation. This determinahon ,

F. Instrumentation shan include, where possible, conyarison d the '

instrument with other independent instruments measunng the same vanable.

1. Functional Test - A functional test is the manual operation  ;

or initiation of a system, subsystem, or component to 5. Instrument Channel Functional Test - An instrument  !

verify that it functions within design tolerances (e.g., the channel functional test means the injection of a simulated manual start of a core spray pump to verify that it runs and signal into the instrument pnmary sensor where possible that it pumps the required volume of water).

to venfy the proper instrument channel response, alarm i

2. Instrument Channel Cahbration - An instrument channel cahbration means the adjustment d an instrument signal 6. Logic System Function Test - A logic system functional test output so that it corresponds, within acceptable range, and means a test of relays and contacts of a logic circuit from accuracy, to a known value(s) of the parameter which the sensor to activated device to ensure cunycnents are  !

instrument monitors. Calibration shall . encompass the operable per design intent. Where practicable, action will entire instrument channel including m* N, alarm or trip. go to cowyecs i.e., pumps t

9 1

. JAFNPP-1.0 (conid) opened to perform necessary operahonal activites. R. Safety Umits - The safety lunts are Imts withm wtuch the reasonable maintenance of the fuel ci&M;ig integnty and the i

. 2. At least one door in each airlock is closed and sealed. - reactor coolant system integnty are namrod. _ Violation of such a .

limit is ra ma for unit shutdown and review by the Atomic Energy  !

3. All automatic containment lardmeion valves are operable or Commission before resumption of unit operation. Operation '

.de-actuated in the isolated position. beyond such a limit may not in itself result in senous '

consequences but it indicates an operahonal deficiency subject >

4. All blind flanges and manways are cloeod. to regulatory review. l t

N. Rated Power - Rated power refers to operation at a reactor S. Secondary Containment integnty -

Secondary contamment I power of 2,536 MWt. This is also termed 100 percent power and integrity means that the reactor building is intact and the t is the maximum power level authorized by the operating k:ense. following conditions are met:

Rated steam flow, rated coolant flow, rated nuclear system pressure, refer to the values of these parameters when the 1. At least one door in each ama=* opening is closed.

l reactor is at rated power (Reference 1). -

2. The Standby Gas Treatment System is operable.

O. Reactor PowerOperation - Reactor power operation is any operation with the Mode Switch in the Startup/ Hot Stan@y or 3. All automahc venblabon system isolation valves are Run position with the reactor entical and above 1 percent rated operable or secured in the isolated position.

thermal power.

T. Surveillance Frequency- Penodic P. ReactorVesselPressure - Unless otherwise irhad, reactor vessel pressures listed in the Technical Specificahons are those measured by the reactor vessel steam space sensor.

Q. RefuehngOutage - Refuehng outage is the period of time between the shutdown of the unit prior to refueling and the startup of the Plant subsequent to that refuehng.

Amendment No. f, If,

JAFNPP Z. Top of Active Fuel AD. Core Operating Limits Report (COLR)

The Top of Active Fuel, corresponding to the top of the ennched This report is the plant-specific _ document that provides the core fuel column of each fuel bunde, is located 352.5 mches above operating Istruts for the current operating cycle. These cycle-vessel zero, which is the lowest point in the inside bottom of the spedric operating limits shall be.determmed for each reload reactor vessel. (See General Electric drawmg No. 919D890BD.) cycle in accordance with Specification 6.9.AA. Plant operation withm these operating limits is addressed in individual Techrucal AA. Rod Density S;+:r-:4x=.

Rod density is the number of control rod notches maerted AE. ' References expressed as a frachon of the total number of control rod notches. All rods fully inserted is a condebon represenhng 100 1. General Electric Report NEDC-32016P,

• Power Uprate percent rod density. Safety Analysis for the James A. FitzPatnck Nuclear Power Plant," December.1991 (proprietary).

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

AC. Ventog Venting is the controlled process of releasmg air or gas from a confinement in such a inanner that replacement air or gas is not provided or required.

Amendment No. }d, ,1 ,

JAFNPP 2.1 (cont'd)

2. Reactor Water Low Level Scram Trip Setting Reactor low water level scram setting shall be >177 in. i above the top of the active fuel (TAF) at normai @ rating conditions.

3. Turbine Stop Valve Closure Scram Trip Setting Turbine stop valve scram shall be <10 percent valve closure from full open when the reactor is at or above 29%

of rated power.

4. Turbine Control Valve Fast Closure Scram Trip Setting Turbine control valve fast closure scram control oil i pressure shall be set at 500 <P <850 psig.

l

5. Main Steam Une isolation Valve Closure Scram Trip Setting Main steam line isolation valve closure scram shall be < 10 t percent valve closure from full open

6. Main Steam Une isolation Valve Closure on Low Pressure When in the run mode main steam line low pressure initiation of main steam line isolation valve closure shall be

>825 psig.

Amendment No. [, , ,

JAFNPP BASES 2.1 FUELCLADDINGINTEGRi1Y limit, it is required that the resultog MCPR does not decrease below the Safety Umit MCPR at any time dunng the transient.

The abnormal operabonal transients applicable to operabon of the FitzPatnck Unit have been analyzed throughout the spectrum The most imtog transients have been analyzed to determine -

of planned operating conditions up to the thermal power which result in the largest ratt rhon in CRITICAL POWER RATIO.

l condition of 2,536 MWt. The analyses were based upon plant The type of transients evaluated were increase in pressure and operation in accordance with the opersing map given in the power, positive reactmty inserbon, and coolant temperature I current load line limit analysis. In addition,2,536 MWt is the decrease. TM limiting transient yteids the largest delta MCPR.

licensed maximum power level of Fitzr.e ick, and this represents When added to the Safety Lmt, the requred operating limit the maximum steady-state power which shall not knowingly be MCPR in the Core Operating Limits Report is obtamed.

c.a, re -

The evaluation of a given transeent begms with the system initial 1 The transient analyses performed for each raioed are described parameters shown in the current reload analysis and Reference 2 in Reference 2. Modois and model conservatism are also that are input to the core dynamic behavior transient computer descnbod in this reference. As reanmaari in Reference 4, the programs desenbod in Reference 2. The output of these core wide transient analysis for one recirn dadan pump operation programs along with the initial MCPR form the input for the is conservatively bounded by two-loop operation analysis, and further erdi n of the INmidi limited bundle with a single the flow-dependent rod block and scram setpoint arymeinns are channel transient thermal hydraulic code. The pnncipal result of acqusted for one-pump operation. Reference 1 evaluates the the evaluehon is the reduction in MCPR can mari by the transient.

safety significance of uprated power operation at 2,536 MWt.

This evaluation is consistent with and demonstrates the acceptabikty of the transient erdi n required by Reference 2.

Fuel cladding integrity is assured by the applicable operating limit MCPR for steady state conditions given in the Core Operating Umits Report (COLR). These operating limit MCPR's are derived from the estabhshed fuel cladding integnty Safety Umit, and an analysis d abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating Amendment No. [, p, [, f,1/2,

m . . _ . _. . .

I JAFNPP '

2.1 BASES (Cont'd)

The MCPR operating limits in the COLR are conservatively A. Trip Settings ,

assumed to exist prior to initiation of the transsents. l The bases for individual trip settogs are rimammari in the i 4

This choice of using ooneervative values of controhng foNowmg paragraphs.

e parameters and initiating transients W the design power level, produces more possimistic answers than would result by usmg 1. Neutron Flux Trip Settmas  ;

expected values of control parameters and analyzing at higher t powerlevels. a. IRM Flux Scram Trip Setting Steady-state operation without forood recirculation is not The IRM system consists of 8 chambers,4 in each of permitted. The analysis to support operation at various power the reactor protection system logic channels. The and flow E'"x=i4. has considered operation with either one IRM is a SAearia instrument which covers the  !

or two recirculation pumps. range of power level between that covered by the SRM and the APRM. The 5 derariam are covered by in summary: the IRM by rneens of a range switch and the 5  ;

darwian are broken down into 10 ranges, each

. The abnormal operational transients were analyzed to the being onehalf of a dearia in size. The IRM scram licensed maximum powerlevel. trip setting of 120 divisions is active in each range of ,

the IRM. For example, if the instrument were on i l . The hcensed maximum power level is 2,536 MWt. Range 1, the scram settmg would be a 120 dmssons for that range; hkowise, if the instrument were on

. Analyses of transients employ artarpealy conservative range 5, the scram would be 120 dmsions on that values of the controllmg reactor parameters. range. Thus, as the IRM is ranged up to ,

accommodate the increase in power level, the scram

. The erd,1; cal procedures now used result in a more trip setting is also ranged up. The most rJyaic6nt logical answer than the attemative method of assuming a sources of reactuty change dunng the power higher startmg power in conjunction with the expected increase are due to control rod withdrawal. For values for the parameters. insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM i scram would result in a reactor shutdown well before any Safety Limit is exceeded Amendment No. If, If, ,

, If,

I JAFNPP 2.1 BASES (cont'd)

5. Main Steam Une isolation valve Closure Scram Trip setting 1 3. Turtune Stop Valve Closure Scram Trip Settmas The low pressure isolation of the main steam lines at 825 psig The turtune stop . valve closure scram trip anticipates the was provided to give protection agamst rapid . reactor pressure, neutron flux and heat flux increase that could result depressunzation and the resulting rapid cooldown of the vessel.

from rapid closure of the turbine stop valves. With a scram trip Advantage was taken of the scram feature which occurs when setting of $10 percent of valve closure from full open, the the main steam line isolation valves are closed, to provide for '

resultant increase in surface heat flux is limited such that MCPR reactor shutdown so that high power operation at low reactor remams along the Safety Limit even during the worst case pressure does not occur, thus providing protection for the fuel transient that assumes the turtune bypass is closed. This scram cl.4.g integnty safety limit. Operation of the reactor at is bypassed when reactor power is below 29% of rated, as pressures lower than 825 psig requires that the Reactor Mode man =med by turtune first stage pressure, consistent with the Switch be in the Startup position where protection of the fuel I safety analysis discussed in Reference 1. cladding integnty safety limit is provided by the APRM high neutron flux sesam and the IRM. Thus, the combination of main

4. Turbine Control Valve Fast Closure Scram Trip Setting stream ime lou pressure isolation and isolation valve closure scram assure; the availability of neutron flux scram protection This turbine control valve fast closure scram anhenpates the over the entine range of applicability of the fuel cladding integrity pressure, neutron flux, and heat flux increase that could result safety limit. In addition, the isolation valve closure scram from fast closure of the turbine control valves due to load anticipates the pressure arid f'ax transeents which occur during rejection exceedmg the capability of the turtune bypass. The Reactor Protection System initiates a scram when fast closure of normal or inadvertent isolation valve closure. With the scrams set at $ 10 percent valve closure, there is no increase in neutron the control valves is initiated by the fast actmg solenoid valves.

This is achieved by the action of the fast actmg solenoid valves flux.

in rapidly reducing hydraulic control oil pressure at the main

6. Main Steam Une isolation Valve Closure on Low Pressure turtune e.ontrol valve actuator disc dump valves This loss of pressure is sensed by pressure switches whose contacts form The low pressure isolation minimum limit at 825 psig was the one-out-of-two-twice logic mput to the reactor protectum provided to give protection against fast reactor depressurization system. This trip setting, a nommally 50 percent greater closure and the resulting rapid cooldown of the vessel. Advantage was time and a different valve characteristic from that of the turbme taken of the scram feature which occurs when the main steam stop valve, combine to produce tiansients very similar and no line isolation valves are closed to provide for reactor shutdown more severe than for the stop valve. No ssgnificant change in so that operation at pressures lower than itose specified in the MCPR occurs. Relevant transient analyses are dimcad in thermal hydraulic safety limit does not occur, although operation j

Section 14.5 of the Final Safety Analysis Report and Reference 1. at a pressure lower than 825 psig would not necessarily I This scram is bypassed when reactor power is below 29 percent of rated, as measured by turbine first stage pressure.

constitute an unsafe condition.

Amendment No. ,[,[, [ f ,d 19

, JAFNPP 2.1 BASES (Cont'd)

C. References

1. General Deyu-ic Report, NEDC-32016P, ' Power Uprate -

Safety Analysis for the James A. FitzPatrick Nuclear Power Plant" December 1991 (piupiny).

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)

4. FitzPatrick Nuclear Power Plant Single-Loop Operation, NEDO-24281, August,1980.

s g g g g (Next page is 23)

I

, JAFNPP '

1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR COOtANT SYSTEM APPUCABluTY: APPUCABluTY:

Apphes to Imts on reactor coolant system pressure. '

Applies to trip settings of the instruments and devices whch are  ;

provided to prevent the reactor coolant system safety hmits from bemg j exceeded. -

OBJECTIVE: OBJECTIVE:  !

To establish a limit below which the integrity ci the Reactor Coolant To define the level of the process variables at wNch automabc protectwe I System is not threatened due to an overpressure condition. achon is initsated to prevent the safety Imts from being exceeded.

SPECIFICATION: SPECIFICATION- I

1. The reactor coolant system pressure shall not exceed 1. The Unwtog Safety System settmg shall be specified 1,325 psig at any time when irradiated fuel is present in the below:

reactor vessel. t A. Reactor coolant high pressure scram shall be "

11,000 peig. li B. Reactor coolant system safety / relief valve nominal  !

settings shall be $1,145 psig. The allowable l ;

setpoint error for each safety / relief valve shall be 1 percent.

Amendment No. , , , , ,

-- JAFNPP 1.2 and 2.2 BASES t

The reactor coolant pressure boundary integrity is an important The current reload analysis shows that the main steam isolation bamer in the prevention of uncontrolled release of fissson valve clnsure transeent, with flux scram, is the most severe event i

i products. It is essenhal that the integnty of tHs boundasy be resulhng directly in a reactor coolant system pressure increase. i protected by MW a pressure limit to be observed for all The reactor vessel pressure code limit of 1,375 psig, given in operating condibons and whenever there is irradiated fuel in the FSAR Section 4.2, is above the peak pressure produced by the reactor vessel. event above. Thus, the pressure safety limit (1,375 psig) is well  !

above the peak pressure that can result from reasonably ,

The pressure safety limit of 1,325 peig as measured by the vessel '

expected overpressure transsents. (See current reload analysis steam space pressure indicator is equivalent to 1,375 psig at the for the curve produced by this analysis.) Reactor pressure is lowest elevation of the Reactor Coolant System. The 1,375 psig continuously indicated in the control room during operation.

value is denved from the design pressures of the reactor ,

pressure vessel and reactor coolant system pepeng The A safety limit is applied to the Residual Heat Removal System  !

respective design pressures are 1250 peig at 575*F for the (RHRS) when it is operating in the shutdown cooling mode reactor vessel,1148 psig at 56ErF for the recirculation suction  !

When operating in the shutdown cooling mode, the RHRS is  !

piping and 1274 psig at 575* for the discharge piping. The included in the reactor coolant system. .j pressure safety limit was chosen as the lower of the pressure i

transients permitted by the applicable design codes: 1965 The numerical safety / relief valve setpoint shown in 2.2.1.B is  :

ASME Boiler and Pressure Vessel Code, Section lit for pressure justified by analyses desenbed in the General Electric report ,

vessel and 1989 ANSI B31.1 Code for the reactor coolant system NEDC-32016P and assures that the structural acceptance  ;

peping. The ASME Boiler and Pressure Vessel Code permsts criteria set forth in the Mark 1 Containment Short Term Program j pressure transeents up to 10 percent over desegn pressure (110% are sabsfied. j x 1,250 - 1,375 psig) and the ANSI Code permits pressure  :

transients up to 20 percent over the design pressure (120% x  !

1,150 - 1,380 psig). The safety limit pressure of 1,375 psig is i referenced to the lowest elevation of the Reactor Coolant System.

j t

t t

Amendment No. ,

, %, i

i r

JAFNPP I

3.1 BASES (cont'd) is discharged from the reactor by a scram can be The IRM high flux and APRM <15% power scrams provide ~

ecc0iTiin0 dated in the dscharge pepmg. Each scram discharge Wmte coverage in the startup and intermediate range. Thus, f instrument volume. accommodates in excess of 34 gallons of the IRM and APRM systems are required to be operable in the I water and is the low point in the piping. No credit was taken for refuel and startup/ hot standby modes. The APRM <120% l this volume in the design of the discharge pepeng as concems the power and flow referenced scrams provide required protection in i amount of water which must be accommodated during a scram. the power range (reference FSAR Sechon 7.5.7). The power  !

range is covered only by the APRMs. Thus, the IRM system is  ;

Dunng normal operation the dscharge volume is empty; not required in the run mode. i however, should it fill with water, the water docharged to ths i piping from the reactor could not be accommodated, which The high reactor pressure, high drywell pressure, reactor low t

would result in slow scram times or perhal control rod insertion. water level and scram discharge volume high level scrams are i To preclude this occummce, level detection instruments have required for startup and run modes of plant operation. They are, been provided in each instrument volume which alarm and therefore, required to be operahonal for these modes of reactor scram the reactor when the volume of water reaches 34.5 operation. j gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the The requirement to have the scram funchons indicated in Table i scram times or amount of insertson of the control rods. This 3.1-1 operable in the refuel mode assures that shifting to the function shuts the reactor down while sufficsont volume remains refuel mode during reactor power operation does not diminish to accommodate the discharged water and precludes the the protection provided by the Reactor Protechon System.  ;

situation in which a scram would be required but not be able to

• perform its function +M_-M _"i. Turbine stop valve closure occurs at 10 percent of valve closure.

Below 29% of rated reactor power, the scram signal due to l A Source Range Monitor (SRM) System is also provided to turbine stop valve closure is bypassed We the flux and i supply additional neutron level information during startup but has pressure scrams are adequate to protect the reactor.

no scram functions (reference paragraph 7.5.4 FSAR).

)

t i

\

Amendment No.d J , f, 34 '

JAFNPP 3.1 BASES (cont'd) 4 Turbine control valves fast closure initiates a scram based on i pressure switches sensmg electro-hydraulic control (EHC) '

system oil pressure.. The switches are located between fast i closure solenoids and the disc dump valves, and are set relative  !

(500<P<850 psig) to the normal (EHC) oil pressure of 1,600 psig so that based on the small system volume, they can rapidly detect valve closure or loss of hydraulic pressure.  ;

The requirement that the IRM's be inserted in the core when the i

APRM's read 2.5 indicated on the scale in the start-up and refuel i 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. i i

B. The limiting transient which determines the required steady state MCPR limit depends on cycle exposure. The operating limit  :

MCPR values as determmed from the transient analysis in the current reload submittal for various core exposures are specified i in the Core Operatmg Umsts Report (COLR). '

The ECCS performance analyses assumed reactor operation will be limited to MCPR = 1.20, as desenbod in NEDO-21662 and  !

l NEDC-31317P includog latest rewmon, errata and (Wa The Technical Specifications limit operation of the raretor to the  ;

more conservative MCPR based on consideratice of the limiting -  !

transsent as specified in the COLR.  !

I Amendment No. , ,If,If,

_%..u-w __. _ =._ --m_ m _m -.-__u

JAFNPP f TABLE 3.1-1 (cont'd)

REACTOR PROTECDON SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Mirumum No. Modesin Which Funchon of Operable Must be Operable Total Number of l

. Instrument Instrument Channels  ;

Channels Per Refuel Startup Run Pronded by Design Action l Trip System (1) Trip Function Trip Level Sethng 1 for Both Trip Systems

~

(6) (1) 4 2 APRM Downscale > 2.5 indicated on X 6 Instrument Channels A or B i

-scale (9) 1 2 High Reactor Pressure 51,000 psig X(8) X X 4 Instrument Channels A l s 2 High Drywell Pressure 52.7 peig X(7) Xft) X 4 IrAJment Channels A 2 Reactor Low Water -> 177 in. above TAF X X X 4 Instrument Channels A i Level 3 High Water Levelin $34.5 gallons per X(2) X X 8 Instrument Channels A Scram Discharge Volume Instrument Volume l 2 X X X A Main Steam Line < 3x normal full 4 Instrument Channels High Radianon -power background (16) 4 Main Steam Line 510% valve closure X(5) 8 Instrument Channels A Isolation Valve closure 4 Turbine Stop 510% valve X(4)(5) 8 Instrument Channels A or C  ;

valve Closure closure i

I i

-no. p. 9.p.p.ir.ys. ,p. p. ,,,

~,

JAFNPP i TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT 1

NOTES OF TABLE 3.1-1

1. There shall be two operable or tripped trip systems for each function, except as specifMxi in 4.1.D. From and after the time that the minimum number of operable instrument channel for a trip system cannot be met, that affected trip system shMI be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.

A. Initiate insertion of operable rods and complete 'msertion of all operable rods within four hours.

B. Reduce power level to IRM range and place Mode Switch in the Startup Position within eight hours.

C. Reduce power to less than 29 percent of rated.

l

2. Permissible to bypass, if Refuel and Sht.4down positions of the Reactor Mode Switch.

3. Deleted.

4. Bypassed when the reactor power is less than 29 percent of rated.  !

5. The design permits closure of any two lines without a scram being initiated.

6. When the reactor is subcritical and the reactor. water temperature is less than 212 F, only the following trip functions need to be operable:

A. Mode Switch in Shutdown B. Manual Scram.

Amendment No. [,[ p!d, %,

42

ATTACHMENT lli to JPN 92-028 NEDC-32016P " POWER UPRATE SAFETY ANALYSIS FOR THE JAMES A. FITZPATRICK NUCLEAR POWER PLANT" (JPTS 91-025) l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50333 DPR-59

JAFNPP L 1

3.5 (Coned) 4.5 (Cont d)

DELETED C. HIGH PRESSURE COOLANT INJECTION (HPCI SYSTEM) C. HIGH PRESSURE COOLANT INJECTION (HPCI SYSTEM)

Survedlance of HPCI System shall be pah6ved as follows i

provided a reactor steam supply is available. If steam is not avadable at the time the survedlance test is scheduled to be i performed, the test shall be performed within 10 days of continuous operation from the time steam becomes available.

i

1. The HPCI System shall be operable whenever the reactor 1. HPCI System testing shall be as specified in 4.5.A.1.a. b, c, '

pressure is greater than 150 psig and reactor coolant d, f, and g except that the HPCI pump shall deliver at least temperature is greater than 2127 and irradiated fuel is in 4,250'gpm against a system head corresponding to a the reactor vessel, except as W below- reactor vessel pressure of 1,195 psig to 150 psig.

l i

I Amendment No. [,[, J d7, 117 i

, JAFNPP 35 (cont'd) 4.5 (cont'd)

The RCIC pump shall dehver at least 400 gpm for a system head correspondmg to a reactor pressure of 1,195 psig to 150 psig. l,

2. When it is detemnmed that the RCIC System is inoperable at a time when it is required to be operable, the HPCI System shall be venfied to be operable immediately and daily thereafter.

l Amendment No. pd, }/8, 121a

P l

t JAFNPP ,

Figure 3.5-1  ;

Thermal Power and Core Flow Limits of l'

. Specifications 3.5.J.1,3.5.J.2, and 3.5.J.3 i

70

. Stabety Monitoring Stabaty Stabety Monitoring i (APRM and LPRM) Required Monito ing (APRM and LPRM) Required .

(APRM and L During Two{oop Opwah For Single Loop Operation .

8 W

SO - LPRM)

Required i

< During r E Single and ,

y 50 - Single-Loop Operation Two-Loop Operation

• m g Pr m ed i Line A W . .

6: 40 - i ja E i i ~

W 35 -  :  :

3:  :  : i

k. 3 0 -  !

I i <  :  :

2 g

!. i l

W 20 - Stability M6nitoring No(Required I

F i i W  :  :

E  :  :

O 10 - i  ;

: i O ii iii iiiilg j i i i 30 40 45 50 60 70 CORE FLOW (PERCENT RATED)

Amendment No. [ [, ,h .

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

7. Reactor Vessel Flux Monitoring The reactor vessel Flux Monitoring Surveillance Program complies with the intent of the May,1983 revision to 10 CFR 50, Appendces G and H. The next flux monitoring  ;

surveillance capsule shall be removed after 15 effective full i power years (EFPYs) and the test procedures and ,

reporting requirements shall meet the requirements of ASTM E 185-82.

I B. Deleted B. Deleted C. Coolant Chemistry C. Coolant Chemistry

1. The reactor coolant system radioactivity concentration in 1. a. A sample of reactor coolant shall be taken at least l water shall not exceed the equilibrium value of 0.2 pCi/gm every 96 hr and analyzed for gross gamma activity. i of dose equivalent 1-131. This lirnit may be exceeded

• following a power transient, for a maximum of 48 hr. b. M@ Msis of a @ d re WM Mi During this iodine activity transient the iodine be made at least once/ month.

concentrations shall not exceed the equilibrium limits by c. A sample of reactor coolant shall be taken prior to more than a factor of 10 whenever the main steamline startup and at 4 hr intervals during startup and isolation valves are open. The reactor shall not be analyzed for gross gamma activity. '

opwated more than 5 percent of its W poww d.

operation under this exception to the equilibrium limits. If During plant steady state operation and fo!!owing an the sodine concentration exceeds the equilibrium limit by offgas activity increase (at the Steam Jet Air more than a factor of 10, the reactor shall be placed in a Ejectors) of 10,000 pCi/sec within a 48 hr. period or  !

cold condition within 24 hr. a power level change of >20 percent of full rated '

power /hr reactor coolant samples shall be taken  !

and analyzed for gross gamma activity. At least  !

three samples will be taken at 4 hr intervals. These ,

sampling requirements may be omitted whenever I the equilibrium I-131 concentration in the reactor  !

coolant is less than 0.007 Ci/ml.  ;

Amendment No. [

139  !

JAFNPP 3.6 and 4.6 BASES (cont'd)

The expected neutron fluence at the reactor vessel wall can be addenda). The RTNOT values for the reactor vessel flange determined at any point during plant life based on the linear region and for the reactor vessel shell beltline region are 30 F, relationship between the reactor thermal power output and the based on fabrication test reports. The RT NOT for the remainder corresponding number of neutrons produced. Accordingly, of thevesselis 40*F.

neutron flux wires were removed from the reactor vessel with the surveillance specimens to establish the correlation at the capsule The first surveillance capsule containing test specimens was ,

location by expenmental methods. The flux distribution at the withdrawn in April, i985 after 6 EFPY. The test specimens vessel wall and 1/4 ti-sc.krmss (1/4T) depth was analytically removed were tested according to ASTM E 185-82 and the determined as a function of core heeght and azimuth to establish results are in GE report MDE49G86 The next surveillance the peak flux location in the vessel and the lead factor of the capsule will bc removed after 15 EFPYs of operation and the surveillance specimens. results of the examination used as a basis for revision of Figure 3.6-1 curves A, B and C for operation of the plant after 16 EFPYs.

Regulatory Guide 1.99, Revision 2 is used to predict the shift in RTNDT as a function of fluence in the reactor vessel beltline Flyze 3.6-1 is comprised of three parts: Part 1, Part 2, and Part region. An evaluation of the irradiated surveillance specimens, 3. Parts 1,2, and 3 establish the pressure-temperature limits for whech were withdrawn from the reactor in April,1985 (6 EFPY), plant operations through 12,14, and 16 Effective Full Power shows a shift in RTNDT less than that predicted by Regulatory Years (EFPY) respectively. The appropriate figure and the Guide 1.99, Revision 2. pressure-temperature curves are dependent on the number of accumulated EFPY. Figure 3.6-1, Part 1 is for operation through Operating limits for the reactor vessel pressure and temperature 12 EFPY, Figure 3.6-1, Part 2 is for operation at greater than 12 during normal heatup and cooldown, and during in-service EFPY through 14 EFPY, and Figure 3.6-1, Part 3 is for operation hydrostatic and leak testing were established using 10 CFR 50 at greater than 14 EFPY through 16 EFPY. The curves contained Appendix G, May,1983 and Appendix G of the Summer 1984 in Figure 3.6-1 are developed from the General Electric Report Addenda to Section til of the ASME Boiler and Pressure Vessel DRF 137-0010, " Implementation of Regulatory Guide 1.99, Code. These operating limits assure that the vessel could safely Revision 2 for the James A. FitzPatrick Nuclear Power Plant,"

accommodate a postulated surface flaw having a depth of 0.24 dated June,1989.

inch at the flange-to-vessel junction, and one-quarter of the material thickness at all other reactor vessel locations and Figure 3.6-1 curve A establishes the minimum temperature for discontinuity regions. For the purpose of setting these operating hydrostatic and leak testing required by the ASME Boiler and i limits, the reference temperature, RTNDT , of the vessel material Pressure Vessel Code,Section XI. Test pressures for in-service was estimated from impact test data taken in accordance with hydrostatic and leak testing are a function of the testing ,

the requirements of the Code to which the vessel was designed temperature and the component material. Accordingly, the and manufactured (1965 Edition including Winter 1966 maximum hydrostatic test pressure will be 1.1 times the operating pressure or about 1,144 psig. l Amendment No.1[,1[,

147

r f

i JAFNPP l

t 3.6 and 4.6 BASES (cont'd)

B' Deleted annunciating at appropriate concentration levels such that l sampling for isotopic analysis can be initiated. The design  !

details of such a system must be submitted for evaluation and .

C. Coolant Chemistry

• accepted by the Commission prior to its implementation and incorporation in these Technical Specifications. f A radioactivity concentration limit of 20 pCi/ml total iodine can j be reached if the gaseous effluents are near the limit as set Since the concentration of radioactivity in the reactor coolant is i forth in Radiological Effluent Technical Specification Section not continuously measured, coolant sampling would be 3.2.a if there is a failure or a prolonged shutdown of the ineffective as a means to rapidly detect gross fuel element ,

cleanup domineralizer. failures. However, some capability to detect gross fuel element failures is inherent in the radiation monitors in the offgas j

• In the event of a steam line rupture outside the drywell, a more system and on the main steam lines. l restrictive coolard activity level of 0.2pCi/gm of dose equivalent 1-131 was assumed. With this coolant activity level and Materials in the Reactor Coolant System are primarily 304 adverse meteorological conditions, the calculated radiological stainless steel and Zircaloy fuel cladding. The reactor water l dose at the site boundary would be less than 30 rem to the chemistry limits are established to prevent damage to these '

thyroid. The reactor water sample will be used to assure tr,d materials. Umits are placed on chloride concentration and the limit of Specification 3.6.C is not exceeded The total conductivity. The most important limit is that placed on  !

radioactive iodine activity would not be expected to change chloride concentration to prevent stress corrosion cracking of l rapidly over a period of 96 hr. In addition, the trend of the stack the stainless steel. The attached graph, Fig. 4.6-1, illustrates l offgas release rate, which is continuously monitored, is a good the results of tests on stressed 304 stainless steel specimens.

indicator of the trend of the iodine activity in the reactor Failures occurred at concentrations above the curve; no coolant. Also during reactor startups and large power changas failures occurred at concentrations below the curve. According  ;

which could affect iodine levels, samples of reactor coolant to the data, allowable chloride concentrations could be set  :

shall be analyzed to insure iodine concentrations are below several orders of magnitude above the established limit, at the allowable levels. Analysis is required whenever the I-131 oxygen concentration (02-0.3 ppm) experienced during power concentration is within a factor of 100 of its ' allowable operation. Zircaloy does not exhibit similar stress corrosion i equilibrium value. The necessity for continued sampling failures.  !

following power and offgas transients will be reviewed within 2  !

years of initial plant startup. However, there are various conditions under which the ,

dissolved oxygen content of the reactor coolant water could bc  :

The surveillance requirements 4.6.C.1 may be satisfied by a higher than 0.2-0.3 ppm, such as refueling, reactor startup, and j continuous monitoring system capable of determining the total hot standby. During these periods with steaming rates less iodine concentration in the coolant on a real time basis, and j Amendment No. [ 149 I

JAFNPP 4.7 (cont'd)

(4.) See table 4.7-2 for exceptions.

(5.) Acceptance criterion - The combined les'xage rate for all penetrations and valves subject to type B and C tests shall be !ess than 0.60 La. Leakage from containment isolation valves that are sealed with fluid from a seal system may be excluded when determining the ceruned leakage rate provided that the installed isolation valve '

seal-water system fluid inventory is sufficient to assure the sealing function for at least 30 days.

d. Otherleak rate tests (1) The leakage rate for containment isolation valves 10-AOV-68A, B (penetration X-13A, B) for Low Pressure Coolant injection system and 14-AOV-13A, 8 (penetration X-16A, B)  :

for Core Spray System shall be less than 11 cubic feet per minute per valve -

(pneumatically tested at 45 psig with amtwent temperature) or 10 gallons per minute per valve (hydrostabcally) tested at 1,035 psig l i with ambient temperature.

Amendment No. f,1 ,

._a__. u -4 6 - +4>, ~ 4 - a

.e . ,

.a.~ 4 1

, JAFNPP 4 3.7 BASES A. Primary Contamment The pressure suppresson pool water prcmdes the heat sink for the Reactor Coolant System' energy . release felW a The integrity of the primary containment and operation of the postulated rupture of the system. The pressure suppression i Emergency Core Cooling Systems in combination limit the chamber water volume must absorb the associated decay and ,

offsite doses to values less than those specified in 10 CFR 100 in structural sensible heat released dunng reactor coolant system l the event of a break in the Reactor Coolant System pipeg. blowdown from 1,040 peig. ll Thus, containment integrity is required whenever the potential for i viniatino of the Reactor Coolant System integrity exists. Smce all of the gases in the drywell are purged into the pressure  !

Concem about such a violation exists whenever the reactor is suppression chamber ' air space during a loss of coolant t critical and above atmospheric pressure. An excephon to the accident, the pressure resultmg from isothermal compression ,

requirement to maintain primary containment integrity is allowed plus the vapor pressure of the liquid must not exceed 56 psig,  ;

during core loading and during low power physics testing when the suppression chamber design pressure. The design volume ready anname to the raarent vessel is required. There will be no of the suppression chamber (water and air) was obtamed by  ;

pressure on the system at this time, which will greatly reduce the considering that the total volume of reactor coolant to be '

chances of a pipe break. The reactor may be taken. critical condensed is discharged to the suppressKm chamber and that .

during this period, however, restrictive operating prreart res and the drywell volume is purged to the suppression chamber  !

operation of the RWM in accordance with Specification 3.3.B.3 (Section 5.2).  ;

minimize the probability of an accident occurring. Procedures in conjunction with the Rod Wbrth Mnimize Technical Specifications limit individual control worth such that the drop of any in-sequence control rod would not result in a peak fuel enthalpy greater than 290 calories /gm. In the unilhely event that an excursion did occur, the reactor bulk 5ng and Staney Gas Treatment System, wtuch shall be operational during this time, offers a sufficient barrier to keep offalle rinaam well within 10 CFR 100.

~

i Amendment No.1)l,1[,

187

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

JAFNPP 3.7 BASES (cont'd)

Using the mmimum or maximum downcomer submergence pumps), contamment pressure is required to rnaintain adequate l

levels gwen in the specsllcahon, cordainment pressure dunng the not positive suction head (NPSH) for the core spray and LPCI design basis acendent is approximately 45 psig which is below pumps.

the design of 56 psig. The minimum downcomer submergence g suppresse W tempwatwe to 105*F W RCIC, l[

of 51.5 in. results in an approximate suppression chamber water ,w pw W stwed H

I volume of 105,900 ft.3 The maiority of the Bodega tests (9) were ,

enwgy are rh frwn the pmary systwn by hgog run with a submerged length of 4 ft. and with complete reactor steam directly to the suppression chamber assures Thus, e W to h % e mar n ap al h any time dwing RCIC, this specificahon is ariarpda Additional JAFNPP specific er.0,eee done in connachon with the Mark l Contamment-

• UE" '

Suppression Chamber Integrity Program indicate the adequacy Expenments indicate that unacceptably high dynamic of the sp cb range of submergence to ensure that dynamic contamment loads may result from unstable cornn6ahon when 3

forces associated with pool swell do not result in overstress of suppression pool water temperatures are high near SRV the suppression chamber or associated structures. Level discharges. Action statements limit the maximum pool ,

instrumentahon is provided for operator use to maintain temperature to assure stable condensation. These actions downcomer submergence withm the specified range. include: limiting the maximum pool temperature of 95"F during OPwM % a reacts saam N Mng a transient The maximum temperature at the end of blowdown tested during as a stuck open W W tanpwatwo exM M -

the Humboldt Bay (10) and Bodega Bay tests was 1707, and and depressunzmg the reactor if pool temperature exceeds  :

this is conservatwely Men to be the M fa complete 120*F. T-quenchers diffuse steam discharged from SRVs and l d the reacts W, M N occw Mempwatwes h 1M romote stable condensation. The presence of T-quenchers and t i

curylierce with these achon statements assure that stable Containmort analyses predict a 46*F increase in pool water condensabon will occur and containment loads will be temperature, after complete LOCA blowdown. These analyses ar rapaahia.

an W W pool waiw tempwatwo of 95*F NEDC-24361P (August 1981) summanzes analyses performed to empwa was W N M W pant 6 of the assume an initial transients using these temperature hmits at a power level of 2535  :

tempwatwo. Thwh, @e h b -M MWt (104% of rated). NEDC-24361P also substanhates the durmg a LOCA because the maximum pool temperature (1417) is h than the 1707 tempwatwo seen dwing h W Bay d the M W W h W W M d i NUREG-0661. NEDO-30832 (December 1984) shows that SRV tests. condensabon inarin are low csnpiwed to other des!gn loads for For an initial maximum suppression chamber water temperature plants with T-quenchers. NEDO-30832 describes why local pool of 95'F, assuming the worst case complement of containment temperatures need not be analyzed at a rated power level of l cooling pumps (one LPCI pump and two RHR service water 2536 MWt.

~mdment No. If, f, If, fl 188 i

t

, JAFNPP

'f 4.7 BASES A. Pnmary Contamment Design basis accidents were evaluated as discussed in Section 14.6 of the FSAR and the power uprate safety evaluation, The water in the suppression chamber is used only for coolmg in Reference 18. The whole body and thyroid doses in the control- i the event of an accident; i.e., it is not used for normal operation; room, low populahon zone (LPZ) and site boundary meet the therefore, a daily check of the temperature and volume is requirements of 10 CFR Parts 50 and 100. The techrucal support adequate to assure that adequate heat removal capabikty is center (TSC), not designed to these hcensing bases, was also .

present analyzed. The whole body and thyroid dose acceptance criteria used for the main control room are met for the TSC when initial  !

The pnmary containment preoperational test pressures are wma to the TSC and occupancy of certain areas in the TSC is based upon . the calculated primary containment pressure restricted by admmistrative control. The LOCA dose evaluation, response corresponding to the design basis loss-of-coolant Reference 19, a=muned: the primary contamment leak rate was r accident. The peak drywell pressure would be about 45 peig 1.5 volume percent per day; source term releases were in which would rapidy reduce to 27 psig within 30 sec. following accordance with TID-14844; and the standby gas treatment the pipe break. Followng the pipe break, the suppression system filter efficiency was 99% for halogens. These doses are  ;

chamber pressure rises to 26 psig within 30 sec, equalizes with also based on the drywell pressure and thereafter rapidy decays with the drywell  ;

pressure decay (14).

The design pressure of the drywell and suppression chamber is 56 psig(15). The design basis accident leakage rata is 0.5 ,

percent / day at a pressure of 45 psig. As pointed out above, the drywell and suppression chamber pressure following an accident  !

would equalize fairly rapidly. Based on the primary containment j

pressure response and the fact that the drywell and suppression  !

l chamber function as a unit, the primary containment will be i tested as a unit rather than the individual components separately.

l

! Amendment No. '

193 5 t

l l

JAFNPP (A) ROUTINE REPORTS (Continued)

4. CORE OPERATING UMITS REPORT i

a. Core operating limits shall be established prior to startup from eae cycle, or prior to any remaining portion of a reload cycle for the followine

. The Average Planar unear Heat Generation Rates (APLHGR) of Specification 3.5.H; The Minimum Critical Power Ratio (MCPR) and MCPR low flow adjustment factor, K, , of Specifications 3.1.B and 4.1.E;

. The Unear Heat Generation Rate (LHGR) of Specification 3.5.l;

. The Reactor Protection System (RPS) APRM flow biased trip settings of Table 3.1 1; and The flow biased APRM and Rod Block Monitor (RBM) rod block settings of Table 3.2-3.

and shall be documented in the Core Operating Umits Report (COLR).

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC as described in:

1. " General Electric Standard Application for Reactor Fuel," NEDE-24011 P, latest approved version and amendments.

2. " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR LOCA Loss-of Coolant Accident Analysis,' NEDC-31317P, October,1986 including latest revision, errata and addenda. l

3. " Loss ofCoolant Acadent Analysis for James A. FitzPatrick Nuclear Power Plant," NEDO 21882 2, July,1977 including latest errata and addenda.

c. The core operating limits shall be determined such that all applicable limits

(e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident

. analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements thereto, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

Amendment No.1[,

2s4c v v s ~ r e -

P i

JAFNPP

[

[

7.0 REFERENCES

t 1 (1) E. Janssen, " Multi-Rod Burnout at Low Pressure," ASME (11) Section 5.2 of the FSAR.

Paper 62-HT-26, August 1962.

, (12) TID 20583, " Leakage Characteristics of Steel ,

(2) K.M. Backer, "Bumout Conditions for Flow of Boiling Containment Vessel and the Analysis of Leakage Rate Water in Vertical Rod Clusters," AE-74 (Stockholm, Determinations.* .

Sweden), May 1962.

(13)- Techrwcal Safety Guxle, " Reactor Containment Leakage (3) FSAR Sect'on 11.2.2. Testing and Surveillance Requirements," USAEC, Division of Safety Standards, Revised Draft, (4) FSAR Section 4.4.3. December 15,1966.  ;

(5) 1.M. Jacobs, " Reliability of Engineered Safety Features as (14) Section 14.6 of the FSAR. .

a Function of Testing Frequency," Nuclear Safety, Vol. 9, No. 4, July-August 1968, pp 310-312. (15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, Section ill Maximum allowable internal pressure is 62 (6) Benjamin Epstein, Albert Shiff, UCRL-50451, improving psig.

Availability and Readiness of Field Equipment Through  ;

Periodic inspection, July 16,1968, p.10, Equation (24), (16) 10CFR50.54, Appendix J, Reactor Containment Testing Lawrence RarGahon Laboratory. Requirements."

(7) 1.M. Jacobs and P.W. Mariott, APED GuKlelines for (17) 10CFR50, Appendix J February 13,1973.

Determining Safe Test Intervals and Repair Times for Engineered Safeguards- April 1969. (18) General Electric Report NEDC-32016P, " Power Uprate Safety- Analysis for the James A. FitzPatrick Nuclear (8) Bodega Bay Preliminary Hazards Report, Appendix 1, Power Plant," December 1991 (proprietary).

Docket 50-205, December 28,1962.

(19) James A. FitzPatrick Calculation JAF-CALC-RAD-00008, (9) C.H. Robtzns, " Tests of a Full Scale 1/48 Segment of the "Rcdiological Consequences of Design Basis Accidents Humbolt Bay Pressure Suppression Containment," at James A. FitzPatrick," November 1991.

GEAP-3596, November 17,1960.

(20) General Electric Report GE-NE-187-45-1191 P, (10) " Nuclear Safety Program Annual Progress Report for " Containment Systems Evaluation," (proprietary).

l Period Ending December 31,1966, ORNL-4071."

Amendment No.

285

~

L-l l

l l ATTACHMENT 11 to JPN 92-028 l

SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES l POWER UFRATE (JPTS-91-025)

1. PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed changes is to revise the Technical Specifications to permit operation of the James A. FitzPatrick Nuclear Power Plant at an uprated power of 2536 MWt.

Engineering analyses and evaluations confirm that the plant can be operated at an uprated power. The increase in the rated power from 2436 MWt to 2536 MWt corresponds to a 4.8 percent increase in rated steam flow (Reference 3, Section 1.2). The increase in rated power l remains below the plant design power level of 2,550 MWt which was the basis for the original plant safety evaluation, Reference 14.

The Technical Specification changes necessary for power uprate are identified and evaluated in i

this safety evaluation. The changes to the Technical Specifications were identified from the results and conclusions of References I to 6. These include two generic licensing topical reports prepared by General Electric: NEDC 31897P-A

• Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Reference 1, referred to as LTR 1, and; NEDC-31984P ' Generic l

Evaluations of General Electric Boiling Water Reactor Power Uprate," and Supplement 1, Reference 2, referred to as LTR 2. They also include plant specific analyses: General Electric Report NEDC-32016P " Power Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant," Reference 3, referred to as the PUSAR; Stone & Webster Engineering Corporation

• Core l

Power Uprate Engineering Report for James A. FitzPatrick Nuclear Power Plant,' Reference 4, referred to as the Engineering Report; General Electric Report NEDC-31317P 1, Revision 1,

' James A. FitzPatrick Nuclear Power Plant SAFER /GESTR Loss-of-Coolant Accident Safety Analysis Report," Reference 5, referred to as the ECCS-LOCA Analysis, and; James A.

FitzPatrick Calculation JAF CALC RAD-00008, " Radiological Consequences of Design Basis Accidents at James A. FitzPatrick," Reference 6, referred to as the Dose Analysis.

This change request is limited to the changes necessary for operation at power uprate conditions. Additional margin associated with the computer models being used for safer analyses have not been used to relax requirements (e.g., ECCS pump flows) except w' required to support operation at uprated power. This change request includes no rer approval of plant operations using special features such as increased core flow or tr Extended Operating Domain.

}

Tb 'USAR provides a plant specific safety c"*.luation for power uprate that diset of the evaluations performed for power uprate. The information in the PUSAR is when directly applicable to a change in order to avoid repetition. PUSAR Table peak containment pressure for a LOCA as 41.2 psig. This is lower than the per pressure of 45 psig now identified in the Te:hnical Specifications. PUSAR Tar the Technical Specifications (i.e., pages 167,172,173,173a,188,193 and 19 changed to reflect the lower pressure. These changes are not being reques order to minimize the changes necessary for operation at uprated power.

• made after the issuance of the PUSAR.

II. DESCRIPTION AND SAFETY IMPLICATION OF THE PROPOSED CH l

l The Operating Ucense with its attached Technical Specifications repre conditions which the plant must conform with in order to assure publ'

- ~

as the protection of the environment. The Operating Ucense provide the authorizations and limitations for plant operation. The Technice safety limits, limiting safety system settings and limiting control se'

1. ^ o

Attachment il to JPN 92-028 l SAFETY EVALUATION

. Page 2 of 45

analyses and evaluations. The Radiological Effluent Technical Specifications contain the l provisions for limiting the release of radioactive materials to unrestricted areas during normal j operations.

The proposed changes were identified in a systematic review of the Technical Specifications.

The necessity for changes was determined using the generic reports LTRs 1 and 2, and the plant specific PUSAR. Where necessary, the supporting documentation such as the engineering report, the LOCA analysis and the dose analysis were used. These documents provide both generic and plant specific evaluations and, where necessary, reanalyses to support James A.

FitzPatrick operations at uprated power. The PUSAR is based on the generic format and content for power uprate licensing reports given in LTR 1. It diamanan the scope of the engineering and i safety evaluations performed for the James A. FitzPatrick power uprate.

The changes affect the operating parameters of the reactor, operational restrictions, setpoints for safety systems, analytical results and test requirements. There are also administrative changes.

The changes in each of these categories are summarized as follows:

e Reactor Parameters: The effect on reactor parameters is limited. Higher power is

, achieved by control rod pattom adjustments to increase reactor thermal power (changes A.2,3 and 4) in a more uniform (flattened) power distribution to increase steam flow without increasing core recirculation flow. This requires an increased reactor dome pressure (changes A.1 and 5) for adequate turbine inlet pressure.

. Operational Umits: The increased thermal power requires a change B.1) to the limitation

. on operation in the high power low flow portion of the power / flow map to limit therma!

I hydraulic instabilities and power oscillations.

. Setooints: The increased reactor pressure had a direct impact on the high pressure scram setpoint (changes C.4 and 8) and the safety relief valve setpoint (changes C.5 and 6).

Additionally, the bypass for the turbine stop valve closure and control valve fast closure scram was changed (changes C.1,2,3,7 and 9) in proportion to the increase in thermal power.

. Analysis Results: Analyssa of uprated power transients and accidents required changes to various technical specifications and their bases. Operational parameters and assumptions used in Feywe were revised (changes D.1,2 and 3) to reflect their use as initial conditions. Revised radiological analyses changed dose results (change D.7). The results of the accident ardin required revisions to property reflect plant capabilities (changes D.4,5 and 6).

l

. Testing: A number of changes to testing requirements resulted from power uprate. The Increase to reactor pressure had a direct effect on hydrostatic leakage testing pressure (changes E.3 and 4). The test pressure for HPCI and RCIC pumps was revised (changes E.1 and 2) to reflect SRV setpoints assumed in analyses.

. Administrative: Administrative changes (i.e., adding references, revising references and correcting anauwad errors) were also made (changes F.1,2,3,4,5,6 and 7).

No changes to the Radiological Effluent Technical Specifications were identified. For each Technical Specification change, this safety evaluation identifies the specific change proposed, j the purpose of the change and the safety implications of the change. This information is l presented for each page that is effected so that tM need for each change can be clearly j identifled and its safety significance evaluated. Rt forencing between the page changes is used to avoid unnecessary repetition of information.

The proposed changes, presented page by page, are as follows:

a

Attachment ll to JPN 92-028 SAFETY EVALUATION Page 3 of 45  ;

A. Reactor Parameters

1. Page 2, Specification 1.0.D - Definition of Hot Standby Condition

a. DESCRIPTION Replace the value "1,005 peig" with the value "1,040 psig.'

b. PURPOSE The change revises the definition of the hot standby condition to reflect the operating pressure of the reactor at uprated power condtions.

c. SAFETY IMPUCATIONS This change reflects a revision to reactor dome pressure and redefines the hot standby condition which is based, in part, on the reactor dome pressure at rated power. The reactor dome pressure is one of the initial parameters selected for evaluating power uprate. The basis for selecting this parameter and the safety implications are dia==M below.

The pressure in the reactor is measured at the reactor dome. An increase in the reactor vessel dome pressure is required to achieve good control characteristics for the turbine control valves at the uprated power condition.

Proper pressure regulation is provided if the control valves are <97% of their wide open position at the uprated power steam flow. This is equivalent to a turbira inlet pressure of 975 peig. Since there is a 55 poig steam line pressure drop at this in'et pressure, the reactor dome pressure must be at least 1,030 peig. A sligWy higher reactor dome pressure of 1,040 peig was chosen based on coordinc.clon of the reactor heat balance with the turbine capability. This value is trAd as the basis for defining the plant operating characteristics and performing plant safety evaluations at up'ated power.

Section 1.3 of the PUSAR identifies the increase in the reactor dome pressure to 1,040 peig. The safety implications of operating at this increased thermal power are die =*M throughout the balance of the PUSAR considering the thermal hydraulic parameters established from the heat balance at this power level. This safety evaluation defines the safety basis for concluding that there are no significant safety impacts for power uprate operation.

- d. ASSOCIATED CHANGES Changes A.5, C.4, C.5, C.6, C.8, E.3 and E.4 relate to this change.

e. REFERENCES Reference 3, Section 1.3

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

Attachrnent il to JPN.926 SAFETY EVALUATION i j Page 4 of 45 l 1  !

) 2. Page 5, Specification 1.0.N Definition of Rated Power l j a. DESCRIPTION i Replace the value "2,436 MWt* with the value "2,536 MWt" and add

"(Reference 1)" to the end of the sentence.

i

b. PURPOSE

The change revises the definition of rated power to reflect the increased j therrnal power at uprated power conditions and provides a reference to the

safety evaluation submitted in support of power uprate which has been added to Technical Specification page 6a as a reference.

I

c. SAFETY IMPUCATIONS l This change reflects an increase of 4.1% to the rated thermal power for power uprate and redefines the defirwtlon of rated power in the Tec,in ; cal i Specifications. The revised thermal power level la the basic parameter for all l l power uprate evaluations.

1 j The increase in thermal power was evaluated using the reactor heat balance to

establish thermal hydraulic parameters. The steam flow from the reactor vessel was increased to approximately match the original design flow. The i l 4.8% increase in steam flow with an increase in dome pressure of 35 peig i provides good turbine operating characteristics without any turbine 2 modifications. This power level is achieved with an increase in the power flow map along existing flow control lines.

l Section 1.3 of the PUSAR identifies the increase in reactor rated power to 2,536 MWt. The safety implications of operating at this increased thermal 4

power are discussed throughout the balance of the PUSAR considering the j thermal hydraulic parameters established from the heat balance at this power

level. This safety evaluation defines the safety basis for concluding that there j are no significant safety impacts for power uprate operation.

i The addition of the reference to the safety evaluation is administrative in nature and can have no safety impact.

l

d. ASSOCIATED CHANGES l

j Changes A.3 and A.4 relate to this change.

1 i

4

e. REFERENCES Reference 3.

]

i l a 1

4

Attachrnent 11 to JPN<92 028 SAFETY EVALUATION Page 5 of 45

3. Page 15, Bases 2.1 Fuel Cladding Integrity

a. DESCRIPTION in the first paragraph replace the value "2436 MWt" with the value "2,536 MWt*

In two locations.

l In the first sentence of the second paragraph, replace the word "given" with the word " described."

At the end of the second paragraph, add the following two sentences:

" Reference 1 evaluates the safety significance of uprated power operation at 2,536 MWt. This evaluation is consistent with and demonstrates the acceptability of the transient analyses required by Reference 2."

^

b. PURPOSE The changes revise the Bases to reflect the increased rated thermal power at 5

uprated power conditions and the associated supporting references. The first change identifies the twrw thermal power level and the reference, added on Technical Specification page 20, is the PUSAR.

c. SAFETYIMPUCATIONS The safety implications are dimuw in change A.2.

d. ASSOCIATED CHANGES Changes A.2, A.4, F.1, F.2 and F.7 relate to this change.  !

e. REFERENCES Reference 3, Section 11

" Reference 2" is " General Electric Standard Application for Reactor Fuel',

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

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Attachment 11 to JPN-926 SAFETY EVALUATION Page 6 of 45

4. Page 16, Bases 2.1 Ucensed Maximum Power Level

a. DESCRIPTION Replace the value "2436 MWt* with the value "2,536 MWt."

b. PURPOSE The change revises the Bases to reflect the tww maximum licensed power level at uprated power conditions.

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c. SAFETY IMPUCATIONS i The safety implications are discussed in change A.2.

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d. ASSOCIATED CHANGES Changes A.2 and A.3 relate to this change.

e. REFERENCES Reference 3 O

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5. Page 187, Bases 3.7 Suppression Chamber Blowdown

a. DESCRIPTION in the second paragraph, replace the value "1,020 psig" with the value "1,040 psig."

b. PURPOSE The change revises the Bases section to reflect the increased operating pressure. This section identifies the suppression chamber water volume function of absorbing the heat released from reactor coolant blowdown. The discussion currently identifies blowdown from 1,020 psig. This value is an editorial error since the intended blowdown is from the rated power pressure of 1,005 psig (i.e.,1,020 psia). This editorial error was in the original issuance.

The change indicates that the blowdown is from the pressure of 1,040 psig at uprated power.

c. SAFETY IMPUCATIONS The safety implications of the uprated power operating pressure are diam maM in change A.1. The esfety implications of the temperature rise associated with blowdown are dien maM in change D.4. There are no safety implications associated with correcting a typographical error.

d. ASSOCIATED CHANGES Changes A.1, D.4 and D.6 are related to this change.

e. REFERENCES Reference 3, Section 4.1 i

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Attachment 11 to JPN 92-028 SAFETY EVALUATION Page 8 of 45 B. Operational Umits  !

1. Page 134, Floure 3.51 - Thermal Power and Core Flow Umits

a. DESCRIPTION Replace the existing Figure 3.51 with the revised Figure 3.51.

b. PURPOSE Revise the core thermal power versus core flow operating map for operation at uprated power.

c. SAFETf IMPUCATIONS Uprated power will shift the line of core thermal power versus core flow which is used to control thermal hydraulic stability. The core thermal power value at which stability monitoring is not required is reduced by 0.98% at each core flow value on Figure 3.51. This reduction is in proportion to the ratio of rated power (2436 MWt) to uprated power (2536 MWt). This change assures that the relationship between thermal power and flow represented by "Une A' on Figure 3.51 will not change. Therefore, the thermal power cutoff point used, at various flows, to prevent single loop operation or to require stability monitoring for single and two loop operation remains the same. With no change to these values, the margin of safety remains unchanged. The safety implications have been generically evaluated in Section 3.2 of LTR-2, as noted in Section 2.4 of the PUSAR.

d. ASSOCIATED CHANGES No changes relate to this change,

e. REFERENCES Reference 2, Section 3.2 Reference 3, Section 2.4 t

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Attachment 11 to JPN-92 028 SAFETY EVALUATION Page 9 of 45 C. Setpoints

1. Page 11, Specification 2.1.A.3 Turbine Stop Valve Closure Scram Trip Setting

a. DESCRIPTION Replace the phrase "above 217 psig turbine first stage pressure" with the phrase "the reactor is at or above 29% of rated power."

b. PURPOSE The change replaces the turbine stop valve closure scram bypass setpoint  !

pressure with a reference reactor power. The actual pressure setpoint for this l bypass has varied over time as the turbine first stage pressure associated with  !

the 30% power level has changed. This variation makes a single pressure setpoint inappropriate. The change also makes the setpoint consistent with its power uprate safety bas:s, reactor power.

c. SAFETYIMPUCATIONS The setpoint for the scram (i.e., when the turbine stop valve reaches less than j or equal to 10% closure from full open) is not changed. The plant transient  ;

analyces at uprated power were performed with the setpoint for scram bypass I equivalent to 30% of uprated power as discussed in Sections 5 and 9 of the PUSAR. The proposed setpoint for the bypass is conservatively set at less than 29% of the new rated power.

J The generic approach to power uprate discusses bypass of this setpoint in .

Section F.4.2 of LTR-1. The setpoint bypass is chosen to allow operational  !

margin for a scram so that it can be avoided by transferring steam to the i turbine bypass system during turbine generator trips at low power. The transient events below the setpoint bypass are non-1,miting from a safety viewpoint allowing two options. The first is to keep the setpoint bypass at the current value (this roouires adjustment downward to reflect the higher steam 4

flow a uprated power). The second is to maintain the se point bypass at the same power level, perform plant specific analysis and reaqust instrument setpoints to reflect the increased pressure at that power level.

An analysis was performed for the James A. FitzPatrick plant assuming that the setpoint bypass was at the same power level. The setpoint will be conservatively maintained when reactor power is at or above 29%. This setpoint bypass is sufficiently high to avoid unnecessary scrams and below the analytically required setpoint for the bypass.

. The first stage turbine pressure can vary over the life of the plant at the setpoint power level. Calculating the pressure that is equivalent to the setpoint power level avoids revising the Technical Specifications when a variance occurs. The proposed change requires the pressure setpoint for the bypass to be calculated using current methodologies (Reference 7) that assure accurate control of the bypass. This change preserves the current margins of safety because it is conservative with respect to plant analyses, reemenari in Sections 5 and 9 of the PUSAR, and provides for more accurate control of the bypass.

Attachrnent !! to JPN-92-028 SAFETY EVALUATION Page 10 of 45

d. ASSOCIATED CHANGES Changes C.2, C.3, C.7, C.9, F.1, F.2 and F.7 relate to this change.

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e. REFERENCES i

1 l Reference 1, Section F.4.2 '

l Reference 3, Sections 5 and 9 Reference 7 l

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Attachment 11 to JPN-92-028 SAFETY EVALUATION Page 11 of 45

2. Page 19, Bases 2.1.A.3 - Turbine Stop Valve Closure Scram Trip Setting

a. DESCRIPTION in the last sentence of Section 2.1.A.3, replace the phrase " turbine steam flow is below 30%" with the phrase " reactor power is below 29%."

In the last sentence of Section 2.1.A.3, add the phrase ", consistent with the safety analysis discussed in Reference 1" at the end of the sentence.

b. PURPOSE The changen revise the Bases to reflect the change proposed to Technical Specifications 2.1.A.3 and 3.1.A (Table 3.1 1, footnote 4) to the value at which the turbine stop valve closure scram is bypassed and provide a reference to the PUSAR to identify a safety discussion of the supporting analyses.

c. SAFETY IMPUCATIONS The safety implications are discussed in change C.1.

d. ASSOCIATED CHANGES Changes C.1, C.3, C.7, C.9, F.1, F.2 and F.7 relate to this change.

e. REFERENCES Reference 1, Section F.4.2 Reference 3, Sections 5 and 9 Reference 7 D

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Attachment 11 to JPN-92-028 SAFETY EVALUATION Page 12 of 45

3. Page 19, Bases 2.1.A.4 - Turbine Control Valve Fast Closure Scram Trip Setting

a. DESCRIPTION In Section 2.1.A.4, add the phrase "and Reference 1" to the end of the next to last sentence.

l in the last sentence of Section 2.1.A.4, replace the phrase " turbine steam flow is below 30 percent" with the phrase " reactor power is belo.y 29 percent."

b. PURPOSE The changes revise the Bases to reflect the change proposed to Technical l Specification 3.1.A (Table 3.1-1, footnote 4) to the value at which the turbine l control valve fast closure scram is bypassed and provide a reference to the '

PUSAR to identify a safety discussion of the supporting analyses.

c. SAFETY IMPUCATIONS The safety implications dim me in change C.1 are applicable to the turbine l control valve fast closure scram trip bypass setting.

d. ASSOCIATED CHANGES Changes C.1, C.2, C.7, C.9, F.1, F.2 and F.7 relate to this change,

e. REFERENCES Reference 1, Section F.4.2 Reference 3, Sections 5 and 9 Reference 7 9

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Attri.J Twit il to JPN-92 028 SAFETY EVAUJATION Page 13 of 45 i

( 4. Page 27, Specification 2.2.1.A - Reactor Coolant System Umiting Safety System l Setting l

a. DESCRIPTION Replace the value "1,045 psig" with the value '1,000 peig."

b. PURPOSE The reactor high pressure scram setpoint was revised to reflect changes in the plant operating conditions during power uprate. The current reactor high pressure scram limiting safety system setting of 1,045 psig is increased by 35 psig to 1,000 psig to reflect the 35 peig increase in the steam dome during operation.

c. SAFETY IMPUCATIONS The acceptability of revised setpoint was confirmed by analysis as discussed in Section 5.1.2.1 of the PUSAR. This change in setpoint is within the design envelope and supports power uprate.

l l d. ASSOCIATED CHANGES Changes A.1 and C.8 relate to this change.

i e. REFERENCES Reference 3, Section 5.1.2.1 j l

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Attachment 11 to JPN.92-028 SAFETY EVALUATION Page 14 of 45

5. Page 27. Specification 2.2.1.B Reactor Coolant System Umiting Safety System l Setting  ;

a. DESCRIPTION j Replace everything after " settings shall be" except the last sentence and replace it with "$1,145 psig." The specification now reads " Reactor coolant ,

system safety / relief valve nominal settings shall be $1,145 psig. The l allowable setpoint error for each safety / relief valve shall be 21 percent." I i

b. PURPOSE  !

The reactor safety / relief valve (SRV) setpoints are revised to reflect changes in the plant operating conditions during power uprate.

c. SAFETYIMPUCATIONS The proposed SRV upper bound setpoint of 1,145 psig represents a 35 psig increase over upper bound setpoint proposed at the currently authorized power level (Reference 8). The 35 psig value is consistent with the increase in the dome pressure which will maintain the same simmer margin (i.e., the difference between the valve spring setpoint pressure and normal operating pressure) and be sufficient to maintain overpressure protection. Section 3.2 of the PUSAR discusses an er,4yse of the. limiting pressurization event, MSIV closure with a failure of valve position scram, using a SRV analysis setpoint (i.e., pressure at which the SRVa' open in the analysis) of 1,179 psig. This allows for a 3% margin over the setpoint of 1,145 poig to account for setpoint drift. SRV operation at 1,179 psig was also used in the transient analyses in Section 9.1 of the PUSAR. The ATWS event reen seem,1 ln Section 9.3.1 used a 1% margin over the 1,145 peig setpoint. This is consistent with the proposed setting.

The power uprate safety analyses included performance improvement features and equipment out of service assumptions as diaminaar1 in Section 1.3.2 of the PUSAR. Two safety relief valves out of service and a single upper bound SRV setpoint were two of the improvement features included. The proposed changes include the single upper bound setpoint since this is how the analyses were performed. The proposed change does not include the out of service allowance or change the 1% allowable setpoint drift since the purpose of the proposed change is to make only those changes noosesary for efficient uprated power operation. Changes that can be justified haranea of the i

additional margin in current analyses are not being requested as part of this amendment request.

~

The application (Reference 8), to use a single setpoint at 1,110 psig is pending. The evaluation of SRV containment dynamic loads for power uprate 4 using a SRV setpoint of 1,195 peig (i.e., pressure at which the SRVs' open in the analysis) is provided in Section 4.1.2.2 of the PUSAR. This evaluation is 4 based on the analysis described in Reference 9. Reference 9 was developed

in support of the pending Technical Specification change.

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Attachment 11 to JPN-926 SAFETY EVALUATION Page 15 of 45

d. ASSOCIATED CHANGES Changes A.1, C.6, E.1, E.2, F.1, F.2 and F.7 relate to this change.

e. REFERENCES Reference 3, Sections 3.2,9.1 and 9.3.1 Reference 8 Referen::e 9 l

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Attachment 11 to JPN-92 028 SAFETY EVALUATION Page 16 of 45

6. Page 29, Bases 1.2 and 2.2 - SRV Settings

a. DESCRIPTION Delete the first part of the last sentence that reads "The numerical distribution of safety / relief valve setpoints shown in 2.2.1.B ( 2 @ 1000 psi,2 @ 1106 psi,7

@ 1140 psi) is justified by analyses described in the General Electric report NEDCh 24129-1, Supplement 1," and replace it with "The numerical safety / relief valve setpoint shown in 2.2.1.B is justified by analyses described in the General Electric report NEDC-32016P.*

b. PURPOSE This change corrects the Bases by providing the referex:e used to justify the setpoint change for power uprate.

c. SAFETY IMPUCATIONS The safety implications of this change are dieneaM in change C.S.

d. ASSOCIATED CHANGES Changes A.1, C.5, E.1, E.2, F.1, F.2 and F.7 relate to this change.

e. REFERENCES Reference 3 1

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Attachment il to JPN-92-028 SAFETY EVALUATION i Page 17 of 45 l

7. Page 34, Bases 3.1 - Turbine Stop Valve Closure Scram Trip Settino l l

a. DESCRIPTION )

Replace "217 psig turbine first stage pressure (30 percent of rated)" with "29%

of rated reactor power" in the second sentence of the last paragraph.

I b. PURPOSE The change revises the Bases to reflect the change to Table 3.1 1 of j Specification 3.1.A.

c. SAFETYIMPUCATIONS l The safety implications are discussed in change C.1.

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! d. ASSOCIATED CHANGES l Changes C.1, C.2, C.3, C.9, F.1, F.2 and F.7 relate to this change.

e. REFERENCES l Reference 1, Section F.4.2 l Reference 3, Sections 5 and 9 l Reference 7 l.

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SAFETY EVALUATION l Page 18 of 45 i

8. Page 41a, Table 3.1-1 Reactor Protection System (SCRAM) instrurnentation i Requirement I

a. DESCRIPTION in the trip level setting column, for the high pressure trip function replace the value ' < 1045 psig" with the value " < 1,000 psig."

l b. PURPOSE The change revises the reactor high pressure scram setpoint to reflect the reactor operating pressure. The reactor high pressure scram limiting safety system setting of 1,045 psig is increased by 35 psig to 1,000 psig to reflect the l 35 psig increase in the steam dome during operation.

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c. SAFETY IMPLICATIONS t

The safety implications are discussed in change C.4.

l l d. ASSOCIATED CHANGES l Changes A.1 and C.4 relate to this change.

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l e. REFERENCES Reference 3, Section 5.1.2.1 l

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Attachment 11 to JPN-92-028 SAFETY EVALUATION Page 19 of 45

9. Page 42, Table 3.1 1 Turbine Stop Valve Closure and Turbine Control Valve Fast Closure Scram Trip Setting

a. DESCRIPTION in note 1.C, replace the value '30 percent" with the value "29 percent."

In note 4, delete the phrase " turbine first stage pressure is less than 217 psig or less than 30 percent" and replace it with the phrase "the reactor power is less than 29 percent."

b. PURPOSE

! The change provides the revised value at which the turbine stop and turbine i control valves can have their valve closure / fast valve closure ser sms l bypassed. This also establishes the limit on operating power w%n instrumentation is not available.

! c. SAFETY IMPUCATIONS l

The safety implications are discussed in change C.1.

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! d. ASSOCIATED CHANGES l Changes C.1, C.2, C.3, C.7, F.1, F.2 and F.7 relate to this change.

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e. REFERENCES l Reference 1, Section F.4.2

! Reference 3, Sections 5 and 9 l

Reference 7 i

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Attachment 11 to JPN 92 028 SAFETY EVALUATION Page 20 of 45 D. Analysis Results Page 139, Specification 3.6.C.1 Coolant Chemistry 1.

a. DESCRIPTION Replace the value "3.1 pCl/gm" with the value '0.2 pCi/gm."

b. PURPOSE The change reduces the reactor coolant system radioactivity operating limit to be consistent with new accident and transient analyses performed at power

uprate conditions.

l l c. SAFETY IMPUCATIONS l The analyses for James A. FitzPatrick power uprate conditions included an

) evaluation of transients and accidents as discussed in Section 9 of the PUSAR. These analyses assumed a value of 0.2 microcuries per gram of dose

. equivalent 1-131. The does analyses for the accident conditions are discussed i in Section 9.2 and summarized in Table 9-3 of the PUSAR. The limitatior.s on specific activity in the primary coolant system are the basis for evaluating thyroid and whole body doses from the main steamline failure outside j containment.

)

l The value selected for analysis is consistent with NUREG 0123, Revision 3, the i

BWR Standard Technical Specifications, Reference 10. Section 3/4.4.5 of NUREG-0123 Identifies 0.2 microcuries por gram as an interim limit selected i

by the NRC based upon a parametric evaluation of typical site locations. l 1

These analyses demonstrate the acceptability of operation at the 0.2 l microcuries per gram level. This change is more restrictive wa= lt reduces

the absolute value of the source term for the main steem line rupture.

d. ASSOCIATED CHANGES

] Change D.2 relates to this change.

e. REFERENCES

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Reference 3, Section 9.2 i Reference 6 Reference 10 4

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Attachrnent ll to JPN-92-028 SAFETY EVALUATION Page 21 of 45 4

2. Page 149, Bases 3.6.C and 4.6.C - Coolant Chemistry And Dose Analysis

a. DESCRIPTION in the second paragraph, replace the first sentence that says:

"In the event of a steam line rupture outside the drywell, with this coolant activity level, the resultant radiological dose at the site boundary would be 33 rem to the thyroid, under adverse meteorological conditions assuming no more than 3.1pCl/gm of dose equivalent 1131."

with 4

"In the event of a steam line rupture outside the drywell, & more

. restrictive coolant activity level of 0.2pCi/gm of dose equivalent l-

- 131 was assumed. With this coolant activity level and adverse meteorological conditions, the calculated radiological dose at the ,

site boundary would be less than 30 rem to the thyroid."

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b. PURPOSE l The change revises the Bases to make it consistent with the revised dose i analysis for main steam line break. This analysis was performed using a

- revised reactor coolant specific acuvity. The revised specific activity is now smaller than the specific activity allowed by the Radiological Effluent Technical Specification limit. The change makes this clear.

c. SAFETY IMPUCATIONS.

The safety implications are discussed in change D.1.

d. ASSOCIATED CHANGES Changes D.1, D.7, F.1, F.2 and F.7 are related to this change.

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e. REFERENCES l

Reference 3, Section 9.2 i Reference 6 Reference 10 e

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3. Page 188, Bases 3.7 Torus Water Volume

a. DESCRIPTION In the first paragraph, replace the phrase *a minimum suppression chamber" with the phrase "an approximate suppression chamber" and replace the value "105,600 ft3 ' with the value "105,900 ft3."

b. PURPOSE This change revises the discussion in the Bases of the suppression chamber water volume to reflect the uprated power containment analyses.

c. SAFETY IMPLICATIONS The revised suppression chamber water volume reflects the latest calculation, Reference 11, of the suppression chamber water volume at minimum water level. The calculated value is 105,930 ft3 . This change represents a minor increase, about 0.3%, in the value used in the original plant calculations.

There is no actual change to the water level or volume in the torus. The ,

change is the result of a more accurate calculation. The description of the I volume, as approximate, reflects the potential for slight changes in volume l with recalculation. The change in calculated volume has no safety impact.

d. ASSOCIATED CHANGES There are no changes related to this change. l

e. REFERENCES  ;

Reference 3, Section 4.1 Heference 11 9

Attachment il to JPN-92-028 SAFETY EVALUATION Page 23 of 45

4. Page 188, Bases 3.7 - Suppression Chamber Water Temperature

a. DESCRIPTION

Replace the third paragraph that says

"Using a 407 rise (Section 5.2 FSAR) in the suppression chamber water temperature and a maximum initial temperature of 957, a temperature of 1457 is achieved, which is well below the 1707 temperature which is used for complete condensation."

j with:

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' Containment analyses predict a 467 increase in pool water temperature, after complete LOCA blowdown. These analyses l assumed an initial suppression pool water temperature of 957 and a rated reactor power of 2536 MWt. LOCA analyses in Section 14.6 of the FSAR also assume an initial 957 pool temperature. Therefore, complete condensation is assured during a LOCA because the maximum pool temperature (1417) is less than the 1707 temperature seen during the Bodega Bay tests."

i b. PURPOSE I This change revises the discussion in the Bases of the calculated temperature rise in the suppression chamber based upon the uprated power analyses and

suppression pool temperature Umiting Conditions for Operation.

c. SAFETYIMPUCATIONS The peak calculated suppression chamber water temperature due to uprated power will increase due to the increased heat in the core. This change does not affect the current Technical Specifications but the Bases are being revised to reflect the current calculations and clearly identify the supporting documentation for the Technical W.

The Bases are clanfied to indicate the correct initial temperature assumed in

plant LOCA analyses. The existing Bases identifies a 407 increase in water

! temperature and a peak blowdown water temperature of 145Y. By inference, 4

the initial water temperature for the LOCA analysis was 1057 based on the i

allowable 107 rise for testing in the Technical Mdms. The LOCA analysis assumed that the plant was at the normal operating water 2

temperature limit of 957 as noted in FSAR Figure 14.6 7. At uprated power,

. Section 4.1.1.4 of the PUSAR identifies a post blowdown torus water temperature of 141T when the initial water temperature is 957. The rewritten Bases clarifles that the initial water temperature for LOCA analysis was 957.

There is no safety significanoe to this change since the LOCA analyses assumed thisinitialtemperature.

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1 Attachrnent 1112 JPN-92-028 SAFETY EVALUATION Page 24 of 45 l

The Bases are revised to reflect the increase in torus water ternperature due to the LOCA blowdown at uprated power conditions. Section 4.1.1.4 of the PUSAR identifies a 46T increase in water temperature from LOCA blowdown.

The Bases are revised to indicate that there is a 467 water temperature rise from 967 to 1417 due to LOCA blowdown. There is no safety significance to j this change since the torus water temperature remains well tmlow the 1707 l

temperature limit for complete condensation based on the Humboldt Bay and Bodega Bay tests. The calculated torus temperatures are within current design values. l

d. ASSOCIATED CHANGES Changes A.5, D.5 and D.6 are related to this change.

I e. REFERENCES Reference 3, Section 4.1 l

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Attachment ll to JPN 92-028 SAFETY EVALUATION Page 25 of 45

5. Page 188, Bases 3.7 ECCS Pump NPSH

a. DESCRIPTICM Replace the fourth paragraph which says:

Tor an initial maximum suppression chamber water temperature of 957 and assuming the normal complement of containment cooling pumps (two LPCI pumps and two RHR service water pumps) containment pressure is not required to maintain adequate not positive suction head (HPSH) for the core spray LPCI and HPCI pumps."

with:

Tor an initial maximum suppression chamber water temperature of 957, assuming the worst case complement of containment cooling pumps (one LPCI pump and two RHR service water pumps), containment pressure is required to maintain adequate not positive suction head (NPSH) for the core spray and LPCI pumps."

b. PURPOSE This change revises the discussion in the Bases of the NPSH requirements for the ECCS pumps. The design basis and analyses do not consider two trains to be available and the Bases should reflect the design condition.

c. SAFETY IMPUCATIONS The existing Bases section riam maan NPSH capabilities of the ~ECCS pumps for a case where there is no single failure assumed. The change is necessary har= ma no reenetfele of suppression chamber water temperature was performed for the case where two loops of containment cooling were available.

The design basis assumes a single failure and the Bases section has been rewritten to reflect this. The worst case is the failure of an Emergency Diesel Generator (EDG) system which results in the loss of 2 RHRSW pumps and 2 RHR pumps. Two RHRSW pumps and 1 RHR pump (the other is assumed to  ;

be discharging to a broken recirculation loop) remain. The suppression i chamber rises to 208.77, as indicated in Table 41 of the PUSAR, when the  !

single failure of one EDG is assumed. The pumps require up to 2 psig of torus pressure at this temperature as discussed in Section 4.1 of the PUSAR and Section 3.9 of the Erg;r.; ; g Report. The use of the containment pressure is consistent with the current plant design bases discussed in Section 6.5.1 of  !

the FSAR (Reference 13). The change does not represent a change in the  !

ability of the ECCS to perform its intended function. j i

Attachment il to JPN-92 028 l SAFETY EVALUATION Page 26 of 45 l

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d. ASSOCIATED CHANGES Changes A.5, D.4 and D.6 are related to this change.

e. REFERENCES Reference 3, Section 4.1 Reference 4, Section 3.9 Reference 12, Section 6.5.1 I

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6. Page 188, Bases 3.7 - Torus Temperature Umits

a. DESCRIPTION Delete the last paragraph on the page " Experimental data indicates that excessive steam condensing loads can ... to avoid the regime of potentially high suppression chamber loadings." Replace the deleted material with:

• Experiments indicate that unacceptably high dynamic l

l containment loads may result from unstable condensation when

suppression pool water temperatures are high near SRV discharges. Action statements limit the maximum pool temperature to assure stable condensation. These actions include

limiting the maximum pool temperature of 957 during normal operation; initiating a reactor scram if during a transient (such as a stuck open SRV) pool temperature exceeds 1107; and depressurizing the reactor if pool temperature exceeds 1207. T-quenchers diffuse steam discharged from SRVs and promote stable condensation. The preserce of T-quenchers and compliance with these action staterronts assure that stable l condensation will occur and containment beds will be acceptable.

NEDC-24361P (August 1981) summarizes eWyses performed to predict pool temperatures and containment loads during plant transionis using these temperature limits at a power level of 2515 MWt (104% of rated). NEDC-24361P also substantictos the acceptability of the plant design using the local pool limits of NUREG-0661. NEDO-30832 (December 1964) shows that SRV condensation loads are low compared to other design loads for plants with T quenchers. NEDO-30832 describes why local pool 1:mperatures need not be analyzed at a rated power level of 2536 MWt.'

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b. PURPOSE The change eliminates the discussion in the Bases of the peak suppression pool temperature limit of 1807 used to avoid excessive loads due to steam condensing during blowdown to the torus. This temperature limit was adopted by utilities in 1974 when the phenomenon was initially defined. The power uprate evaluation has identified plant specific and generic analyses that supersede this temperature limit and establish new justifications for the torus temperature limits. The purpose of the change is to reconcile these assessments and their relation to the torus temperature limits.

c. SAFETYIMPUCATIONS There are no safety concems associated with this change. The existing paragraph is no longer applicable. The paragraph being deleted was added in Amendment 16 as the basis for the torus temperature limits (i.e.,957 normal operating,1107 scram requirement,1207 isolated reactor depressurization) that were also added.

Attachment 11 to JPN 92-028 SAFETY EVALUATION Page 28 of 45 l l

The NRC identified a local temperature limit (i.e.,2007) to replace the bulk l torus temperature. This limit and its supporting bases were issued in NUREG-0661. Transient analyses performed for FitzPatrick (NEDC 24361P) demonstrate compliance with the acceptance criteria of NUREG-0661. These analyses assumed that the initial temperature of the torus water was 957, that reactor scram would be initiated at a water temperature of 1107 and that i depressurization would begin at a water temperature of 120*F, if the reactor were isolated.

Subsequent research and testing lead to a conclusion that local temperature limits were not required for plants with quenchers. The bases for this  !

conclusion is provided in NEDO 30832 which demonstrates that the condensation loads with quenchers over the full range of pool temperature up to saturation are low compared to loads due to SRV discharge line air clearing and LOCAs which have already been considered in containment design evaluations.

The conclusion that condensation loads do not require local temperature limits  ;

did not eliminate the need to limit the pool water temperature. NEDO-30832 has not eliminated the current licensing basis, NEDC-24361P, but has been relied upon to eliminate the need for further analysis of local pool temperature at uprated power conditions. Umits on water temperature assure that torus water temperature is maintained below the saturation temperature limits in NEDO-30832. Also, pool temperature limits have been used as initial conditions in transient and accident analyses that are part of the design basis for structures and equipment.

d. ASSOCIATED CHANGES Changes A.5, D.4, D.5 and F.7 are related to this change.

e. REFERENCES Reference 3, Section 4.1 Reference 15 Reference 16 i

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j' Attachrnent 11 to JPN 92 028 SAFETY EVALUATION Page 29 of 45 l

7. Page 193, Bases 4.7.A - LOCA Dose Analysis l

a. DESCRIPTION f Delete the first part of the fourth paragraph "The design basis loss-of-coolant i accident ... unlikely event of a design basis loss-of-coolant accident." Replace

! the deleted rnatorial with: l 1

" Design basis accidents were evaluated as discussed in Section 14.6 of the FSAR and the power uprate safety evaluation, Reference 18. The whole body and thyroid doses in the control

} room, low population zone (LPZ) and site boundary meet the

! requirements of 10 CFR Parts 50 and 100. The technical support

. center (TSC), not designed to these licensing bases, was also

analyzed. The whole body and thyroid dose acceptance criteria j used for the main control room are met for the TSC when initial access to the TSC and occupancy of certain areas in the TSC is restricted by administrative control. The LOCA dose evaluation, Reference 19, assumed: the primary contenment leak rate was 1.5 volume percent per day; source term releases were in accordance with TID 14844; and the standby gas treatment system filter efficiency was 99% for ha60 gens." I

b. PURPOSE i l

The change eliminates the discussion in the Bases of the specific dose results I I

from the LOCA dose calculation. A change was necessary ham m the dose analysis at uprated power changed the calculational models and results. A reference to the uprate power safety evaluation was also necessary. The 1 change discusses the assumptions used in the dose calculations in the same l level of detail for consistency. The results of the does calculations are  !

discussed generally since it is necessary to show compliance with the acceptance criteria and not calculational results. The results of the dose  ;

analysis will be included in an FSAR update following power uprate apprual.

c. SAFETYIMPLICATIONS The done analyses for power uprate were performed as described above as well as with additional methodologies and assumptions consistent with current NRC acceptance criteria in NUREG 0800, Revlalon 1. The analyses and specific dose results are discussed in Section 9.2 of the PUSAR. As noted in that section, access to the TSC is administratively restricted based upon measured activity. This control, applicable to all accidents, was initiated as a result of the MSLB dose analysis. The MSLB will result in an unacceptable thyroid dose in the TSC if it is activated immediately after the accident. Using more reat!stic activity levels (e.g., the design activity level of 0.11 pCi/gm of

&se equivalent 1-131 rather than the Technical Specification limit of 0.2 pCl/cm) would allow immediate access hamina the dose following a MSLB at the design actMty level would meet 10 CFR 50, egendix A, GDC 19 dose criteria. This approach assures compliance with federal guidelines and allows access in a reasonable time.

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Attachment 11 to JPN 92-028 SAFETY EVALUATION Page 30 of 45 l

! d. ASSOCIATED CHANGES l 1

r Changes D.2 and F.7 are related to this change.

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l e. REFERENCES Reference 3, Section 9.2 1 Reference 6 l l

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Attachment ll to JPN-92-028 SAFETY EVALUATION Page 31 of 45 E. Testing

1. Page 117, Specification 4.5.C.1 - HPCI System Surveillance Test Pressure

a. DESCRIPTION l Replace the value "1,120 psig" with the value "1,195 psig."

b. PURPOSE The change revises the HPCI test pressure to reflect the analyzed value at which the SRV could be set.

c. SAFETYIMPUCATIONS The HPCI must be able to deliver water to the primary system at the highest pressure allowed by the SRV. Ths peak pressure the primary system can attain corresponds to the setpoint and drift allowed before the SRV's act to a depressurize the system. The change to Specification 22.1.A discusses new l

setpoints for the SRV at 1,145 psig with a 1% setpoint error allowed. This setpoint is consistent with power uprate analyses referred to in Sections 32 and 9.1 of the PUSAR which assumed that the SRV would operate at 1,179 psig. However, other analyses have been performed assuming that the SRV operate at 1,195 psig, see Section 4.1.2.2 of the PUSAR. The revised test pressure for the HPCI pump is conservatively based on the highest analyzed pressure. This provides a margin of over 38 peig between the required delivery pressure and the test pressure. The proposed change is therefore conservative.

9

d. ASSOCIATED CHANGES Changes C.5, C.6 and E.2 relate to this change.

e. REFERENCES Reference 3, Sections 3.2,4.1.2.2 and 9.1 e

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Attachment ll to JPN 92-028 SAFETY EVALUATION Page 32 of 45

2. Page 121a, Specification 4.5.E.1 - RCIC System Surveillance Test Pressure

a. DESCRIPTION Replace the value '1,120 psig" with the value "1,196 psig."

b. PURPOSE Revise the RCIC test pressure to reflect the analyzed value at which the SRV could be set.

c. SAFETYIMPUCATIONS The safety implications are dim maad in change E.1.

d. ASSOCIATED CHANGES Changes C.5, C.6 and E.1 relate to this change.

e. REFERENCES Reference 3, Sections 3.2,4.1.2.2 and 9.1 l

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Attachment il to JPN-92 028 SAFETY EVALUATION Page 33 of 45

3. Page 147, Bases 3.6 and 4.6 Maximum Hydrostatic Test Pressure

a. DESCRIPTION in the last sentence on the page, replace the value "1105 psig" with the value "1,144 psig."

b. PURPOSE The change increases the peak hydrostatic test pressure to reflect the increased reactor operating pressure. The code allows hydrostatic testing to 1.1 times the operating pressure. The power uprate increase in the operating pressure by 35 psig to 1,040 psig results in a 39 peig increase in the peak allowable test pressure to 1,144 peig.

c. SAFETY IMPUCATIONS There are no safety implications because the new test pressure remains below the FitzPatrick reactor vessel and pressure boundary design pressure of 1250 psig and is significantly below the ASME code allowable peak pressure of 1375 psig. This value remains within the design of the system resulting, as discussed in Sections 3.1 and 3.2 of the PUSAR, in no safety concems. The safety implications of the increased operating pressure are described in change A.1.

d. ASSOCIATED CHANGES Changes A.1 and E.4 relate to this change.

e. REFERENCES Reference 3, Sections 3.1 and 3.2

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Attachment 11 to JPN 92 028 SAFETY EVALUATION Page 34 of 45

4. Page 172, Specification 4.7.A.2.d.(1) - Containment Laakage Test Pressure

a. DESCRIPTION Replace the value "1000 psig* with the value "1,035 psig."

b. PURPOSE The change revises the leakage testing criteria to reflect the new operating pressure.

1

c. SAFETYIMPUCATIONS The are no safety implications associated with this change. The revised pressure is based on the revised system operating pressure discussed in change A.1. This pressure is within the system design limit and is less than the pressure used for hydrostatic and leak rate tests discussed in change E.3.

The ASME code requires the hydrostatic test pressure to be less than the normal operating pressure for self seating valves. The differential between operating pressure and test pressure has been maintained. See changes A.1 and E.3 for a further discussion of safety implications,

d. ASSOCIATED CHANGES Changes A.1 and E.3 are related to this change.

e. REFERENCES Reference 3 1

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Attachment 11 to JPN 92-028 SAFETY EVALUATION Page 35 of 45 4

F. Administrative

1. Page6a, Specification AE References

a. DESCRIPTION Add a new specification that reads as follows:

"AE. References

1. General Electric Report NEDC-32016P, " Power Uprate Safety Analysis for the James A. RtzPatrick Nuclear Power Plant," December 1991 (proprietary)."

b. PURPOSE The definition was added to provide a reference to the PUSAR that was prepared in support of the power uprate application.

c. SAFETY IMPUCATIONS There are no safety implicat'ons associated with the addition of a reference. ,

1

d. ASSOCIATED CHANGES Changes A.3, C.1, C.2, C.3, C.6, C.7, C.9, D.2, D.6, F.2 and F.7 are related to this change.

e. REFERENCES Reference 3 3 .

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Attachment 11 to JPN 92-028 SAFETY EVALUATION Page 36 of 45

2. Page 20, Bases 2.1.C - References

a. DESCRIPTION in item 1, replace the word "(Deleted)" and insert the reference " General Electric Report, NEDC-32016P, " Power Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant", December 1991 (proprietary)."

b. PURPOSE The change adds a reference.

c. SAFETYIMPUCATIONS This reference is the PUSAR. It is added to Bases 2.1 to clarify the location of supporting information. It can have no safety significance Wauw it makes no changes.

d. ASSOCIATED CHANGES Changes A.3, C.1, C.2, C.3, C.6, C.7, C.9, D.2, D.6, F.1 and F.7 are related to this change.

e. REFERENCES Reference 3 O

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Attachment ll to JPN 92-028 SAFETY EVALUATION

, Page 37 of 45 1

3. Page 35, Bases 3.1.B Reactor Protection System

a. DESCRIPTION in the last paragraph, replace the reference "NEDC-31317P' with the reference j "NEDC-31317P including latest revision, errata and addenda."

I b. PURPOSE l This change revises the reference in the Bases to property reflect the ECCS-LOCA Analysis for pcmenprate. Section 4.3 of the PUSAR Identifies the j analysis performed and its applicability.

c. SAFETY IMPUCATIONS I The purpose of this change is to correct 'Jw report in the Bases to reflect the l

ECCS performance under all loss of coolant accident conditions to satisfy the 5

reqJrements of 10 CFR 50.46 and 10 CFR 50 Appendix K. The added i reference is discussed in Section 4.3 of the PUSAR where it is concluded that the analysis is done using NRC approved methods. This correction has no ,

j safety implication. j i i j d. ASSOCIATED CHANGES I Change F.8 in related to this change.  ;

l l e. REFERENCES l i

i Reference 3, Section 4.3 i Reference 5

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Attachment ll to JPN 92-028 SAFETY EVALUATION Page 38 of 45

4. Page 41a, Table 3.1-1 Reactor Protection System (SCRAM) Instrumentation Rsquirement

a. DESCRIPTION Move the trip function " Turbine Stop Valve Closure" from page 42 to the bottom of page 41a.

l

b. PURPOSE The trip function was moved from page 42 to the bottom of page 41a to place it with the balance of the trip functions. This is an editorial change made during the processing of the pages.

c. SAFETY IMPUCATIONS There are no safety impilcations associateo with the movement of text.

d. ASSOCIATED CHANGES No other change relates to this change.

e. REFERENCES None O

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Attachment ll to JPN 92 028 SAFETY EVALUATION Page 39 of 45

5. Page 188, Bases 3.7 Primary Containment

a. DESCRIPTION in the second paragraph, delete the phrase "the limit for complete

. condensation of."

1 In the fifth paragraph, delete the word " form" and the value "1307" and replace them with tte word "from" and the value "1057."

^

b. PURPOSE The changes, correct typographical errors in the text that were introduced during the amendment process.

The phrase "the limit for complete condensation of," is a repeat of an existing phrase that was added in Amendment 168. This change corrects this error.

l ne misspelling of the word "from" was introduced in Technical Specification Amendment 168 and is corrected here.

The value "1307" was part of the original Technical Speerfication Bases for restricting the temperature rise in the torus pool during the use of the RCIC, HPCI or relief valves. It reflects the condensation limit for blowdown of 1707, based on the Humboldt Bay and Bodega Bay tests, less the 407 pooi temperature rise associated with blowdown. Amendment 16 changed this value to "105Y" as part of the amendment that added torus pool temperature 3

limits. These limits addressed containment issues. The 1057 includes the i 107 pool temperature rise over the normal Cd7 limit that was allowed for RCIC, HPCI and relief valve testing. The 1057 was inadvertently changed

back to 1307 in Amendment 36. This change corrects this error,

c. SAFETYIMPUCATIONS There are no safety implications associated with the correction of a typographical error,

d. ASSOCIATED CHANGES No other change relates to this change.

e. REFERENCES None i

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Attachment 11 to JPNWG8 l SAFETY EVALUATION Page 40 of 45

6. Page 254-c, Administrative Controls Section 6.9.(A)4.b2

a. DESCRIPTION I l Add the word " revision," after the phrase 'NEDC-31317P, October,1986 j l including latest."

b. PURPOSE l

This change revises the reference used for administrative control to reflect the l

need to use the latest ECCS LOCA Analysis. Section 4.3 of the PUSAR identifies the analysis performed for power uprate and its applicability. That NEDC reference is Reference 5 to this safety evaluation. .

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c. SAFETY IMPUCATIONS l l

l The safety implications are discussed in change F.3.

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d. ASSOCIATED CHANGES Change F.3 relates to this change.

e. REFERENCES Reference 3, Section 4.3 Reference 5 i

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Attachment ll to JPN-92 028 i SAFETY EVALUATION Page 41 of 45

7. Page 285, Section 7.0 - References

a. DESCRIPTION Add references "(18) General Electric Report NEDC-32016P, " Power Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant," December 1991 (proprietary)."; '(19) James A. FitzPatrick Calculation JAF CALC RAD-00008, " Radiological Consequences of Design Basis Accidents at James A.

FitzPatrick," November 1991."; and "(20) General Electric Report GE NE 187-45-1191 P, " Containment Systems Evaluation," (proprietary)."

Delete from Reference (10) the phrase " Progress Report for Period Ending December 31,1966."

b. PURPOSE The identified references were used in mak!ng changes to the prior pages.

The deletion from Reference (10) is administrative. The deleted phrase was

, repeated twice and represents a typographical error made when adding the reference.

c. SAFETY IMPUCATIONS There are no safety implications associated with the addition of references or ,

correction of typographical errors.  ;

d. ASSOCIATED CHANGES Changes A.3, C.1, C.2, C.3, C.6, C.7, C.9, D.2, D.7, F.1 and F.2 are related to this change.

i e. REFERENCES Reference 3 Reference 6 i Reference 12 e

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i Attachment 11 to JPN 92 028 I SAFETY EVALUATION Page 42 of 45 l l

lit. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION 1

Operation of the FitzPatrick plant at a therrnal power of 2536 MWt will not involve a significant I hazards consideration as defined in 10 CFR 50.92, since it would not:

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

The James A. FitzPatrick nuclear power plant was reviewed for operation at a rated power of 2550 MWt at the time of its operating license, Reference 13. This review was based on the original design of the plant. Since that tirne, a number of safety lasues of a generic and plant specific nature as well as plant modifications have changed the originally reviewed design.

Generic criteria, methodologies and evaluation scope required to uprate BWRs up to 5%

were prepared by General Electric and submitted to the NRC in LTR 1. This was supplemented by the submittal of generic evaluations in LTR-2 to determine: which NRC and industry generic communications were applicable to power uprate and how they should be treated; analytical evaluations that could be generically approved; bounding evaluations of cei6penents and equipment, and; the effect of power uprate on safety margin. These generic evaluations are supplemented by plant specific evaluations. The Power Uprate Safety Analysis Report (PUSAR) describes the dependence placed on References 1 and 2, the additional analyses that were performed, the results of these additional analyses and overall conclusions on the safety impacts of power uprate.

The plant systems and components will be within design limits at power uprate conditions with minor modfications. At uprated power, the power plant will not be operated in a manner that is different from current operations except for limited changes to operating parameters such as primary system pressure, steam flow and feedwater temperature.

Setpoints are revised as tweenmy to reflect new operational conditions and analyses.

The ECCS-LOCA analysis using current practices demonstrates compliance with design and regulatory acceptance criteria at uprated power.

The radiological consequences of accidents have been evaluated using more current 1

moti-M-Wr with consistent assumptions and continue to meet acceptance criteria.

Compliance with NRC does criteria using current methodologies is discussed in Section 9.2 of the PUSAR. The effect of power uprate on does analyses now discussed in the FSAR were =>-"'#di assessed recognizing that power uprate increases doses in direct proportion to the 4.1% increase in thermal power. An increase of 4.1% to the calculated doses currently identifled in FSAR Chapter 14 Indicates that a reevaluation using the original methodology would have demonstrated compliance with current NRC dose criteria. A review of Table 14.4 2 indcates that, with the 4.1% increase, offsite doses would be sybstantially less than NRC allowable values. A review of Table 14.81 indicates that, with the 4.1% increase, control room doses would be substantially less than NRC allowables empt for the main steam line break (MSLB). However, the MSLS dose would drop wd; below allowables once the proposed change on allowable coolant activity (reduces the lim 4 by more than a factor of ten) is accounted for.

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Attachment 11 to JPN 92028 SAFETY EVALUATION Dage 43 of 45

2. create the possibility of a new or different kind of accident from any accident previously evaluated.

Operation at uprated power involves no changes to the manner in which the plant is operated. There we changes to operational parameters and setpoints but analyses of these identified no new failure modes or accident scenarios. The effects of transients and accidents fall within design capabilities. Systems and components are capable of operating and performing their safety functions at uprated power. No mechanisms for creating a new or different accident were identified.

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

The power uprate will not result in significant increases to primary system temperature and pressure due to postulated operating transients or accidents. These and other margins of safety have been discussed in the PUSAR, where it is demonstrated that there will be no reductions in the margin of safety because the plant will still meet its design and regulatory acceptance criteria. For example, the core will continue to be operated with the same margin to the safety limit minimum critical power ratio. Fuel thermal limits will continue to meet NRC acceptain criteria. Plant systems and equipment are designed for uprated power conditions and have been evaluated for their capability to perform 6 uprated conditions. They will continue to perform within design limits.

IV. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not adversely affect the Fire Protection Program at the FitzPatrick plant. Since there are no plant configuration or combustible load changes, there is no affect on the fire suppression or detecion system. The increase in thermal power will increase the normal source terms in the primary system but these will continue to be well below design values. The impacts on the ALARA program are therefore expected to be minimal. The changes will effect the environment but the impacts will be minimal. There will be no need to change the currently approved radioactive material discharge limits for gaseous and liquid discharges. The increases in radiological levels from the primary system will be proportional to the increase in thermal power. These are restricted to a new and lower limit of 0.E pci/gm of dose equivalent 1-131. This reduction is more than an order of mcgnitude. The thermal discharges to the lake will inct asse slightly and a request to modify the State Pollu'Jon Discharge Elimination System (SPDES) is currently planned. The limits in this permit will continue to be met.

V. CONCLUSION Because thp changes will slightly increase the consequences of a power dependent accident, they constitute an unroviewed safety question as defined in 10 CFR 50.50. The dose increases 4.1% with an increase of 4.1% in the thermal power level when analyses are performed using the same methodology.

Operation of the FitzPatrick plant in accordance with the proposed amendment has been assessed in the power uprate safety evaluation, Reference 3, and it has been demonstrated in accordance with 10 CFR 50.92 that the changes would not:

Attachment 11 to JPN 92-028 SAFETY EVALUATION Page 44 of 45

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

2. create the possibility of a new or different kind of accident from any accident previously evaluated; or

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

VI. REFERENCES

1. General Electric Ucensing Topical Report NEDC-31607P A ' Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," (LTR-1) (proprietary)

2. General Bectric Ucensing Topical Report NEDC 31984P " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," July 1991 and Supplement 1, October 1991 ,

(LTR 2) (proprietary) 3

3. General Bectric Report NEDC-32016P " Power Uprate Safety Analysis for the James A.

RtzPatrick Nuclear Pcwor Plant," December 1991 (PUSAR) (proprietary)

4. Stone & Webster Engineering Corporation " Core Power Uprate Engineering Report for James A. FitzPatrick Nuclear Power Plant," December 1991 (Engineering Report)

5. General sectric Report NEDC-31317P-1, Revision 1, " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR Loss-of-Coolant Accident Safety Analysis Report," August 1991 (ECCS LOCA Analysis) (proprietary)

6. James A. FitzPatrick Calculation JAF CALC RAD 00008, " Radiological Consequences of Design Basis Accidents at James A. FitzPatrick," November 1991 (Dose Analysis)

7. James A. FitzPatrick calculation JAF 91-002, Revision 1, " Turbine First Stage Pressure Scram Bypass Setpoint (Uprated Condition)," November 1991

8. ' NYPA letter, J. C. Brons to NRC dated December 20,1989 (JPN-89-084) regarding proposed enangas to the Technical Specifications for S/RV single setpoint performance q (JPTS-89017)

9. General Electric Report NEDC-31897P 1, Revision 1, " Updated S/RV Performance 1 Requirements for the James A. FitzPatrick Nuclear Power Plant," October 1991 (proprietary)

10. NUREG 0123, " Standard Technical Specifications For General Bectric Boiling Water Reactors," BWR/4 j

11. JAF Document No. 22A5747, Revision 1, " Containment Data," 1979 l

12. General Bectric Report GE NE 187-451191, " Containment Systems Evaluation,"

(ptcgrietary)  ;

13. James A. AtzPatrick Nuclear Power Plant Updated Final Safety Analysis Report I

14. James A. RtzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements

15. General Bectric Report NEDC-24381P, " James A RtzPatrick Nuclear Power Plant Suppression Pool Temperature Response," August 1981 (proprietary)

Attachment 11 to JPN-92-028 i SAFETY EVALUATION Page 45 of 45 1

1

16. General Electric Report NEDO-30832, "Eimination of Umit on BWR Suppression Pool Temperature For SRV Discharge With Quenchers," December 1964 I

t

i l ATTACHMENT IV to JPN-92-028 l

l GE AFFIDAVIT ON NEDC-32016P I

, (JPTS 91@5) i I

New York Power Authority l

JAMES A. FITZPATRICK NUCLEAR POWER PLANT

! Docket No.fMM33 DPR 59 f

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't ATTACHMENT V to JPN-92 028 i

i l

l NEDC-31317P 1

" JAMES A. FIT 2 PATRICK NUCLEAR POWER PLANT SAFEH/

GESTR LOSS-OF-COOLANT ACCIDENT SAFETY ANALYSIS REPORT" (JPTS 91-025)

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NewYork Power Authority JAMES A. RTZPATRICK NUCLEAR POWER PLANT Docket No. 50333

) DPR-50

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

I GENERAL ELECTRIC COMPANY AFFIDAVIT f

I, DAVID J. R0 BARE, being duly sworn, depose and state as follows

i

! 1. I am Manager, Plant Licensing Services, General Electric Company, and have been delegated the function of reviewing the information described in paragraph 2 which is sought to be withheld and have been authorized  :

to apply for its withholding. i

2. The information sought to be withheld is contained in General Electric l Report NEDC-32016P " Power Uprate Safety Analysis for the James A.

i FitzPatrick Nuclear Power Plant", dated December 1991. The GE

Proprietary portions of this report are identifiable by the "GE l Proprietary Information" designation at the top of the page.

3. In designating material as proprietary, General Electric utilizes the definition of proprietary information and trade secrets set forth in the American Law Institute's Restatement of Torts, Section 757. This definition provides:

"A trade secret may consist of any formula, pattern, device or compilation of information which is used in one's business and which gives him an opportunity to obtain an advantage over competitors who do not know or use it...A substantial element of secrecy must exist, i so that, except by the use of improper means, there would be difficulty in acquiring information...Some factors to be considered in determining whether given information is one's trade secret are (1) the extent to which the information is known outside of his business; (2) the extent to which it is known by employees and others involved in his business; (3) the extent of measures taken by him to guard the secrecy of the information; (4) the value of the information to him and to his competitors; (5) the amount of effort or money expanded by him developing the information; (6) the ease or difficulty with which the information could be properly acquired or duplicated by others."

4. Some examples of categories of information which fit into the definition of Proprietary Information are:

a. Information that discloses a process, method or apparatus where prevention nf its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information consisting of supporting data and analyses, including test data, relative to a process, method or apparatus, the application of which provide a competitive economic advantage, e.g.,

I by optimiz~ation or improved marketability;

GENERAL ELECTRIC COMPANY AFFIDAVIT l l

c. Information which if used by a competitor, would reduce his l expenditures of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality or licensing of a similar product;

d. Information which reveals cost or price information, production l capacities, budget levels or commercial strategies of General Electric, its customers or suppliers;

e. Information which reveals aspects of past, present or future General Electric customer-funded development plans and programs of potential commercial value to General Electric;

f. Information which discloses patentable subject matter for which it 1 may be desirable to obtain patent protection; l

g. Information which General Electric must treat as proprietary according to agreements with other parties.

5. Initial approval of proprietary treatment of a document is typically made by the Subsection Manager of the originating component, the person who is most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within the Company is limited on a "need to know" basis and such documents are clearly identified as proprietary.

6. The procedure for approval of external release of such a document typically requires review by the Subsection Manager, Project Manager, Principal Scientist or other equivalent authority, by the Subsection Manager of the cognizant Marketing function (or delegate) and by the Legal Operation for technical content, competitively effect and determination of the accuracy of the proprietary designation in accordance with the sti'ndards enumerated above. Disclosures outside General Electric are generally limited to regulatory bodies, customers and potential customers and their agents, suppliers and licensees then only with appropriate protection by applicable regulatory provisions or proprietary agreements.

7. The document mentioned in paragraph 2 above has been evaluated in accordance with the above criteria and procedures and has been found to contaic information which is proprietary and which is customarily held in confidence by General Electric.

8. The information to the best of my knowledge and belief has consistently been held in confidence by General Electric Company, no public disclosure has been made, and it is not available in public sources.

l GENERAL ELECTRIC CONPANY AFFIDAVIT

8. All disclosures to third parties have been made pursuant to regulatory provisions of proprietary agreements which provide for maintenance of the information in confidence.

9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to the competitive position of the General Electric Company and deprive or reduce the availability of profit making opportunities. A substantial effort has been expended by General Electric to develop this information.

GENERAL ELECTRIC CONPANY AFFIDAVIT STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

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l David J. Robare, being duly sworn, deposes and says:

i That he has read the foregoing affidavit and the matters stated therein are truly and correct to the best of his knowledge, information, and belief.

StD .

Executed at San Jose, California, this ?.3 day of SE.cE;MB E R 19 )l .

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f bd David J. Robare

%L General Electric Company I

-Subscribed and sworn before me this ,#4 day of hd 19 W .

[? I OFFICIAL SEAL l W

g%QE , , f Notary PGblic, State of California SANTA CLARA COUNTY

) f My Comm. Exp. Mar. 26,1993 !

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l ATTACHMENT VI to JPN-92-028 l

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l l GE AFFIDAVIT ON NEDC-31317P-1 l

l l IJPTS91-025) l l l l

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i NewYork Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 50 O

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ATTACHMENT Vil to JPN 926  :

MARK-UP OF CURRENT TECHNICAL SPECIFICATIONS j

! (JPTS 91-025) l l

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l New York Power Authortfy JAMES A. FITZPATRICK NUCLEAR POWER Pl. ANT Docket No.50333 DPR-59 4

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s JAFNPP 1.0 (cont *d)

C.

Celd Condition - Reactor coolant temperature 3. i 6:212*F. Instrument Chamael

- As lastrument chamael D.

means as arrangement of a sensor and musil-Mot Et== % C- *iria= - Not Standby condition lary equipment required to generate and mesma operation .with ecolant temperature > 212*F, transmit to a trip system a slagte trip the Mode Switch t_up/_ Wet Standby and signal related to the plant parameter  ;

reactor presser monitored by that lastrument chamael, E.

( {pel'g. 4.

Immediate - Immed e h that the required Instrument Check - As lastrument check is a l !

action will be laitiated as soon as practicable qualitative determination of acceptable  !

considering the safe operaties of the malt and operability by observation of instrument i the importance of the required action, behavior durlag operation. This determina-F.

tion shall laclude, where possible, compar-Imatr-- -tation 1som of the lastrument with other independent  !

instrusients measuring the same variable.

1.

Feactional Test - A smactional test is the 5.  !

manual operaties er laitiation of a system, Instrument Channel Functional Test - As suhayates, or component to verify that it instrument chamael functional test means the functiona eithis design tolerances (e.g., lajection of a simulated signal lato the Coe maa d start of a core spray pump to instrusset primary sensor where possible to verify that St resa and that it pumps the verify the proper lastrument chamael re-

• required volume of water). sponse, alarm and/or initiating action. I

2. 6.

Logic System Function Test - A logic system Instrument ch= - 1 Calibration - Am lastrument chaemel calibration means the functional test means a test of relays and adjustment of as lastrument signal output so contacts of a logic circuit from sensor to i that it correspeeds, withia acceptable activated device to ensure components are j range, and accuracy, to a known value(s) of operable per desige intent. Where practi-the parameter which the lastrument cable, action will go to completions i.e., l pumps monitors. Calibration shall encompass the entire instrument chamael including j actuation, alarm er tely. i I

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AmendmentNo.f.I4 b 2

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i JAFMPP 1.0 (cont *d) opened to perform necessary operattomal \

activities. is the period of time between the shutdown of the unit prior to refuelleg and the startup of the l

2. Plant subsequent to that refuellag.

At least one *door la each airlock is closed and sealed. R.

i SattLY_klRILS - The safety limits are limits

3. All. automatic costatemsat isolation valves within which the reasonable malatemance of the i I

are operable er de-activated in the isolated fuel cladding integrity and the reactor coolant i position.

system latogrity are assured. Violation of such i

a limit is cause for unit shutdown and review by

4. All blind flanges and manways are closed. the Atomic Energy Commission before resumption of unit operation. Operation beyond such a limit '

H. Bated Power - Bated power refers to operation at may not la itself result in serious consequences a reactor power of test. This is also but it indicates an operational deticiency subject to regulatory review.

2.,53 termed loe p.reene - r and is the maan - - r level autaora_ _ my the operating license. Rated S. Etcondafr_CenLRIERRRL_lRit9t112 - Secondary steam flow, rated coolant flow, rated nuclear containment latogrity means that the reactor system pressure, refer to the values of these parameters when the reactor is at rat building is intact and the following conditions power are met O. Reactor Power Operatlea - Reactor power operation h tt N A. .

is any operation with the Beode Switch la 1 At least one door in each access openiaq is closed.

Startup/Ilot Staaer or Rua position with the reactor critical and above 1 percent rated 2. The Standby Gas Treatment System is operable.

thermal power. '

3. All automatic ventilation system isolation P. Reactor Vessel Fransare - Unless otherwise valves are operable or secured la the indicated, reactor vessel pressures listed la the isolated position.

Technica#1 Specifications are those measured by the reactor vessel steam space sensor. T. Survalliance Frequency - Periodic '

9 Refueling Outage - Refuellag outage -

Amendment No.

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< JAFNPP AD. Core Operating Umits Report (COLR)

Z. Top of Achve Fuel This report is the plant-apaMc document that provides the core

The Top of Aceve Fuel, corresponding to the top of the ennched operating limits for the current operating cycle. These cycle-

! luel column of each fuel bunsSe, in located 352.5 inches above specMc operating luvuts shall be deterrruned for each reload weasel zero, whu:h is tie lowest point in the inside bottom of the cycle in accordance with Specdicahon 6.9.AA. Plant operation reactor vessel. (See General Bactric drawing No. 9tennannn.) wehen those operating irruts is addressed in indudual Techrucal AA. Rod Dansdy -

Rod density is the number of control rod notches inserted l empressed as a fraction of the total number of control rod notches. Au rodslug yinsertedis a contSelon representing 100 T me.d \

percent rod denemy. ~

i WB. Purge-Purging -

Purge or Purging is the controlled process of dacharging air or gas from a conenement in such a rnanner that replacement air or gas is required to punfy the condnement.

AC. Venhng Venhng is the controlled process of releasing air or gas from a e conEnement in such a rnanner that repiecement air or gas is not prowded or required.

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INSERT 1 i

1- . AE. References

1. General Electric Report NEDC-32016P, ' Power Uprate Safety Analysis for the James A. FazPatrick Nuclear Power Plant," December 1991 (proprietary).

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JAFMPP 2.1 (cont *d)

2. Reactor _Matst_LDE Level Scras his EsLLing l Reactor low water level . scram setting shall be 1 177 in. above the top of the active final (TAF) {

at normal operating conditions.

Q% r 00 3. Turhine slog _Yalva_ Closure. Scram._Irla_ Selling ObW 0 Turbine stop valve scram shall be f 10 percent IPMed PO W - #'

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• II"' * '_ ' " / " * * "* " ""* * ^ ~

4. Turbine _ControLYalve feat Closure _ Scram Irie SaLLing Turbine control valve fast closure scram control oil pressure shall be set at 500 < P ( 8 50 ps ig.

5. Main _ Steam Line_ Isolation _Yalyt. Closure _ Scram Iris.StLLine Main steam line isolation valve closure scram shall be ( 10 percent valve closure f rom full open.

, 6. Mala _ Steam Lina _ Isolation Yalva_ Closure on_Lew Eramaura When in the run mode main steam line low pressure initiation of main steam line isolation valve closure shall be 2825 psig.

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JAFNPP BASES 2.1 FUELCUODINGINTEGRITY q3s5% The abnormal opprationet transiones apphcable to operationThe of most limiting transients have been analyzed to d tho FitrPatrick Unit hope been analyzed throupiout the spectrum which result in the largest rar* reann in CRITICAL PO g of planned concttion operating consfilons up to the thermal power The type of transients evmW were increase in press MlAlt. The analyseewere bened upon power, poonive reactmty insertson, and rmaant terrporatur eccordancewehenoperasngma g currentloadBnelimitanefysis. Inadttelon 2,536_'"'"powenin%=

decromes. The Imtmo transeent yields the largest delt P8" #

pJ.isthe When added to N Sahny unt, the required operating l

""'"""* MCPR in the Core Operating L.imes Report is obtained. {

NN "UI ~

The eve nasenn of a given transient begms with the s L

The trenaient analyses perk rmed for each ribed parameters shown in the current reload analyses and Reier in that are inptd to the core dynamic behavior transaan Reference 2. Models and model consenraelem are programs desenbodin Psference 2. The output of these i described in this reference. As nearmammariin Reference 4, theprograms along with the initial MCPR form the input for the core wide transient analysis for one reckculation pump operation further analyses of the thermally Imted buncSe with a is conservasively bounded by two4oop operasion analysis, and channel transeent thermal hydraulic code. The pnncipal re the SouNiependent rod block and scram seapoint equasions are ' the eve wanti is the rartr*no in MCPR ch by the transsent.

artamead for one pump operatiort M 7.,

Fuel clareing integrity is assured by the appbcable operating simit MCPR for steady state condllions given in the Core

operasing umns Report 00tJg. These operating limit MCPR's

are derived from the aanahs=*ied iuss candding insegnty Safety Umit, and an analysis of abnormal operational transients. For

! any abnormal operating tranment analysis evenatinn with the indial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Umit MCPR at any time during the transient.

l Amendment No.1s,64,3d',plf 1pf

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L INSERT 2

' Reference 1 evaluates the safety significance of uprated power operation at 2,536 MWt. This evaluation is consistent with and demonstrates the acceptability of the transient analyses required by Reference 2.

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2.1 BASES (Cont'd) A. Trip Settings l The MCPR operating limits in the COLR are conservatsvely assumed to exist prior to initiation of the transsents. The bases for individual r. rip settings are discussed in the following paragraphs. i This choice of using consentative values of controlhng parameters and

' initiating transients at the ds=ign power level, prawnma more possmshc I* ""#*" 'IE ""I'D i answers than wodd reset by using expected values of control parameters and analyzin0 at hig$er power levels. a. IRM Flux Scram Trip Setting l -atate ation h W WWis not pwmind h ,

The IRM system consssts of 8 chambers,4 in each of analysis to support operation at various power and Sow r m .. _

.p. the reactor protection system logic channels. The has considered operation with allher one or two redrenW pumps. , ,

!n summary: range of power level between that cowwed by the SRM and the APRM. The 5 decades are covered by

. The abnormal operahonal transsents were analyzed to the the IRM by means of a range switch and the 5 t boensed maximum power level. =

decades are broken down into 10 ranges, each i besng one-half of a decade in size. The IRM scram

. The W nasmum power g1$3g trip sethng of 120 devaions is achve in each range of J_ ,

. Analyses of transients employ --- _ , conbervamie values of theIRM. Forexample,if theinstrumentwereon t i

the controging rear *w parameters. Range 1, the scram sethng would be a 120 divisums for that range; likewise, if the instrument were on g .

The analyncal procedures now used result in a rnore logical range 5, the scram would be 120 divisions on that  ;

h h aNemative method of assuming a W W range. Thus, as theIRM ss ranged up to power in conjunchon with the awpartari values for the accor.6 the incmase in pows M, the scram p,,,,,,,,

trip setting is also ranged up. The most significant i sources of reachvity change during the power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of j change of power is slow enough due to the physical j limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor shutdown well before any Safety Umit is exceeded i t

Amendment No.14, )K,2T,)d 1[

16

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2 ';CS (cunt'd) JA!

3. Turtrir.s Step Valve Closure Screm Trip Settinsts 3. Ma,ta Steau Line Isolation Yalve Closure Scrass Trio

, Ectrina n e turbine stop valve closure scrae' trip anticipates tie.: pressure, sieutron flux and heat flux increase that The low presse sotation of the main steam times could rer. nit frac rapid closure of the turbiase stop at 825 pet vided te give protection osminst

valves. Uith a serem trip setting of g 10 percent of rapid reac surizattua and the resulting valve closure frcm futt epen, the resultant increase la rapid cae of tale vessel. Advantage was taken of surface li::ac flux is 11mitad such that IE:PR remains the scram , ture iMich occurs whom the main steam aleng the Safety Llantt even derSog the worst case line asetet npilvhs are closed, to provide for i (@Cd or "trannicit I".iis scre.m that assumesulien to bypassed the gturbine M =_bypass is closed,m_ reactor l reactor shutdonne so that high power operet ten at t oi, u i " _ jprosauru does not oceier, thus providing pro- 3 y pewy is -9t% uf rated, as seemeurony turbine first stage T tenction for clie faset ctedding integrity safety limit.

I gNg prescutey ConTi d wih M. % gno g j Operation of the reactor at pressures louer than 825 dWCuE io Setett.n(.tl. Pais reagistre9 tiest the Reactor Hede Suitch be in the

4. Tothine Control Valve Fast closure Scres Tris Settien Startup position uttere protection of the fuel clad.

ding isetsgrity safety limit le provided by the APRH Th a tur ine centret valve set sleeure serem antici- high neutron flux scrasa and the IRN. Thus, the com-potes the pressure, neutrom- flux, and heat flier in- bination of main steem line low pressure isolation crease that could result from fast closure of time tur. and isolation valve closure scrass assures the avat!-

bine ccatret valves due to lead rejection exceeding clie ability of neutrosi flux scram protection over the capat>1tity of the turbine hypese, h e peacter Protec- entire range of applicability of the fuel claddina tion Systas initiates a scrass winen fast closure of the integrity safety. liselt. In addition, the isolation I

control valves is initiated by clie fast acting solenoid valve closure scram anticipates the pressure and valves. Tlate is achieved by the acties of the fast ftum transients which occur during normat er inad-i acting solenoid valves la rapidly reducing' hydraulic vertent isolation valve cloeisrap. With the scrases

. control uit pressure at time main tearbine centrol set at s 10 percent valve closure, there is no in- f vetve actuator dine diamp valves. This less of pree- crease la neutron flus. -

sure is sensed by pressure switches whose contacts foram the one-out-of-two-twice logic laput to the re ~ 6. Hein stesse Line Isolation valve Closure on Emma Pressure c.ctor protectica system, This trip setting, a itsninstty ',0 percent greater closure time and a dif- The teu pressure isetation setnimuss limit at 825 psig fer me valve characteristic fress that of tlee turbine was provided to give protection against fast reacrar  !

stop velve, cordsine to produce transients very similar depressurization and the resulting rapid cooldoun et l

and no smore severe then for the stop valve. IIe signifi. the vesset. Advantage was taken of the scrase feature cent change ist MCPR occurs. Relevant transient. tditek occurs when the mein steam line isolation valves (g analyses are discussed in Sectina 14.5 of the Finag Safor ; Analysis Reporta h is scram to bypassed ubem are closed to provide for reactor shutdoien so that

• Peration at pressures loieer th n thoso specified 1i

( k g l ^__'.:. c r r Mc: Er 5 *- " ;rr-" of rated, as the thernet hydraulic safety. timit does not occur, casund by tua*etes.: first stage pr.es_ sere. altlicugh operation at a pressure lever than 825 psig

~ ~ - - y woestd not necessarily constitiste an unsafu condit toes.

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INSERT 3 General Electric Report, NEDC-32016P,

• Power Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant *, December 1991 (proprietary).

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JAFMPP ,

1.2 and 2.2 BASEM h reactor coolant pressure bowadary integrity is an important barrier la the prevention of uncon- ANSI Code permits pressure transients up to 20 percent I trolled release of fission products. It is over the design pressure (120% a 1,150 - 1.3eo peg g),

h safety limit pressure of 1,375 psig is referenced ,

essential that the latogrity of this boundary be protected by establishlag a pressure limit to be to the lowest elevation of the Rocctor Coo 13nr Syst==.

observed for all ob rating corditions and whenever there is irradiated feel in t'te reactor vessel. j The pressure safety limit of 1,325 peig as measured h current reload analysis shows that the main steam by the vessel steam space pressere ladicator is isolation valve closure transient, with flus scram is ,

equivalent to 1,375 pelg et tt,e lowest alevation of the most severe event resulting directly in a reactor cookast system pressure increase. The reactor vessel l the Reactor Coolant liyeten. The 1,375 pelg salue is derived from the design pressures of the reactor pressure code limit of 1.375 psig, given la FSAn Sectica 4.2, pressure vessel and reacter coolest system piping. is above the peak pressura produced by The respective desige pressures are 1250 psig at the event above. Thus, the pressure sn aty limit 575*F for the reacter vessel,,1148 psig at 568*F (1.375 psig) is well above the peak pressure that can for the recirculation section plying and 1274 psig result from reasonably espected overpressure tran- l staats. (See current reload analysis fo the curve at 575* for the discharge piping. The pressure  !

safety limit was chosen as the 1 ewer of the produced by this analysis.) Reactor pressure is pressure transients permitted by the applicable continuously ladicated in the control room during  !

operation. '

daalge codes: 1965 Ases meller and pressure vessel Code,Section III for prcosure vessel and 1969 ANSI 331.1 Code for the reactor coolant system piping. A safety limit is applied to the Residual Heat l The ASME Bo!!er and pressure Vessel Code permits Removat System (RNRS) when it is operating in the f

pressure transleets up to le percent over desige shutdown cooling mode. When operatlag la the shut- i pressure (110% a 1,250 - 1,375 psig) and the 3ews coollag mode, the RHRS is included la the reactor coolant system.

9eme  : -- rir:2 eietr!tztie: er s trfread:: e

-- ! rt e " .-- != ? 2.1.5 '2 * ! 20 p;i, 2 m_

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11'^ --!} 5: ^ :1!!!:d by - : y::: f:::;it d ;  ;

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.;.; ~.22:t; ; ::y :t ~~~"' 22:3^ 2, T_;;?- _ t A and assures that the structural acceptance criteria '

set forth la the Mark I Containment Short Term Program are satisfied.

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Amendment No. , 4e I4 29

INSERT 4 The numerical safety / relief valve setpoint shown in 2.2.1.B is justified by analyses described in the General Electrie report NEDC-32016P.

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i JAFNPP 3.1 SASES (cont *d) i is discharged from the reactor by a scram can be acc M ated la the discharge piping. Each scram The IBM high fluz and APRM 6:

discharge lastrument volume accommodates la escess of 15% power scrans provide adequate coverage in the startup and latermediate range.

34 gallons of teater and is the low point la the piplag. Thms, the IBM and APRM systems are required No credit was tahem for this volume ia the to be operable modes. in the refuel and startup/ hot standb y f design of the dischegge piplag as conceras the enount The AraM 6120% power and flow referenced I of water which must be accommodated durlag a scram. scrans provide required protection la the power range I (reference FSAR Section 7.5.7). The power range is Durlee normal operation the discharge volume is covered only by the APENs.

eardy, however, should it fill with water, the water not required la the rum mode.Thus, the IBM system is ,

l discharged to the piping from the reactor could not ,

L he acc===adated, which would recent la slow scram The high reactor pressure, high drywell pressure times or partial control ted lasertion. To preclude reactor low water level and scram discharge solume this occurrence, level detection lastruments have high level scrane are required for startup and rum hees provided la each lastrument volume which alarm modes of plant operation.

They are, therefore, {'

and setan 34.5 the reactor when the volume of water reaches operation.

gallons. required to be operational for these modes of reactor i

As ladicated above, there is sufficleat volume la the piping to accomunodate the scram without  ;

impairment of tha scram times or amesat of lasertion The requirement to have the scram fonctions indicated of the control rods. la Table 3.1-1 operabia la the refuel mode assures This function shuts the reactor down while sufficiest volemme renales to accommodate that shiftlag to the refuel mode during reactor power t the discharged water and precludes the situation la operation does not dialaish the protection provided which a scram would be required but not be able to by the Reactor Protection System.

perform its function adequately. ~-

Turbine stop valve closure occurs at 10 percent of A Source Range Monitor (SAM) System is also provided valve closure, selow ;7 7 6 : .. ." " _. C h . ; . ; _ m 1

to supply additlosal neutros level information during r P^ ;;;:x; g :st4, the scram sigast due to turbine stop val _

startup graph 7.5.4but FSAE).

has no scram functions (reference para- 4 fluz and pressure ser closure is bypassed because the reactor. are adequate to protect the t

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Amendment No. [

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JAFNPP T * =' ' 3.1-1 (comt'd) 1 RSACTOS PterECTIGE SISTB$ (ECRAM) INSTERSETATIGE REGUIREBENT Minimum Be.

• Modes la teilch Total Number  !

of Operable Trly Level Punction Most be of Instrument i Instrument Trip Function Setting 1 Operable Chamaels Action Chamaels , ,

Provided by (1) per Trly Refuel Startup Rua Design for Roth  !

Bystem (1) (6)' Trip Systems 2 AreN Dowascale ),2.5 Indicated on I 6 Isotrument A or R v

scale (9) ch====1s  !

• I 2 Nigh Reactor [ psig I(8) I I 4 Instrummet A Pressure { Q$

• Ch====1s 2 Nigh Dryuell (2.7 pelg I(7) I(7) I ~ 4 Instrument A Pressure Chamaels 2 Reactor Low Water 3177 la. above TAP I I I 4 Instrument A Level ch====1s 3 Eigh Noter Level 634.5 gallons per 2(2) I I 8 Instrinnent A la Scram Discharge Instrument Volume Chamaels Volume ,

2 Male Steam Line f3a normal full I I I 4 Instruent A Eigh Radiation power background (16) Chamaels i

4 Mala Etess Line f10% valve I(5) 8 Instrimment A Isolation Valve closure -

Chamaele Closure -

4 b bim % p g y)3. W e xu t) B lns4rument C

% Gosse. cbsure. Chemm\3 a-o.d at Ro. 4 23. W g. ar, ,(, p, v'2

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Tans e 3d 1._ICent *3I REACT 9E FatprECTI(M ETEM (BCEastl IBS1M&TIGE REGillREDENT t / I / , /  : _/  : /. _. .

Se DStich Total N O " "" ## .

pusher et Trip Le 1 F lee gemet he

• I Orde le gyerehte Bestrument Ac t i o..

Bestr Trly Fuse es getti

<heeeels ,33 le __

me pre,lded not I startup Per Tr .

<. si h,.eei.e sr== <in nor .eth rly Syst a uv 'Nh '

41.3 < . ..

S .c-- O T.l Cl ore c...or.

tMt3 Lkdb 1. There shall be two ayerable er telyped trip systems for each functies, escept as speelfled w in 4.1.D.

reo= an.t atie.

.. .h.. . ~ . . . . . . . .

e he, en e,or. . in.tr.seet ch.nesi nor a tra,. rote. c...e Py4h == ti- th.s the .iei 8Fote. shall be placed le the safe < tripped) condities, er the appropriate actions listed below shall be taken.

A. Initiate leoerttee of operable rede and cesplete lesettlos et all operehte rods withis four hours.

j E* B***ce yewer level to 3me range and place tende Seltch le the starter Posities within eight hour s. _

c. .o co - r t. no.. t 9 ,orcoat e,r. sed.

m i

[

2. Per.lealble to bypass, if Re me and Shetdeve yesittees of the Beacter Beede Switch. ,

2. .eieto..

4e map:ms 2#

- less then%e. percent et rated.

i

4. DFreesed eY.-eBr eeret __ 7 p::::-:: 5: 5::: t * -- ' ? p:!

S. The deelge perette closure of any two times without a scram helag lettleted.

)

S. 98been the reacter le emberttical as<.1 the reactor watet tempeesture. ls less then 212*r. only the rot taving e n ig.

f unctions *d to be operable: l A. peode switch le shutdowe S. Manual Scram  !

- ~ .. u.. ><. 17 , i11 p 42 i

l

,., , ,.. . .s -

.q -

0 Q -

O s:

JAFMPP

).5 (Cont'dl 4.5 (Cont 'all DELETED C. IIIGH PNESSullE COOLANT INJECTION (HPCI SYSTEMI. C. HICil PHI;SSUf4E COOLANT INJECTION (HPCI SYSTEMI Star vet I lance of HPCI System shall ese per formed as f o l l ow:s provided a reactor steam supply as available. If steam is not available at the tim.= s the nurveillance test is scheduled ~ to tee performed, the test shall be per f ormed within 10 I days of continuous opesation from the time st eam ' .'

become's avai lable.

1. The HPCI system sha!! be operable whenever 1. IIPC I Gystem testing shall be as.spe:ofied in the reactor pressure is greater than 150 4.5.A.I.a, b, c, d, f, and g except that the g psig and reactor coolant temperature is IIPCI pump shall deliver a t_ l ea sQ,250 gpm '

l" greater than 212*F and irradiated fuel is in _

.esainst a system head corresponding o a the reactor vessel, except as speciffad' reactor vessel pressure of +y+@e psig t 150 helow: psig.

[g Amon41 ment Jff. / 7 IIi

l

(,'

• JAFNPP 3.5 (cont'd) 4.5 (cont'd)

The RCIC pump shall deliver at least 400 for a system

, head corresponding to a reactor pressure +;4f9 psig to 150 psig.

4f95

2. When it is determined that the RCIC System is moperable at a time when it is required to be operable, the HPCI System shall be verified to be operable immediately and daily thereafter.

Amendment No.g %

I 121a

t s >

(' .)

Figure 3.5-1 i Thermal Power artd Core Flow Limita of

(\%f Specifications 3.5.J.1, 3.5.J.2 and 3.5.J.3

\/

70

\

Stabili Stability Monitor *.ng i

( APRM and LPRM) Required ,

r Sin 9 1e LO P

~

St ility Monitoring (AP and LPRM) Required ration 60 - Durin Two-Loop Operation, O Line A N 4g , i i

two-Imop '

N' i"

i 50 --

Single- p Operation i u Prohibit j a

W l D I 40 - i m

I 35 = l I

d 30 I i

M I I

m

• e i

a Y n r ng ptRequired i o

20 -

g i o I I

I I

1 10 -

I l

I l

l' 70 s

O .. . . . 50 60 f 30 40 5

/

CORE FLOW (PERCENT RATED) ndmentNo.[,)d,7,f4 I O _.

-_Lrad 5 JAFNPP Figure 3.5-1 Thermal Power and Core Flow Limits of Specificatione 3.5.J.1,3.5.J.2, and 3.5.J.3 70 Stabliity Monitoring Stability Stability Monitoring

( APRM and LPRM) Required M *"""'I"9 During Two* Loop Operation (APRM and (APRM and LPRM) Required ii 6O LPRM) For Single Loop Operation w

Required c

During E 50 ShWe and Two-Loop in Single-Leop Operation Operation / /

'j g Prohibited

/

o .

Line A S 40  :

P2 35 1 *

$ /

2 30

.s

• T E *

• 20 Stability P*onitoring Not Required na -

a '

O 10

.l O

g g g g y g g g g g g---- ---h-- --

g - -. .------g- - - -

- g-30 40 45 50 60 70 CORE FLOW (PERCENT RATED)

.'.mendment No. If,p,%,h,h, 134 '

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

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

i JAFNPP 3 6 (cont'd) 4.6 (cord'd)

7. Reector Vessel Flux Monitoring The reactor vessel Flux Monitoring SurvedBence Program cor12es wilh the intent of the May,1983 th to 10 CFR 50, Appendcas G and H. The next Dux monitoring survelRance capsule sher be removed aner 15 ellective M power year: (EFPYs) and the test procedures and reporting requirements shen meet the requirements of ASTM E 18Mt2. -

I B. Deleted B. Deleted (

C. Coolant Chemistry C. Coolant Chemistry

1. The reactor coolant system radoectivity concentrelon in (}fl. 1. a. A sample of reactor coolant shall be taken at least water shall not ee.ceed the equutrium value of 84 pCl every 96 hr and analyzed for gross gamma activity of dose equevalent 1-131. This limit may be ,

b. Isotopic analysis of a sample of reactor coolant shall following a power transient, for a maximum of 48 hr.

During this lodne activity transient the lodne M M W W mW~

concentrations sher not exceed the equerium Nmits by c. A sample of reactor rannannt sher be taken prior to l more than a factor of 10 whenever the mein steemNne startup and at 4 hr intervals during startup and isolation valves are open. The reactor sher not be snelyzedfor groes gamme actMiy.

operated more then 5 percent of its M power d. During plant steady state operation and losowing an I operation under this exception to the equlubrium limits. If

' '" D h (d h h M M I E)ectors) of 10fND pCl/sec within a 48 hr. period or f more than a factor of 10, the reactor sher be pieced in a a M cold condition within 24 hr. of 20 W of M W !

jg g gg, end analyzed for gross gamme actMty. At least ttree samples wGI be taken at 4 hr intervals. These j

, r-amplig requirements may be omitted whenever i I the equilibnum 1-131 concentration in the ten . !

coolant is less than 0.007 pCi/ml. l t

Amendment No [9 139 4

i

,l  ! ,f ,, -' , ,!!I!

!!  ! l!Ii!jj :; t [ll!i iflj, {:;i l

e y

- no ssesea6 anhchr1 dnr aunnd .,o nr rteohhok eef c rdegee o nc t Tu l

wet nt u gr mdadie at a e e t

ae f

T t nt r i f a ainhhvts a t t beR oi P a -

- ncn eaf si aea mnFt oa f

,r 2epou,pt u F a h_ i s botaz vr e r

y 3 Y. m ue e e y, e r r st lgn s

aonenu t l e p v t cg i l

l eh hT d ms12 usioa r n

o nn t ameAaoarPr r

Pel l tr t

aF f t t nF e

i r eBirhid p

t s .

etsHt pe8xri t il asp .t r6e,f s n e Ee s ioo P,6- i E 1+ emA. mEf ot r f o1 -

t s 1 t 1 -

et st 1 eev a pe s eASesoc 1 c s

sepo r

nprh ee t u1 r s dp .o p3h":

i er

&.1 l t r -

eE o r .o 1 oer g mm mAunAe t

e a sdor ver eh rt of P er napmt ue si so .b

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nypuewr r nii

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Peo hd .i 8 bi a pt 2, T e e deamrg.

er r n 0' t to3F t 1 2dY 0o9 e eT 0 Ear esh1 u .

t 6.r t nP1 7f 1 hr

- e i

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- ade c9 dpoo von eh e 1 eid s /

pait) Y. rut t Frop9 Pgg p

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n nf F i u rn aketahay ee na u2ene t

c r mi n af e i e 9. "t,Aki r

- oeo gbr eapt r eEemB nis l

,t , r p E F ot ad R1 n a ee enlet t s

-. e r F, d NeA e Ger eA a x ,

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D i W E' # -tW*"*

l dht et wnpu Wa( ee 4ee s o i rdeGl u nt o t a wfo Nta Da 6i n

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a x ehcWt Betr r ua e e ceMd as eo s t r t bot nh 6ND1il i

so v r de ' .

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r rr pet r e dfo e ne o v ew mft r

p d = ppAids lWduep h pteuutME c s 6l '""'

l t e 5 on aa e ei 8e -

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MtWd o r ge sic osia nt EaSs ke eoin d

h a efe eu r r iae nded s n w ac tu u 8=

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r u qmT h e

pe R oS gs mnGa s

e .t T o Am e ut s Bna em y 'r l

a mE ,d '

ror a u d e. a4 ohmM c c

t e en s,asm an )a rd .

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9. W d

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g nt sispAl r

i u o p mddA4 l

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s #dg m e nk on wude crE 4id mesa n

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S dab"E t e de del p8 nSe hnWWybe um e nW t

edf a caoGa" yor ys A g, E, r

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tfodnad n md6 es8 d) _

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S1Vndo e da9 pi n h e "0 ot t o r Pnr v e est te oa a nie r eca s ehO9edemar S l p o aw1 4iecRer o t

A Y ri N

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t e

peCmneimheo asn s n r

6 et s"e u e pteaia e ,

pmev0ur ea env i e hei"nhadwal Ce z e thew e

4 Tdr t c e R  ; Ot i r 1SPvh t v m d d n n _

a e _

6 m -

3 A -

4 JAFNPP

' s 3 G and 4.6 BASES (cont'd)

B^ Deleted annunciating at appropriate concentration levels such t C. sampimg for isotopic analysis can be initiated. The design Coolant Chemistry i

details of such a systern nuest be submitted for evaluation accepted by the Commission prior to its inip;e.6o. iation and A radoactivity concentration limit cf 20 4/mi total iodme can incorporation in these Technical Specifications.

be reached if the gaseous effluents are near the limit as set ,

forth in Radmiogeal Effluent Techrucal Specification Section Since the concentration of radioactivity in the ,eacto 3.2.a if there is a failure or a prolonged shutdown of the not continuously measured, coolant sampling would tM cleanup dommeralizer.

ineffective as a means to rapidly detect gross fuel element I "a r~ ;.;.; ;" ; _;~.. Iggg .~. _=x

_2, ..;", . '". ;^.~

failures. However, some capatmlity to detect gross fuel elem faikses is inherent in the radiation monitors in the offg

"'~n ~~'!:t; ' :2. " ; : .=

^

ovo. .a. y m-

^^

6 0 . ..

; 6

T;'.,,

&'2.:, ^* % 2;- system and on the main steam lines. '

' -^_ --t-9 d  :: NE._ 1 1 ~

c' i -

r

. ~. / : 0;nhe rei 7 or"$waterg n: "= ?wsti;"ffg-.

sample be used '

Materials in the Reactor Coolant System are primarily 30

~

stamiess steel and Zircaloy fuel claddog The reactor io assure that the limit of Specificahon 3.6.C is not exceeded. chemistry limits are established to prevent damage to these The total radioactive iodne activity would not be expected to materials. Lirruts are placed on chloride concentration and !

change rapedly over a period of 96 hr. In addebon, the trend of conductivity. The most important limit is that pbced on !

the stack oligas release rate, which is contsnuously monitored, chloride concentration to prevent stress corrosion crackie is a good Indicator of the trend of the iodne activity in the the stainless steel. The attached graph, Fig. 4.6-1, ill reactor coolant. Also during reactor startups and large power the results of tests on stressed 304 stainless steel specimen changes whsch could affect iodme levels, samples of reactor Failures occurred at concentrations above the cunto, no !

coolant shcIl be analyzed to insure iodme concentrahons are failures occurred at concentrations below the curve.

below allowable levels. Analyses is required whenever the 1-131 to the data, allowable chloride concentrahons could be set concentration equilibrium value. is withm a factor of 100 of its allowable several orders of magnitude above the established l The necessity for continued samping oxygen concentration (0.2-0.3 ppm) expenonced during power following power and offgas transsents will be reviewed wettun 2 operation. Zircaloy does not exhibit surular stress corro years ofinitialplant startup. failures.

~

The survesitance requirements 4.6.C.t may be satisfied by a However, there are various conditions under whech the ;

1 ( mtinuous monitoring system capable of determining the total dissolved oxygen content of the reactor coolant water could be sodme concentration in the coolant on a reat time basis, and hegher than 0.2-0.3 ppm. such as refueling, reactor startup, t

hot standby. During these periods with steaming rates less Amendment No  !

I49 l

l

l I

i l

INSERT 6 l In the event of a steam line rupture outside the drywell, a more restrictive coolant activity level of 0.2nci/gm of dase equivalent I 131 was assumed. With this coolant activity level and adverse meteorological conditions, the a.culated radiological dose at the site boundary would be less than 30 rem to the thyroid.

1 j

I i

l 1

1 i

I i

d 4

('

l

, _ _ . _ _ _ _ _ . _ . _ . _ __ __ m. . .. _... ._ . .. . . . _ _

(. V)

JAFMPP 4.7 (cont'd)

(4.) See table 4.7-2 for esceptions.

l (5.) Acceptance criterlom - The combined

• leakage rate for all penetrations and valves subject to type B and C tests  ;

shall be less than 6.60 La. Leakage

, fross containment isolation valves that are sealed with fluid from a seal system may be excluded whom determining the combined leakage rate provided that the installed isolattom valve seal-water system fluid leventory is sufficient to assure the seallag function for at least 30 days.

d. Other leak rate testa (1) The leakage rate for containment isola-tion valves 10-AOV-68A, B (penetration X-13A, 5) for Low Pressure Coolant Injection system and 14-AOV-13A, 3 (penetration X-16A, B) for Core Spray t System shall be less than 11 cubic feet '

per minute per valve (pseumatically tested at 45 psig with ambient temper-ature) or 10 gallons per minute per valve (hydrostatically) tested at M pelg with ambient temperature.

I,019 Amendment No.df , 1 4 172

!!I!

I -

I i .

~

ji }ijlij!ji l l!  !

=

![! __

f

M Afff')r en1 L l yy s.y r< m a s N

g 3.7 BASES (cont'd) d M Using the minimum or medmum downcomer submergence

! levels given in the W containment pressure during %g a 40'F rise (Section 5.2 FSM5 in the suppramminn i

the design basis W is ;---Miri 45 peig which is chamber water temperature and a medmum initial temperature j below the design of ti6 poig. The minimum downcomer of 96'F, a temperature of 14S'F is achieved, which is well below submeroence of st. sin. r=-a in the 170'F temperature which is used irr complete pd, h

• h M; .. .k hMN

- . = n'n.

N A tconden== nan. i l

Bodege tests (9) were run alti a submerged length of 4 ft. W For an initial madrmat suppreselon chamber water I with a mplete condensaNm he, with respect to downcomer tem erature of 95'Fand assuming sie normal complement of "9'" 'h W "'"P " Ad8" "

MPP specinc ansfyses done in connocuan with the M i contamment cooMg pumps (two LPCI pumps and two RHR j

serwce water pumps) containment pressure is not required to r Containment-Suppression Chamber integrity Program indcate mentain Wenot posieve suedon head (HPSH) for the  !

the =i ym of the ap=*werange of sannergence to ensure Lc re spray W and HPCIptsnps.  :

that dynernic forces manaciasmi with pool swell do not result in .s  ;

overstress of the stypression chamber or anww4a8ad bmiting suppresson pool temperature to 199'F dunng RCIC, '

structures. Levelinstrumentation is provided for operator use HPCI, or relief valve operanon, when decay host and stored energy are removed to maintain downoomer submergence within the speedied the primary system by dscharging range. reactor steam direcoy to suppression chamber assures h madmum temperature at the and of blowdown tested adequate CIC, HPCI, or margin for a A,;. n' 'any relief valve operation.

time dunng 4frdM ,

during tio Humboldt Bay (10) and Bodege Bay tests was 170'F, and this is 00f1c"s taken to be the4mveHer- data irwarmena that avemasive steem condensing i desspisteneemfunguBormilho Emit for complete condenansinn loads can be avoided if the peak temperature of the of the reactor coolant, although condensation would occur for suppression pool is maintained below 160'F during any period i temperatures above 170'F. of relief valve operation with sonic conddons at the dscharge I exit. Spec 6cahons have been placed on the envelope of i reactor operating condelons so that the raar+nr can be depressurized in a hmely manner to avoid the regwne of  !

onhalhr high suppression chamber loadings. i Iz m i Amendment No. /,pf, jps , JM  !

188 ,

I t

INSERT 7 Containment analyses predict a 46*F increase in pool water temperature, after complete LOCA blowdown.

These analyses assumed an initial suppression pool water temperature of 95'F and a rated reactor pcwer of 2536 MWt. LOCA analyses in Section 14.6 of the FSAR also assume an initial 95'F pool temperature.

Therefore, complete condensation is assured during a LOCA because the maximum pool temperature (141'F) is less than the 170*F temperature seen during the Bodega Bay tests.

INSERT 8 For an initial maximum suppression chamber water temperature of 95'F, assuming the worst case complement of containment coolir.g pumps (one LPCI pump and two RHR service water pumps), containment pressure is required to maintain adequate net positive suction head (NPSH) for the core spray and LPCI pumps.

INSERT 9 Experiments indicate that unacceptably high dynamic containment loads may result from unstable condensation when suppression pool water temperatures are high near SRV discharges. Action statements limit the j

maximum pool temperature to assure stable condensation. These actions include: limiting the maximum pool '

temperature of 95'F during normal operation; initiating a reactor scram if during a transient (such as a stuck open SRV) pool temperature exceeds 110*F; and depressurizing the reactor if pool temperature exceeds 120*F. I T-quenchers diffuse steam discharged from SRVs and promote stable condensation. The presence of T-quenchers and compliance with these action statements assure that stable condensation will occur and containment loads will be acceptable.

NEDC-24361P (August 1981) summarizes analyses performed to predict pool temperatures and containment loads during plant transients using these temperature limits at a power level of 2535 MWt (104% of rated), i NEDC-24361P also substantiates the acceptability of the plant design using the local poollimits of NUREG-0661. NEDO-30832 (December 1984) shows that SRV condensation loads are low compared to other design loads for plants with T-quenchers. NEDO-30832 describes why local pool temperatures need not be analyzed at a rated power level of 2536 MWt.

4 e

9 W,, - - - - , ,- , , . , ., , , - - , , . , - - -- -

i s ~

L JAFNPP i 4.7 GASES pressure following an accident would  ;

equalize fairly rapidly. Based on i A. Primary Containment the primary containment pressure

' t response and the fact that the m e water in the suppression chamber drywell and suppression chamber is staed only for cooling in the function ac a unit, the primary event of an accident; i.e., it is containment will be tested as a unit not used for normal operations rather than the individual therefore, a daily check of the components separately.

temperature and volume is adequate _

[

to assure that adequate heat removal i~~ . (The design basis loss-of-coolant' capability is present. accident was evaluated in FSAR i Section 14.6 incorporating the i The primary containment preope- primary containment maximum rational test pressures are based allowable accident leak rate of upon the calculated orimary contain- 1.5 percent / day. The analysis  !

ment pressure response corresponding showed that with the leak rate and a .

to the design basis loss-of-coolant standby gas treatment system filter l accident. The peak drywell pressure efficiency of 99 percent for f would be about 45 psig which would halogens, 99 percent for particulate i rapidly reduce to 27 psig within and asstasing the fission product  !

30 sec. following the pipe break. release fractions stated in TID-l Following the pipe break, the 14844, the maximum total whole body suppression chamber pressure rises passing. cloud dose is about .97 rem to 26 psig within 30 sec, equalizes and the maxiansa total thyroid dose t with drywell pressure and thereafter is about 11.4 rem at the site j rapidly decays. with the drywell boundary over an exposure duration i pressure decay (14). of two hours. He resultant thyroid dose that would occur over a 30-day j The design pressure of the drywe!i period is 32.5 rem at the boundary and suppression chamber is of the low population zone (LPZ).

56 psig (15) . The design basis Thus, these doses are the maximum  ;

accident leakage rate is that would be expected in the [

0.5 percent / day at a pressure of unlikely event of a design basis  !

45 peig. As pointed out above, the loss-of-coolant accident. These j drywell and suppression chamber ,paes are also based on the j i

I WW Y D

~ ..

' INSERT 10 Design basis accidents were evaluated as discussed in Section 14,6 of the FSAR and the pow- uprate safety evaluation, Reference 18. The whole body and thyroid doses in the control room, low population zone (LPZ) and site boundary meet the requirements of 10 CFR Parts 50 and 100. The technical support center (TSC), not designed to these licensing bases, was also analyzed. The whole body and thyroid dose acceptance criteria used for the main control room are met for the TSC when initial access to the TSC and occupancy of certain areas in the TSC is restricted by administrative control.The LOCA dose evaluation, Reference 19, assumed: the primary containment leak rate was 1.5 volume percent per day; source term releases were in accordance with TID-14844; and the standby gas treatment system filter efficiency was 99% for halogens.

/

v 9

4 0

0 0

JAFNPP O  :

%- (A) ROUTINE REPORTS (Continued) , l 4

4. CORE OPERATING UMITS REPORT ,

a. Core operatirq limits shall be established prior to startup from each reload  !

cycle, or prior to any remaining portion of a reload cycle for the following:

. The Average Planar Unear Heat Generation Rates (APLHGR) of Speerfication 3.5.H; e The Minimum Critical Power Ratio (MCPR) and MCPR low flow adjustment factor, K,, of Speerfications 3.1.B and 4.1.E; e The Unem Heat Generation Rate (LHGR) of Specification 3.5.l;

. The Reactor Protection System (RPS) APRM flow biased trip settings  ;

of Table 3.1 1; and ,)

. I Tlie flow biased APRM and Rod Block Monitor (RBM) rod block settings of T,able 3.2-3. e -

and shall be documented in the Core Operating Umits Report (COLR).

b. The erdyt.c4 methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC as described in:

1. " General Bectric Standard Application for Reactor Fuel," NEDE.

C 24011 P, latest approved version and amendments.

2. ' James A. FitzPatrick Nuclear Power Plant SAFER /GESTR LOCA reV6 ion; t.oes of Coolant Accident Analysis," NEDC 31317P, October,1986 includnglateysts and addenda.

3. "L.cssef Coolant Accident Analysis for James A. FitzPatrick Nuclear Power Plant," NEDO 21082 2, July,1977 including latest errata and addenda.

1 i c. The core operadng limits shall be determined quch that all applicable limits (e.g., fusi ihermel mechanical limits, core thermal-hydraulic limits, ECCS limus, nucieer umns such as shutdown margin, and transient and accident analysis Imns) of the safety analysis are met.

d. . The COLR, includng any mid cycle revisions or supplements thereto, shall be provided, upon leeuance for each reload cycle, to the NRC Document Conirci Desk with copies to the Regional Administrator and Resident inspector.

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AmendmentNo. If 254c

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JAFWFP

7.0 REFERENCES

(9) C.H. Robbins, " Tests of a Full (1) E. Janssen, " Multi-Rod Burnout at Scale 1/48 Segment of the Hanhlt ,

Bay Pressure Suppression Low Prepsure," ASME Paper 62-HT-26, Containment," GEAP-3596, August 1962. 17, 1960 November (2) K.M. Lacker, " Burnout Conditions (10) " Nuclear for Flow of Boiling Water in Verti- Safety Program Annual I cal Rod Clusters," AE-74 Progress Report for Period Ending (Stockholm, sweden) , May 1962. December 31, 1966, T_ q_;;; I w .;.  ;

L. L.M C. .,  ;- ~ ~.J.- . M ,

(3) FSAR Section 11.2.2. -94647 ORNL-4071.=

(4) FSAR Section 4.4.3. (11) Section 5.2 of the FSAR. '

(12) TID 20583,

• Leakage Characteristics (5) .I.M. Jacobs, " Reliability of Engi-neered Safety Features as a Func- of Steel Containment Vessel and the

• tion of Testing Frequency," Nuclear Analysis of Leakage Rate Determi-i nations."

Safety, Vol. 9, No. 4, July-August i 1968, pp 310-312. t (13) Technical Safety Guide, " Reactor Containment Leakage Testing and (6) Benjamin Epstein, Albert Shiff, I

UCRL-50451, Improving Availability Surveillance Requirements," USAEC, Division of Safety Standards, 4

and Readiness of Field Equiseent t Revised Draft, December 15, 1966.

Through 16, Periodic Inspection, July .

1968, p. 10, Equation (24) ,

Lawrence hadiation Laboratory. (14) Section 14.6 of the FSAR.

(7) I.M. Jacobs and P.W. Mariott, APED (15) ASME Boiler and Pressure Vessel  ;

Code, Nuclear Vessels,Section III.

Guidelines for Determining Safe Maximum allowable internal pressure Test Intervals and Repair Times for is 62 psig.

Engineered Safeguards - April 1969.

(B) Bodega Bay Preliminary Hazards Re- (16) 10CFR50.54, Appendix J Reactor Con-port, Appendix 1, Docket 50-205, tainment Testing Requirements.*

• l December 28, 1962. (17) 10CFR50, Appendix J, February 13, i 1973. ,

i N t b

r Ace A 4 %. '"

1 l

INSERT 11 (18) General Electric Report NEDC.32016P,

• Power Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant,' December 1991 (proprietary).

(19) James A. FitzPatrick Calculation JAF CALC-RAD-00008,' Radiological Consequences of Design Basis Accidents at James A. FitzPatrick,' November 1991.

(20) General Electric Report GE.NE.187 451191P, ' Containment Systems Evaluation,' (proprietary).

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