ML20134M175

Proposed Tech Specs Reflecting Interposed Amend That Was Issued,Updating References to Repts on TS Pp & Changing Rev Bars on Previously Submitted Update Pp Which Were Erroneously Positioned on Pp

ML20134M175
Person / Time
Site:	FitzPatrick
Issue date:	11/20/1996
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20134M167	List: JPN-96-046, Forwards Addl Info Re 920612 Application for Amend to License DPR-59 Re Power Uprate,To Reflect Interposed Amend Issued,Update Ref to Rept on TS Pages & Change Rev Bars on Previous Update Pages ML20134M175
References
NUDOCS 9611220256
Download: ML20134M175 (42)

Text

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l Attachment I to JPN-96-046 i

' 1 UPDATED PAGE CHANGES FOR PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS REGARDING POWER UPRATE (JPTS 91-025) 4 i

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 59 9611220256 961120 PDR ADOCK 05000333 P PDR

JAFNPP AD. Core Operatina Limits Report (COLR)

Z. Too of Active Fuel i This report is the plant-specific document that The Top of Active Fuel, corresponding to the top of provides the core operating limits for the current  ;

the enriched fuel column of each fuel bundle, is operating cycle. These cycle-specific operating  !

located 352.5 inches above vessel zero, which is limits shall be determined for each reload cycle in  ;

the lowest point in the inside bottom of the reactor accordance with Specification 6.9.A.4. Plant vessel. (See General Electric drawing No. operation within these operating limits is addressed j 919D690BD.) in individual Technical Specifications.

l AA. Rod Density AE. References i Rod density is the number of control rod notches 1. General Electric' Report NEDC-32016P-1, inserted expressed as a fraction of the total number " Power Uprate Safety Analysis for James A.

of control rod notches. All rods fully inserted is a FitzPatrick Nuclear Power Plant," April 1993 condition representing 100 percent rod density. (proprietary), including Errata and Addenda Sheet No.1, dated January 1994.

AB. Puroe-Puraina -

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

AC. Ventino Venting is the controlled process of releasing air or j gas from a confinement in such a manner that i replacement air or gas is not provided or required. t 1

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5 Amendment No. 75, 93,1S2, ,

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JAFNPP 1.1 BASES (Cont'd) .

E. Refecences C. Power Transient

1. " General Electric Standard Application for >

Plant safety analyses have shown that the scrams Reactor Fuel," NEDE-24011-P-A-13, August caused by exceeding any safety system setting will 1996.

assure that the Safety Limit of 1.1.A or 1.1.B will not be exceeded. Scram times are checked periodically to 2. FitzPatrick Nuclear Power Plant Single-Loop assure the insertion times are adequate. The thermal Operation, NEDO 24281, August 1980.  !'

power transient resulting when a scram is accomplished other than by the expected scram signal 3. GE12 Compliance with Amendment 22 of  :

(e.g., scram from neutron flux following closure of the NEDE-24011-P-A (GESTAR 11), NEDE-32417P, main turbine stop valves) does not necessarily cause December 1994.  ;

fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the t plant design. The concept of not approaching a Safety Limit provided scram signals are operable is  ;

supported by the extensive plant safety analysis. ,

Reactor Water Level (Hot or Cold Shutdown i D.

Condition) t During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the '

active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capatslity could lead to elevated cladding temperatures and clad i perforation. The core will be cooled sufficiently to i prevent clad melting should the water level be  !

reduced to two-thirds the core height. Establishment i of the Safety Limit at 18 in. above the top of the fuel ,

provides adequate margin. This level will be k.  :

continuously monitored whenever the recirculation pumps are not operating.

Amendment No. M, 90,1 S2, 238, 14

___. .. _ __. _________..m_ _ . _. _. . _ _ . _ _ _ _ . . . . _ _ _ _ _ _ _ _ _. .._____________._._.___m__ _ - _ _ . _ _ .

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

JAFNPP 2.1 BASES (Cont'd)

B. Not Used C. References

1. General Electric Report, NEDC-32016P-1, " Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant," April 1993 (proprietary), including Errata and i Addenda Sheet No.1, dated January 1994.

2. " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A-13, August 1996.  ;

3. (Deleted)

4. FitzPatrick Nuclear Power Plant Single-Loop Opeistion, ,

NEDO-24281, August,1980.

Amendment No.

• i,19, Si, 98,162,190, 20 (Next page is 23) ,

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

i JAFNPP  ;

1.2 and 2.2 BASES i The reactor coolant pressure bo Jndary integrity is an important The limiting vessel overpressure transient event is a main steam ;l' barrier in the prevention of uncontrolled release of fission products. isolation valve closure with flux scram. This event was analyzed It is essential that the integrity of this boundary be protected by within NEDC-32016P-1, " Power Uprate Safety Analysis For James i crtablishing a pressure limit to be observed for all operating A. FitzPatrick Nuclear Power Plant," including Errata and Addenda j conditions and whenever there is irradiated fuel in the reactor Sheet N9.' 1, dated January 1994, assuming 9 of the 11 SRVs were j vessel. operable wit h opening pressures less than or equal to 1179 psig.

The resultant peak vessel pressure for the event was shown to be The pressure safety limit of 1,325 psig as measured by the vessel less than the AGME Code limit of 1375 psig (see current reload  ;

steam space pressu c indicator is equivalent to 1,375 psig at the analysis for the reactor response to the main steam isolation valve i lowest elevation of the Beactor Coolant System. The 1,375 psig closure with flux scram event). ,

v11ue is derived from tbc design pressures of the reactor pressure '

vessel and reactor coolant system piping. The respective design A safety limit is applied to the Residual Heat Removal System l pressures are-1250 psig at 575 *F for the reactor vessel,1148 psig (RHRS) when it is operating in the shutdown cooling mode. When l at 568 'F for the recirculation suction piping and 1274 psig at 575 operating in the shutdown cooling mode, the RHRS is included in

• F for the discharge piping. The pressure safety limit was chosen the reactor coolant system.

Es the lower of the pressure transients permitted by the applicable f d sign codes: 1965 ASME Boiler and Pressure Vessel Code, l Section 111 for pressure vessel and 1969 ANSI B31.1 Code for the  !

rzactor coolant system piping. The ASME Boiler and Pressure  ;

Vassel Code permits pressure transients up to 10 percent over d: sign pressure (110% x 1,250 = 1,375 psig) and the ANSI Code permits pressure transients up to 20 percent over the design ,

t pressure (120% x 1,150 = 1,380 psig). The safety limit pressure j of 1,375 psig is referenced to the lowest elevation of the Reactor t Coolant System. l l

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Amendment No. 58,01,131,190,217,  !

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._ - - _. _-. -_ h_ _ _ _ . - - _ _ _ _ _ _ _ __ _ - _ _ _ _ _ _ _ - .__-___________-____?

t JAFNPP 3.1 BASES (cont'd) l Turbine control valves fast closures initiates a scram based C. References on pressure switches sensing electro-hydraulic control (EHC) system oil pressure. The switches are located between fast 1. " James A. FitzPatrick Nuclear Power Plant  !

closure solenoids and the disc dump valves, and are set SAFER /GESTR-LOCA Loss-of-Coolant Accident

• relative (500 < P< 850 psig) to the normal (EHC) oil pressure Analysis," NEDC-31317P, Revision 2, April 1993. l of 1,600 psig so that based on the small system volume, they can rapidly detect valve closure or loss of hydraulic ,

pressure.  ;

i The requirement that the IRM's be inserted in the core when the APRM's read 2.5 indicated on the scale in the start-up and refuel modes assures that there is propar overlap in the neutron monitoring system functions and thi!s, that  :

adaquate coverage is provided for all ranges of reactor i operation.

B. The limiting transient which determines the required steady ,

state MCPR limit depends on cycle exposure. The operating t limit MCPR values as determined from the transient analysis in the current reload submittal for various core exposures are specified in the Core Operating Limits Report (COLR).  !

The ECCS performance analyses assumed reactor operation will be limited to the MCPR value for each fuel type as ,

I described in Reference 1. The Technical Specifications limit operation of the reactor to the more conservative MCPR  !

based on consideration of the limiting transient as specified in the COLR. l; l

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l Amendment No. 19, Si,109,1S2, 9:i:d 5; '!9C L...: dnd 2/1 S/93, 35 ,

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

5 L JAFNPP 3.3 and 4.3 BASES (cont'd)

" full out" position during the performance of SR 4.3.A.2.a. 3. The Rod Worth Minimizer (RWM) restricts the order of This Frequency is acceptable, considering the low control rod withdrawal and insertion to be equivalent to the Banked Position Withdrawal Sequence (BPWS). These probability that a control rod will become uncoupled when it is not being moved, and operating experience related to sequences are established such that the drop. of any .

uncoupling events. in-sequence control rod from the fully inserted position to i the position of the control rod drive would not cause the  !

reactor to sustain a power excursion resulting in a peak fuel

2. The control rod housing support restricts the outward enthalpy in excess of 280 cal /gm. An enthalpy of 280  :

movement of a control rod to less than 3 in. in the cal /gm is well below the level at which rapid fuel dispersal extremely remote event of a housing failure. The amount could occur (i.e. 425 cal /gm.). Primary system damage in of reactivity which could be added by this small amount of this accident is not possible unless a significant amount of ,

rod withdrawal, which is less than a normal single fuel is rapidly dispersed. Ref. Subsections 3.6.6, 7.7.4.3 '

withdrawal increment, will not contribute to any damage to and 14.6.1.2 of the FSAR, NEDE-24011-P-A-13, August the Primary Coolant System. The design basis is given in 1996 and NEDO-10527 including supplements 1 and 2 to i subsection 3.8.2 of the FSAR, and the safety evaluation is NEDO-10527.

, given in subsection 3.8.4. This support is not required if .

the Reactor Coolant System is at atmospheric pressure in performing the function described above, the RWM is not  !

since there would then be no driving force to rapidly eject required to impose any restrictions at core power levels in i a drive housing. Additionally, the support is not required if excess of 10% of rated. Material in the cited references all control rods are fully inserted and if an adequate shows that it is impossible to reach 280 calories per gram shutdown margin with one control rod withdrawn has been in the event of a control rod drop occurring at power  !

demonstrated, since the reactor would remain subcritical greater than 10% regardless of the rod pattem. This is '

even in the event of complete ejection of the strongest true for all normal and abnormal patterns including those control rod. which maximize the individual control rod worth.  ;

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Amendment No. 30,155,193, j 100

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JAFNPP 1 3.3 and 4.3 BASES (cont'd) t

5. The Rod Block Monitor (RBM) is designed to automatically C. Scram insertion Times prevent fuel damage in the event of erroneous rod withdrawal ,

from locations of high power density during high power level The Control Rod System is designated to bring the reactor operation. Two channels are provided, and one of these may be subcritical at a rate fast enough to prevent fuel damage;i.e.,

bypassed from the console for maintenance and/or testing. to prevent the MCPR from becoming less than the Safety i Tripping of one of the channels will block erroneous rod Limit. Scram insertion time test criteria of Section 3.3.C.1 i withdrawal soon enough to prevent fuel damage. were used to generate the generic scram reactivity curve

. shown in NEDE-24011-P-A-13, August 1996. This generic l This system backs up the operator who withdraws control rods curve was used in analysis of non-pressurization transients according to written sequences. The specified restrictions with to determine MCPR limits. Therefore, the required protection 'j one channel out of service conservatively assure that fuel is provided.

damage will not occur due to rod withdrawal errors when this condition exists.

The numerical values assigned to the specified scram A limiting control rod pattern is a pattern which results in the performance are based on the analysis of data from other i BWR's with control rod drives the same as those on core being on a thermal hydraulic limit (e.g., MCPR limit). During use of such patterns,it is judged that testing of the RBM System JAFNPP. t prior to withdrawal of such rods to assure its operability will ,

assure that improper withdraw does not occur. It is the The occurrence of scram times within the limits, but r responsibility of the Reactor Engineer to identify these limiting significantly longer than the average, should be viewed as an patterns and the designated rods either when the patterns are indication of a systematic problem with control rod drives, <

initially established or as they develop due to the occurrence of especially if the number of drives exhibiting such scram inoperable control rods in other than limiting patterns. times exceeds eight, the allowable number of inoperable  !

rods.

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i Amendment No. 'i, 18,21,30,13,19,53,S7,155,1S2, l 102

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T JAFNPP  !

3.5 BASES i A. Core Sorav System and Low Pressure Coolant iniection (LPCI) Core spray distribution has been shown, in full scale tests of i i Mode of the RHR System systems similar in design to that of the FitzPatrick Plant, to exceed the minimum requirements by at least 25 percent. In addition,  !

- This specification assures that adequate emergency cooling - cooling effectiveness has been demonstrated at less than half the capability is available whenever irradiated fuel is in the reactor rated flow in simulated fuel assemblies with hester rods to  !

i vessel. duplicate the decay heat characteristics of irradiated fuel. The  ;

accident analysis is additionally conservative in that no credit is '

The loss-of-coolant analysis is referenced and described in taken for spray coolant entering the reactor before the intemal General Electric Topical Report NEDE-24011-P-A-13, August pressure has fallen to 113 psi above primary containment  ;

1996. pressure.

The limiting conditions of operation in Specifications 3.5.A.1 The LPCI mode of the RHR System is designed to provide .

~t hrough 3.5.A.6 specify the combinations of operable emergency cooling to the core by flooding'in the event of a i subsystems to asstra the availability of the minimum cooling loss-of-coolant accident. These subsystems are completely systems. No single failure of ECCS equipment occurring during independent of the Core Spray System; however, they function in a loss-of-coolant accident under these limiting conditions of combination with the Core Spray System to prevent excessive fuel operation will result in inadequate cooling of the reactor core. clad temperature. The LPCI mode of J

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Amendment No. 18,

• 18,

• iS, 125 .

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JAFNPP i

3.6 and 4.6 BASES (cont'd)

E. Safetv/ Relief Valves The safety / relief valves (SRVs) have two modes of operation; the with the HPCI and RCIC turbine overspeed systems and the Mark safety mode or the relief mode. In the safety mode (or spring i torus loading analyses. Based on safety / relief valve testing mode of operation) the spring loaded pilot valve opens when the experience and the analysis referenced above, the safety / relief steam pressure at the valve inlet overcomes the spring force valves are bench tested to demonstrate that in-service opening holding the pilot valve closed. The safety mode of operation is pressures are within the nominal pressure setpoints 13% and '

required during pressurization transients to ensure vessel then the valves are returned to service with opening pressures at pressures do not exceed the reactor coolant pressure safety limit the nominal setpoints 11%. In this manner, valve integrity is of 1,375 psig. maintained from cycle to cycle. t in the relief mode the spring loaded pilot valve opens when the The analyses- with NEDC-32016P-1, including Errata and spring force is overcome by nitrogen pressure which is provided Addenda Sheet No.1, dated January 1994, also provide the to the valve through a solenoid operated valve. The solenoid safety basis for which 2 SRVs are permitted inoperable during operated valve is actuated by the ADS logic system (for those continuous power operation. With more than 2 SRVs inoperable,  !

SRVs which are included in the ADS) or manually by the operator the margin to the reactor vessel pressure safety- limit is from a control switch in the main control room or at the remote significantly reduced, therefore, the plant must enter a cold -

ADS panel. Operation of the SRVs in the relief mode for the ADS condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> once more than 2 SRVs are is discussed in the Bases for Specification 3.5.D. determined to be inoperable. (See reload evaluation for the current cycle).

Experiences in safety / relief valve testing have shown that failure or deterioration of safety / relief valves can be adequately detected A manual actuation of each SRV is performed to demonstrate if at least 5 of the 11 valves are bench tested once every 24 that the valves are mechanically functional and that no blockage months so that all valves are tested every 48 months. exists in the valve discharge line. Valve opening is confirmed by i Furthermore, safety / relief valve testing experience has monitoring the response of the turbine bypass valves and the demonstrated that safety / relief valves which actuate within13% SRV acoustic monitors. Adequate reactor steam dome pressure  !

of the design pressure setpoint are considered operable (see must be available to avoid damaging the valve. Adequate steam ANSl/ASME OM-1-1981). The safety bases for a single nominal flow is required to ensure that reactor pressure can be I valve opening pressure of 1145 psig are described in maintained during the test. Testing is performed in the RUN NEDC-32016P-1, " Power Uprate Safety Analysis for James A. mode to reduce the risk of a reactor scram in response to small t FitzPatrick Nuclear Power Plant," including Errata and Addenda pressure fluctuations .thich may occur while opening and l '

Sheet No.1, dated January 1994. The single nominal setpoint is reciosing the valves. .

set below the reactor vessel design pressure (1250 psig) per the requirements of Article 9 of the ASME Code - Section ill, Nuclear Low power physics testing and reactor operator training with r I Vessels. The setting of 1145 psig preserves the safety margins inoperable components will be conducted only when the  !

associated safety / relief valves are l

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Amendment No. 13,134,217,219,229, ,

152 .

t JAFNPP 3.7 (cont'd) 4.7 (cont'd)

(2) During testing which adds heat to the suppression pool, the water temperature shall not exceed 10*F above the normal power operation limit specified in (1) above. In connection with such testing, the pool temperature must ,

be reduced to below the normal power operation limit specified in (1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(3) The reactor shall be scrammed from any operating condition if the pool temperature reaches 110*F. Power i operation shall not be resumed until the pool temperature is reduced below the normal power operation limit .

specified in (1) above.

(4) During reactor isolation conditions, the reactor pressure -

vessel shall be depressurized to less than 200 psig at normal cooldown rates if the pool temperature reaches 120*F.

2. Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water 2. a. Perform required visual examination and leakage rate '

temperature is above 212*F, and fuel is in the reactor vessel, testing of the Primary Containment in accordance with '

except while performing low power physics tests at the Primary Containment Leakage Rate Testing Program.

atmospheric pressure at power levels not to exceed 5 MWt. i

b. Demonstrate leakage rate through each MSIV is s 11.5 ,

scfh when tested at a 25 psig. The testing frequency is t in accordance with the Primary Containment Leakage ,

Rate Testing Program. t

c. Once per 24 months, demonstrate the leakage rate of i 10AOV-68A,B for the Low Pressure Coolant injection

. system and 14AOV-13A,B for the Core Spray system to be less than 11 scfm per valve when pneumatically i tested at a 45 psig at ambient temperature, or less than i 10 gpm per valve if hydrostatically tested at a 1,035 l '

psig at ambient temperature. ,

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Amendment No. 16, 231, 166  ;

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JAFNPP 3.7 BASES (cont'd)

I Using the minimum or maximum torus water level (which are temperature. Therefore, complete condensation is assured based on downcomer submergence levels where 13.88 feet during a LOCA because the maximum pool temperature above the bottom of the torus is 0.005 feet higher than the (141 *F) is less than the 170*F temperature seen during the minimum submergence of 51.5 inches and 14.00 feet above Bodega Bay tests.

the bottom of the torus is equivalent to the maximum

• i submergence of 53 inches assumed in containment analyses) For an initial maximum torus water temperature of 95*F, containment pressure during the design basis accident is assuming the worst case complement of containment cooling i approximately 45 psig which is below the design of 56 psig. pumps (one LPCl pump and two RHR service water pumps), i The minimum downcomer submergence of 51.5 inches results containment pressure is required to maintain adequate net  !

I in a minimum torus water volume of approximately 105,900 positive suction head (NPSH) for the core spray and LPCI i feet . The majority of the Bodega tests (9) were run with a pumps, i submerged length of 4 feet and with complete condensation. .

Thus, with respect to downcomer submergence, this Limiting suppression pool temperature to 105*F during RCIC, I t specification is adequate. Additional JAFNPP specific analyses HPCI, or relief valve operation, when decay heat and stored  ;

done in connection with the Mark l Containment-Suppression energy are removed from the primary system by discharging i i Chamber integrity Program indicate the adequacy of the reactor steam directly to the torus assures adequate margin for - f specified range of submergence to ensure that dynamic forces a potential blowdown any time during RCIC, HPCI, or relief associated with pool swell do not result in overstress of the valve operation. ,

torus or associated structures. Levelinstrumentation is i provided for operator use to maintain downcomer Experiments indicate that unacceptably high dynamic I submergence within the specified range. containment loads may result from unstable condensation  ;

when suppression pool water temperatures are high near SRV t The maximum temperature at the end of blowdown tested discharges. Action statements limit the maximum pool l during the Humboldt Bay (10) and Bodega Bay tests was temperature to assure stable condensation. These actions  !

170 F, and this is conservatively taken to be the limit for include: limiting the maximum pool temperature of 95*F

• I complete condensation of the reactor coolant, although during normal operation; initiating a reactor scram if during a  !

condensation would occur for temperatures above 170 F. transient (such as a stuck open SRV) pool temperature [

exceeds 110 F; and depressurizing the reactor if pool Containment analyses predict a 46*F increase in pool water temperature exceeds 120*F. T-quenchers diffuse steam t temperature, after complete LOCA blowdown. These analyses discharged from SRVs and promote stable condensation. The i assumed an initial suppression pool water temperature of 95 F presence of T-quenchers and compliance with these action l

, and a rated reactor power of 2536 MWt. LOCA analyses in statements assure that stable condensation will occur and ,

Section 14.6 of the FSAR also assume an initial 95*F pool containment loads will be acceptable. j

' Amendment No. 1 S, 3S,1 se,13 3, 3 97, 188  ;

JAFNPP 4.7 BASES  !

I A. Primary Containment Design basis accidents were evaluated as discussed in  ;

Section 14.6 of the FSAR and the power uprate safety .;

The water in the suppression chamber is used only for evaluation, Reference 18. The whole body and thyroid

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I cooling in the event of an accident; i.e., it is not used for doses in the control room, low population zone (LPZ) and normal operation; therefore, a daily check of the site boundary meet the requirements of 10 CFR Parts 50 temperature and volume is adequate to assure that and 100. The technical support center (TSC), not .

adequate heat removal capability is present. designed to these licensing bases, was also analyzed. The  !

whole body and thyroid dose acceptance criteria used for The primary containment preoperational test pressures are the main control room are met for the TSC when initial  !

based upon the calculated primary containment pressure access to the TSC and occupancy of certain areas in the response corresponding to the design basis loss-of-coolant TSC is restricted by administrative control. The LOCA accident. The peak drywell pressure would be about 45 dose evaluations, References 19, 20, and 21, assumed:

psig which would rapidly reduce to 27 psig within 30 sec. the primary containment leak rate was 1.5 volume percent following the pipe break. Following the pipe break, the per day; source term releases were in accordance with suppression chamber pressure rises to 26 psig within 30 TID-14844 and Regulatory Guide 1.3, and were consistent -

r sec, equalizes with drywell pressure and thereafter rapidly with the Standard Review Plan; and the standby gas .

decays with the drywell pressure decay (14). treatment system filter efficiency was 99% for halogens.  !

These doses are also based on the  !

The design pressure of the drywell and suppression  !

chamber is 56 psig(15). The design basis accident t leakage rate is 0.5 percent / day at a pressure of 45 psig. I As pointed out above, the drywell and suppression ,

chamber pressure following an accident would equalize '

j fairly rapidly.- Based on the primary containment pressure t response and the fact that the drywell and suppression chamber function as a unit, the primary containment will '

be tested as a unit rather than the individual components separately, i

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i Amendment No.

193

s JAFNPP

7.0 REFERENCES

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

62-HT-26, August 1962.

(12) TID 20583, " Leakage Characteristics of Steel Containment (2) K.M. Backer, " Burnout Conditions for Flow of Boiling Water in Vessel and the Analysis of Leakage Rate Determinations."

Vertical Rod Clusters," AE-74 (Stockholm, Sweden), May l 1962. (13) Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program", dated September 1995.  !

(3) FSAR Section 11.2.2.

(14) Section 14.6 of the FSAR.

(4) FSAR Section 4.4.3.

(15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, *

(5) 1.M. Jacobs, " Reliability of Engineered Safety Features as a Section 111. Maximum allowable internal pressure is G2 psig.

Function of Testing Frequency," Nuclear Safety, Vol. 9, No. 4, July-August 1968, pp 310-312. (16) 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B -

(6) Deleted Performance Based Requirements", Effective Date October 26, 1995 (7) 1.M. Jacobs and P.W. Mariott, APED Guidelines for Determining Safe Test Intervals and Repair Times for (17) Deleted .

Enginected Safeguards - April 1969. ,

(18) General Electric Report NEDC-32016P-1, " Power Uprate >

(8) Bodega Bay Preliminary Hazards Report, Appendix 1, Docket Safety Analysis for James A. FitzPatrick Nuclear Power Plant,"

50-205, December 28,1962. April 1993 (proprietary), including Errata and Addenda Sheet No.1, dated January 1994.

(9) C.H. Robbins, " Tests of a Full Scale 1/48 Segment of the .

Humbolt Bay Pressure Suppression Containment," (19) James A. FitzPatrick Calculation JAF-CALC-RAD-OOO23, Rev.  !

GEAP-3596, November 17,1960. O, " Power Uprate Program - Technical Support Center Post- i-Accident Radiological Habitability Study," August 1996.

(10) " Nuclear Safety Program Annual Progress Report for Period Ending December 31,1966, ORNL-4071." (20) James A. FitzPatrick Calculation JAF-CALC-RAD-OOO42, Rev. L 0, " Control Room Radiological Habitability Under Power Uprate .

Conditions and CREVASS Reconfiguration," September 1995. l

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JAFNPP

7.0 REFERENCES

(continued)

(21) James A. FitzPatrick Calculation JAF-CALC-RAD-OOO48, Rev. O, " Power Uprate Project - Radiological impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents," May 1996 (22) General Electric Report GE-NE-187-45-1191,

" Containment Systems Evaluation for the James A.

FitzPatrick Nuclear Power Plant," November 1991 (proprietary).

Amendment No.

285a

I i Att chmInt il 13 JPN-96-046 l EVALUATION OF UPDATED PAGE CHANGES FOR PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS REGARDING POWER UPRATE (JPTS-91-025) i I

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New York Power Authority ~

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

l Attichm:nt 11 to JPN-96-046 EVALUATION OF UPDATED PAGE CHANGES Page 1 of 8

1. DESCRIPTION OF THE PROPOSED CHANGES This section provides a description of the proposed changes to the Technical Specifications (TS). Minor changes in format, such as type font, margins or .

hyphenation, are not described in this submittal. These changes are typographical l and do not affect the content of the TS. )

Technical Soecifications Pace 6a. Definitions Section 1.0.AE.1 This change adds the current revision of the Power Uprate Safety Analysis Report (NEDC-32016P-1, Reference 1). This change supersedes the change originally proposed in JPN-92-028 (Reference 2):

l Add the following: l "AE. References l

1. General Electric Report NEDC-32016P-1, ' Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant,' April 1993 i (proprietary), including Errata and Addenda Sheet No.1, dated l January 1994" l Technical Snecifications Paae 14. Bases Section 1.1.E.1 l This change adds the current revision of the General Electric Standard Application  !

for Reactor Fuel. This change is necessary to reflect the change proposed in JPN-96-043 (Reference 3) to TS Page 254c, Section 6.9.(A).4.b.1.

Replace the following:

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

with:

"1. ' General Electric Standard Application for Reactor Fuel,' NEDE-24011-P-A-13, August 1996" Technical Soecifications Paae 20. Bases Section 2.1.C.1 This change adds the current revision of the Power Uprate Safety Analysis Report (NEDC-3201tiP-1, Reference 1). The change supersedes the change originally proposed in JPN-92-028 (Reference 2) and in JAFP-96-0306 (Reference 4).

Replace:

. (Deleted)"

with:

"1. General Electric Report, NEDC-32016P-1, ' Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant,' April 1993 (proprietary),

including Errata and Addenda Sheet No.1, dated January 1994."

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

a

, AttachmInt ll to JPN 96-046 EVALUATION OF UPDATED PAEE CHANGES Page 2 of 8 Technical Soecifications Paae 20. Bases Section 2.1.C.2 This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change is necessary to reflect the change proposed in JPN-96-043 (Reference 3) to TS Page 254c, Section 6.9.(A).4.b.1.

Replace the following:

"2. ' General Electric Standard Application for Reactor Fuel', NEDE 24011-P-A (Approved revision number applicable at time that reload fuel analyses are performed)."

with:

"2. ' General Electric Standard Application for Reactor Fuel,' NEDE-24011-P-A-13, August 1996."

Technical Snecifications Paae 29. Bases Section 1.2 and 2.2 i l

This change adds the current revision of the Power Uprate Safety Analysis Report I (NEDC-32016P-1, Reference 1), l l

Replace the following change originally proposed in JPN-92-028 (Reference 2) and I superseded by the change proposed in JAFP-96-0306 (Reference 4). j

"...NEDC-31697P, ' Updated SRV Performance Requirements for the JAFNPP,' )

assuming 9 of the 11 SRVs were operable with opening pressures less than or '

equal to 1195 psig. The resultant peak vessel pressure for the event was shown to be less than the vessel pressure code limit of 1375 psig. (See current reload

. analysis for the reactor response to the main steam isolation valve closure with flux scram event). The value of 1195 psig is the SRV opening pressure up to which plant performance has been analyzed, assuming 2 SRVs are inoperable. Therefore, SRV opening pressures below 1195 psig ensure that the ASME Code limit on peak reactor pressure is satisfied..."

with:

"...NEDC-32016P-1, ' Power Uprate Safety Analysis For James A. FitzPatrick Nuclear Power Plant,' including Errata and Addenda Sheet No.1, dated January  ;

1994, assuming 9 of the 11 SRVs were operable with opening pressures less than '

or equal to 1179 psig. The resultant peak vessel pressure for the event was shown to be less than the ASME Code limit of 1375 psig (see current reload analysis for the reactor response to the main steam isolation valve closure with flux scram event)..."

Attachmtnt 11 to Jr 1-96-046 I

EVALUATION OF UPDATEO PAGE CHANGES I Page 3 of 8 Technical Soecifications Paae 35. Bases Sections 3.1.B. 3.1.C.1. and 3.1.C.2 This change deletes reference to an outdated Loss of Coolant Accident Analysis Report for FitzPatrick (NEDO 21662, Reference 5). In addition, this change provides the current reference for the Loss of Coolant Accident Analysis Report applicable to FitzPatrick (NEDC-31317P, Reference 6). These changes are necessary to reflect the changes proposed in JPN-96-043 (Reference 3) to TS Page 254c, Section 6.9.(A).4.b. JAFP-96-0306 (Reference 4) stated that no change was required to TS Page 35. However, for the reasons stated above, a change is required. j

1. Replace the following in Bases Section 3.1.B:

l

...NEDO 21662 (Reference 1) and NEDC 31317P (Reference 2) including the latest revision, errata and addenda..."

with:

... Reference 1..."

2. Replace the following from Bases Section 3.1.C:

"1. General Electric Topical Report NEDO-21662, Revision 2, ' Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear i Power Plant (Lead Plant)', July 1977 with errata and addenda. I

2. General Electric Topical Report NEDC-31317P, ' James A. FitzPatrick Nuclear Power P! ant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis', October 1986 with revisions, errata and addenda."

with:

"1. ' James A. FitzPatrick Nuclear Power Plant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,' NEDC-31317P, Revision 2, April 1993."

Technical Soecifications Paae 100. Section 3.3 and 4.3 of the Bases item B.3 This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change is necessary to reflect the change proposed in JPN-96-043 (Reference 3) to TS Page 254c, Section 6.9.(A).4.b.1.

Replace the following:

... N E D E-2401 1.. . "

with:

...NEDE-24011-P-A-13, August 1996..."

l Att:chmint il to JPN-96-046 EVALUATION OF UPDATED PAGE CHANGES

Page 4 of 8

Technical Soecifications Paae 102. Section 3.3 and 4.3 of the Bases item C.

1 j This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change is necessary to reflect the change proposed in JPN-

96-043 (Reference 3) to TS Page 254c, Section 6.9.(A).4.b.1.

Replace the following:

i

". ..N E DE-2401 1 -P-A..."

T l with:  !

! "...NEDE-24011 P-A-13, August 1996..."

1

! Technical Snecifications Paae 125. Bases Section 3.5.A. l This change adds the current revision of the General Electric Standard Application

for Reactor Fuel. This change is necessary to reflect the change proposed in JPN- l 96-043 (Reference 3) to TS Page 254c, Section 6.9.(A).4.b.1. '

j: Replace the following:

l .. . N E D E-2401 1 -P-A. . . "

with:

"...NEDE-24011 P-A-13, August 1996..."

Technical Soecifications Paae 152. Section 3.6 and 4.6 of the Bases item E.

This change adds the current revision of the Power Uprate Safety Analysis Report (NEDC-32016P-1, Reference 1) and adds a revision bar which was omitted in JAFP-96-0306. This change supersedes the change originally proposed in JPN-92-028 (Reference 2) and in JAFP-96-0306 (Reference 4).

Replace:

"...The safety bases for a single nominal valve opening pressure of 1110 psig are described in NEDC-31697P, ' Updated SRV Performance Requirements for the JAFNPP.' The single nominal setpoiric is set below the reactor vessel design pressure (1250 psig) per the requirements of Article 9 of the ASME Code - Section lil, Nuclear Vessels. The setting of 1110 psig..."

and:

...The analyses with NEDC-31697P also..."

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

Att: chm:nt il to JPN-96-046 EVALUATION OF UPDATED PAGE CHANGES

, Page 5 of 8 j with "The safety bases for a single nominal valve opening pressure of 1145 psig are  ;

described in NEDC-32016P-1, ' Power Uprate Safety Analysis for James A.

FitzPatrick Nuclear Power Plant,' including Errata and Addenda Sheet No.1, dated

, January 1994. The single nominal setpoint is set below the reactor vessel design i pressure (1250 psig) per the requirements of Article 9 of the ASME Code - Section lil, Nuclear Vessels. The setting of 1145 psig..."

and: l

"...The analyses with NEDC-32016P-1, including Errata and Addenda Sheet No.1, l

'. dated January 1994, also..."

j Technical Soecifications Paae 166. Section 4.7.A.2.c.

I

! This change is made to reflect the issuance of TS Amendment 234 (Reference 7). I Amendment 234,in part, deleted TS Page 172, and relocated TS Section

~

i 4.7.A.2.d.(1) to TS Page 166 and renumbered this section as 4.7.A.2.c. This

change affects relocation only and does not affect the technical information previously submitted under JPN-92-028 (Reference 2).

! Replace:

1 l' ...hydrostatically tested at a 1000 psig..."

with:

l "...hydrostatically tested at a 1,035 psig..."

! Technical Snecifications Paae 172. Section 4.7.A.2.d.(1)

! No change is required to TS Page 172. Amendment 234, in part, deleted TS Page 5

172.

i Technical Snecifications Paae 188. BASES Section 3.7 i

i This change only places certain revision bars on TS page 188 in the correct position. Revision bars on this page were not correctly shown in JAFP-96-0306

(Reference 4),

i Technical Soecifications Paae 193. BASES Section 4.7 Add References 20 and 21 after 19 in the proposed submittal (i.e., JPN-92-028).

These three references (i.e.,19,20, and 21) reflect the current calculations regarding radiological consequences of design basis accidents.

1 I

4 j

t

, Attrchmtnt 11 to JPN-96-046 EVALUATION OF UPDATED PAGE CHANGES

. Page 6 of 8 s

1 Add to the proposed submittal (JPN 92-028) that the LOCA dose evaluations, References 19,20, and 21, assumed source term releases were in accordance with TID-14844 and Regulatory Guide 1.3, and were consistent with the Standard Review Plan (NUREG-0800).

These changes do not alter and are consistent with the conclusions presented in References 1 and 2. The Authority previously submitted updated radiological consequences of design basis accidents to the NRC under Reference 4.

These are the only changes to that previously submitted under JPN-92-028.

Replace:

"The design basis loss-of-coolant accident was evaluated in FSAR Section 14.6 incorporating the primary containment maximum allowable accident leak rate of 1.5 percent / day. The analysis showed that with the leak rate and a standby gas treatment system filter efficiency of 99 percent for halogens,99 percent for particulate and assuming the fission product release fractions stated in TID-14844, the maximum total whole body passing cloud dose is about .97 rem and the maximum total thyroid dose is about 11.4 rem at the site boundary over an exposure duration of two hours. The resultant thyroid dose that would occur over a 30-day period is 32.5 rem at the boundary of the low population zone (LPZ).

Thus, these doses are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident."

with:

" 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 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 evaluations, References 19,20, and 21, assumed: the primary containment leak rate was 1.5 volume percent per day; source term releases were in accordance with TID-14844 and Regulatory Guide 1.3, and were consistent with the Standard Review Plan; and the standby gas treatment system filter efficiency was 99% for halogens."

Technical Goecifications Paae 285 and 285a References Section 7.0 New Page 285a is added to the TS to reflect additional references.

There is no change to Reference 10 from that previously submitted under JAFP 0426 (Reference 8). JAFP-9f:-0426 superseded JPN-92-028 (Reference 2) and JAFP-96 0306 (Reference 4) with regards to Page 285.

Reference 18 is changed to add the current revision of the Power Uprate Safety Analysis Report (NEDC-32016P-1, Reference 1) from that previously submitted under JAFP-96-0426.

_.________.._._._____.._._____m_..___.._.___ _ _ . . _ ,

4 Attachm:nt ll to JPN-96-046 j EVALUATION OF UPDATED PAGE CHANGES i Page 7 of 8 1

! Proposed Reference 19 is deleted and replaced with three references, Numbered 4

19,20, and 21. These three references reflect the current calculations regarding

radiological consequences of design basis accidents. Reference 21 will be located on Page 285a. This change supersedes the Reference 19 change submitted under i

JAFP 96-0426.

Proposed Reference 20 is renumbered as Reference 22 (GE-NE-187-45-1191) due to addition of references regarding radiological consequences. In addition, JAFP- i

, 96-0426 erroneously stated that a "P" was at the end of this document. (i.e., GE-

NE-187-451191P). Although this document is proprietary, there is no "P" at the end of the document number. As such, the "P" has been deleted. This reference j- will be located on TS Page 285a. This change supersedes the Reference 20 change i submitted under JAFP-96-0426.

l 1. Replace the following on Page 285:

1

( "(10) ' Nuclear Safety Program Annual Progress Report for Period Ending j_ December 31,1966, Progress Report for Period Ending December 31,

1966, ORNL-4071.'"

[ with:

! "(10) ' Nuclear Safety Program Annual Progress Report for Period Ending l December 31,1966, ORNL-4071.'"  !

2. add the following to Page 285:

i

"(18) General Electric Report NEDC-32016P-1, ' Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant.' April 1993 j (proprietary), including Errata and Addenda Sheet No.1, dated l January 1994.

(19) James A. FitzPatrick Calculation JAF-CALC-RAD-00023, Rev. O,

' Power Uprate Program - Technical Support Center Post-Accident

. Radiological Habitability Study,' August 1996.

I 1

!' (20) James A. FitzPatrick Calculation JAF-CALC-RAD-00042, Rev. O, i ' Control Room Radiological Habitability Under Power Uprate i Conditions and CREVASS Reconfiguration,' September 1995."

3. add the following to Page 285a:

i.

"(21) James A. FitzPatrick Calculation JAF-CALC-RAD-00048, Rev. O, i ' Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents,' May 1996

. (22) General Electric Report GE-NE-187-45-1191, " Containment Systems Evaluation for the James A. FitzPatrick Nuclear Power Plant,"

l November 1991."

i i

l 2

~

I Attachmtnt ll to JPN-96-046 EVALUATION OF UPDATED PAGE CHANGES Page 8 of 8

11. SAFETY IMPLICATIONS OF THE PROPOSED CHANGES There are no safety implications associated with these proposed changes. The Authority has reviewed these proposed changes and has determined that adoption of these changes do not affect the bases or conclusions of the no significant hazards considerations described in NEDC-32016P-1 (Reference 1) and in JPN 028 (Reference 2). The TS changes provided in Attachment I are administrative in nature and support the overall conclusions presented in References 1 and 2. The PORC and SRC have reviewed these proposed changes to the TS and concur with this conclusion.

Vll. REFERENCES

1. General Electric Report NEDC-32016P-1, " Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant," April 1993 (proprietary),

including Errata and Addenda Sheet No.1, dated January 1994

2. NYPA Letter, R. E. Beedle to the NRC, (JPN-92-028), " Proposed Changes to the Technical Specifications Regarding Power Uprate (JPTS-91-025)," dated June 12,1992

3. NYPA Letter, W. J. Cahill, Jr. to the NRC, (JPN-96-043), " Response to Request for Additional Information Regarding Power Uprate," dated November 14,1996

4. NYPA Letter, M. J. Colomb to the NRC, (JAFP-96-0306). " Updated Page Changes for Proposed Change to the Technical Specifications Regarding Power Uprate (JPTS-91-025)," dated August 15,1996

5. General Electric Topical Report NEDO-21662, Revision 2, " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead  :

Plant)," July 1977 with errata and addenda. ,

6. General Electric Topical Report NEDC-31317P, " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,"

October 1986 with revisions, errata and addenda

7. NRC Letter, K. R. Cotton to W. J. Cahill Jr., Regarding issuance of

~

Amendment 234 for the James A. FitzPatrick Nuclear Power Plant (TAC No. ,

M95099), dated October 4,1996

8. NYPA Letter, M. J. Colomb to the NRC, (JAFP-96-0426), " Submittal of Updated Pages Regarding Proposed Changes to the Technical Specifications Contained in the Referenced Letters," dated October 23,1996

Attrchm:nt lll ts JPN-96-046 l l

MARKUP TO REFLECT UPDATED PAGE CHANGES FOR PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS REGARDING POWER UPRATE (JPTS-91-025) l l

NOTE 1: Deletions are shown in Otrikeout, and additions are in bold.

NOTE 2: Previous amendment revision bars are shown and will be deleted.  !

l l

l l

l l

I 1

i l

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

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

i JAFNPP

AD. Core Operatina Limits Report (CQLR1 Z. Too of Active Fuel ,

This report is the plant-specific document that .

The Top of Active Fuel, corresponding to the top of provides the core operating limits for the current the enriched fuel column of each fuel bundle, is operating cycle. These cycle-specific operating located 352.5 inches above vessel zero, which is limits shall be determined for each reload cycle in  !

the lowest point in the inside bottom of the reactor accordance with Specification 6.9.A.4. Plant <

vessel.' (See General Electric drawing No. operation within these operating limits is addressed 919D690BD.) in individual Technical Specifications.  !

AA. Rod Density AE. References Rod density is the number of control rod notches 1. General Electric Report NEDC-32016P-1,  !

inserted expressed as a fraction of the total number " Power Uprate Safety Analysis for James A. i of control rod notches. All rods fully inserted is a FitzPatrick Nuclear Power Plant," April 1993 !

condition representing 100 percent rod density. (propdetary), including Errata and Addenda l Sheet No.1, dated January 1994. i AB. Purae-Puraina l 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 purify the confinement.

AC. Ventina  ;

I Venting is the controlled process of releasing air or t gas from a confinement in such a manner that replacement air or gas is not provided or required. l b

t i

1 i e

Amendment No. 75, 92, ? S2, 6a ,

JAFNPP 1.1 BASES (Cont'd) .

E. References C. Power Transient

1. " General Electric Standard Application for Plant safety analyses have shown that the scrams Reactor Fuel," NEDE-24011-P, ' .:;;; ;;;rc;d caused by exceeding any safety system setting will  :;.id;n ;nd ;=;c.42;..:.. A-13, August 1996, assure that the Safety Limit of 1.1.A or 1.1.8 will not be exceeded. Scram times are checked periodically to 2. FitzPatrick Nuclear Power Plant Single-Loop assure the insertion times are adequate. The thermal Operation, NEDO 24281, August 1980.

power transient resulting when a scram is a'ccomplished other than by the expected scram signal 3. GE12 Compliance with Amendment 22 of (e.g., scram from neutron flux following closure of the NEDE-24011-P-A (GESTAR 11), NEDE-32417P, i main turbine stop valves) does not necessarily cause December 1994. -

fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only i accomplished by means of a backup feature of the ,

plant design. The concept of not approaching a l Safety Limit provided scram signals are operable is 1 supported by the extensive plant safety analysis.  ;

D. Reactor Water Level (Hot or Cold Shutdown  !

Condition)

During periods when the reactor is shut down,  !

consideration must also be given to water level  !

requirements due to the effect of decay heat. If i reactor water level should drop below the top of the active fuel during this time, the ability to cool the core ,

is reduced. This reduction in core cooling capability  ;

could lead to elevated cladding temperatures and clad  ;

perforation. The core will be cooled sufficiently to  !

prevent clad melting should the water level be reduced to two-thirds the core height. Establishment i of the Safety Limit at 18 in. above the top of the fuel -

provides adequate margin. This level will be continuously monitored whenever the recirculation '

pumps are not operating. t Amendment No. 'i, 90,162,238, 14 ,

JAFNPP 2.1 BASES (Cont'd) i B. Not Used C. References

1. (Oc!cicd) General Electric Report, NEDC-32016P-1,

" Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant," April 1993 (proprietary), including Errata and Addenda Sheet No.1, dated January 1994.

2. " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A (App cycd rc.icien number Opp!!ccb!c c: t5.0 thct rc!ced fuc! Onc!yccc crc perfer=cd: 13, August 1996.

3. (Deleted)

4. FitzPatrick Nuclear Power Plant Single-Loop Operatior:,

NEDO-24281, August,1980.

i I

1 Amendment No. , 4 0, Si, 98,162,100, 20 (Next page is 23)

JAFNPP ,

i i

1.2 and 2.2 BASES The reactor coolant pressure boundary integrity is an important The limiting vessel overpressure transient event is a main steam barrier in the prevention of uncontrolled release of fission products. isolation valve closure with flux scram. This event was analyzed it is essential that the integrity of this boundary be protected by within N5DC 31S97", "Upd:::d SRV "sfnunx " g;;;x:: f= l c::tablishing a pressure limit to be observed for all operating $c J.^7N'"," cxu"n; 9 ef 2: 11 S".V: c;r: ;;;b!: ;;hh  !

conditions and whenever there is irradiated fuel in the reactor p:Hng p:x ::: !:= $:n = :;=! :: 1195 ;dg. W :::ch::4  !

vissel. p :k V:x ! g:xu f= $ :nn: :: 1 r;- :: 50 !::: :Pn2 v:=c! p;nu : ::d: F .h of 1375 ;d;. 2: : ;;x: ::!=d The pressure safety limit of 1,325 psig as measured by the vessel en:F,-d fa 2: ::::t= x;;n : :: $: m:!n c'::= ::d::! v:!= i st am space pressure indicator is equivalent to 1,375 psig at the  !:== .;!$ f!u =:= Ov nt). % :!= cf '195 pd; :: 2: S".V lowest elevation of the Reactor Coolant System. The 1,375 psig sp:Nng g ::u up : c;Plch p'n; ;;"sm =: Pc: 5: n :n:P, :d, i v;lue is derived from the design pressures of the reactor pressure  :::=ing 2 ERV: =0 n;;;;b!:. Th xfac, SRV p:dng p::===

vsssel and reactor coolant system piping. The respective design be!:re ' 105 ;d; en=0 $0: $ .^.S".'5 C:d: Fm!: On ph :::::=

• pressures are 1250 psig at 575 'F for the reactor vessel,1148 psig p := : :::::f!:d.NEDC-32016P-1, " Power Uprate Safety (;

st 568 *F for the recirculation suction piping and 1274 psig at 575 Analysis For James A. FitzPatrick Nuclear Power Plant," including

• F for the discharge piping. The pressure safety limit was chosen Errata and Addenda Sheet No.1, dated January 1994, assuming 9 7 as the lower of the pressure transients permitted by the applicable of the 11 SRVs were operable with opening pressures less than or dssign codes: 1965 ASME Boiler and Pressure Vessel Code, equal to 1179 psig. The resultant peak vessel pressure for the i S:ction lli for pressure vessel and 1969 ANSI B31.1 Code for the event was shown to be less than the ASME Code limit of 1375 peig reactor coolant system piping. The ASME Boiler and Pressure (see current reload analysis for the reactor response to the main Vcssel Code permits pressure transients up to 10 percent over steam isolation valve closure with flux scram event).  :

dssign pressure (110% x 1,250 = 1,375 psig) and the ANSI Code  !

permits pressure transients up to 20 percent over the design A safety limit is applied to the Residual Heat Removal System pressure (120% x 1,150 = 1,380 psig). The safety limit pressure (RHRS) when it is operating in the shutdown cooling mode. When  !

of 1,375 psig is referenced to the lowest elevation of the Reactor operating in the shutdown cooling mode, the RHRS is included in  ;

Coolant System. the reactor coolant system.

i i

Amendment No. 58, Si,131,190, 217, 29

JAFNPP 3.1 BASES (cont'd) ,

Turbine control valves fast closures initiates a scram based C. References on pressure switches sensing electro-hydraulic control (EHC) system oil pressure. The switches are located between fast 1. C cnc :! E! Ori T pic:! R:p;n t?EDO 21SS2, Rc;iden closure solenoids and the disc dump valves, and are set 2, "L :: Of C::': : ^.00! dent ^.n;!y:'; Rep;" f:-

relative (500< P< 850 psig) to the normal (EHC) oil pressure J:m:0 ^. Pt:" ::! k Mus!::: ": :: "ent (L d '

of 1.600 psig so that based on the small system volume, "cnt)", Ju!y 1977 re $ :::: : Ond Oddende.

they can rapidly detect valve closure or loss of hydraulic pr. essure. 2. Cen:::! E!::::i T p?: ! ":p;n t?EOC 31317P, " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR-The requirement that the IRM's be inserted in the core when LOCA Loss-of-Coolant Accident Analysis,", Oct:50:

the APRM's read 2.5 indicated on the scale in the start-up 1985 rdh : vid: .0, ::::::nd Oddende NEDC-and refuel modes assures that there 6 proper overlap in the 31317P, Revision 2, April 1993.

neutron monitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation.

B. The limiting transient which determines the required steady state MCPR limit depends on cycle exposure. The operating limit MCPR values as determined from the transient analysis in the current reload submittal for various core exposures are specified in the Core Operating' Limits Report (COLR).

The ECCS performance analyses assumed reactor operation will be limited to the MCPR value for each fue! type as described in NEDO 21SS2 (Rcfcc ne: 1) Ond NEOC 31317 iReferone 2) in !udin;; the ! :::t vi !:n, :::: Ond Oddende Reference 1. The Technical Specifications limit operation of the reactor to the more conservative MCPR based on consideration of the limiting transient as specified in the COLR.

Amendment No. 49, Si,109,1S2, 90 cited by N9C !:nce d:::d 3/18/93, 35

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

t JAFNPP 3.3 and 4.3 BASES (cont'd) i

" full out postion during the performance of SR 4.3.A.2.a. 3. The Rod Worth Minimizer (RWM) restricts the order of ,

This Frequency is acceptable,- considering the low control rod withdrawal and insertion to be equivalent to the probability that a control rod will become uncoupled when Banked Position Withdrawal Sequence (BPWS). These  !

it is not being moved, and operating experience related to sequences are established such that the drop of any  !

uncoupling events. in-sequence control rod from the fully inserted position to the position of the control rod drive would not ceuse the  ;

reactor to sustain a power excursion resulting in a peak fuel  :'

2. The control rod housing support restricts the outward enthalpy in excess of 280 c91/gm. An enthalpy of 280 movement of a control rod to less than 3 in. in the cal /gm is well below the level at which rapid fuel dispersal  !

extremely remote tevent of a housing failure. The amount could occur (i.e. 425 cal /gm.). Primary system damage in j of reactivity which could be added by this small amount of this accident is not possible unless a significant amount of  !

rod withdrawal, which is .less than a normal single fuel is rapidly dispersed. Ref. Subsections 3.6.6, 7.7.4.3 j withdrawal increment, will not contribute to any damage to and 14.6.1.2 of the FSAR, NEDE-24011-P-A-13, August  :

the Primary Coolant System. The design basis is given in 1996 and NEDO-10527 including supplements 1 and 2 to r subsection 3.8.2 of the FSAR, and the safety evaluation is NEDO-10527. .

given in subsection 3.8.4. This support is not required if f the Reactor Coolant System is at atmospheric pressure in performing the function described above, the RWM is not [

since there would then be no driving force to rapidly eject required to impose any restrictions at core power levels in ,

a drive housing. Additionally, the support is not required if excess of 10% of rated. Material in the cited references  ;

all control rods are fully inserted and if an adequate shows that it is impossible to reach 280 calories per gram .

shutdown margin with one control rod withdrawn has been in the event of a control rod drop occurring at power  !

demonstrated, since the reactor would remain subcriticai greater than 10%, regardless of the rod pattern. This is - ,

even in the event of complete ejection of the strongest true for all normal ar d abnormal pattems including those  :

control rod. which maximize the individual control rod worth. t i

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i Amendment No. 30,155,193, -

100  !

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JAFNPP 3.3 and 4.3 BASES (ct at'd) i-

'5. The Rod Block Monitor (RBM) is designed to automatically C. Scram insertion Times prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level The Control Rod System is designated to bring the reactor operation. Two channels are provided, and one of these may be subcritical at a rate fast enough to prevent fuel damage;i.e.,

bypassed from the console for maintenance' and/or testing. to prevent the MCPR from becoming less than the Safety Tripping of one of the channels will block erroneous rod Limit. Scram insertion time test critseia o' Section 3.3.C.1 ,

withdrawal soon enough toprevent fuel damage. were used to generate the generic scram reactivity curve shown in NEDE-24011-P-A-13, August 1996. This generic This system backs up the operator who withdraws control rods curve was used in analysis d non-pressurization transients ; ,

according to written sequences. The specified restrictions with to determine MCPR limits. Thereic e. the required protection one channel out of service conservatively assure that f. net is provided. i damage will not occur due to rod withdrawal errors when this condition exists.

The numerical values assigned - to the specified scram A limiting control rod pattern is a pattern which results in the performance are based on tha analysis of data from other l core being on a thermal hydraulic timit (e.g., MCPR iimit). During BWR's with control rod dnves the same as those on  !

use of such patterns, it is judged that testing of the RBM System JAFNPP.

prior to withdrawal of such rods to assure its operability will assure that improper withdraw does not occur. It is the The occurrence of scram -times within the limits, but responsibility of the Reactor Engineer to identify these limiting significantly longer than the average, should be viewed as an patterns and the designated rods either when the patterns are indication of a systematic problem with control rod drives, initially established or as they develop due to the occurrence of especially if the number of drives exhibiting such scram' inoperable control rods in other than limiting patterns. times exceeds eight, the allowable number of inoperable rods.

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Amendment No. 'i, 10,21,30,13,19,53,S7,155,1S2, 102 '

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JAFNPP

-3.5 BASES A. Core Sorav system acd_Lo;< Pressure Coolant iniection (LPCl) Core spray distribution has been shown, in full scale tests of Mode c,? the RHR Systei? systems similar in design to that of the FitzPatrick Plant, to exceed the minimum requirements by at least 25 percent. In addition, This speci'.ication asswas that adequate emergency cooling cooling effectiveness has been demonstrated at less than half the capability' s available whenever irradiated fuel is in the reactor rated flow in simulated fuel assemblies with heater rods to i vessel. duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additionally conservative in that no credit is The loss-of-coolant analysis is referenced and described in taken for spray coolant entering the reactor before the internal General Electric Topical Report NEDE-24011-P-A-13, August pressure has fallen to 113 psi above primary containment 1996. pressure.

The LPCI mode of the RHR System is designed to provide The limiting conditions of operation in Specifications 3.5.A.1 through 3.5.A.6 specify the combinations of operable emergency cooling to the core by flooding in the event of a subsystems to assure the availability of the minimum cooling loss-of-coolant accident. . These subsystems are completely ,

systems. No single failure of ECCS equipment occurring during independent of the Core Spray System; however, they function in a loss-of-coolant accident under these limiting conditions of combination with the Core Spray System to prevent excessive fuel t operation will result in inadequate cooling of the reactor core. clad temperature. The LPCI mode of t

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i JAFNPP 3.6 and 4.6 BASES (cont'd) l E. Safetv/ Relief Valves i

The safety / relief valves (SRVs) have two modes of operation; the with the HPCI and RCIC turbine overspeed systems and the Mark 1 I torus loading analyses. Based on safety / relief valve testing safety mode or the relief mode. In the safety mode (or spring mode of operation) the spring loaded pilot valve opens when the experience and the analysis referenced above, the safety / relief. ,

steam pressure at the valve inlet overcomes the spring force valves are bench tested to demonstrate that in-service opening holding the pilot valve closed. The safety mode of operation is pressures are within the nominal pressure setpoints 13% and  ;

required during pressurization transients to ensure vessel then the valves are returned to service with opening pressures at  !

pressures do not exceed the reactor coolant pressure safety limit the nominal setpoints 11%. In this manner, valve integrity is  ;

of 1,375 psig. maintained from cycle to cycle.

• In the relief mode the spring loaded pilot valve opens when the The analyses with NEDC-3449'7P 32016P-1,Inclus5nOErrata and spring force is overcome by nitrogen pressure which is provided Addenda Sheet No.1, dated January 1934, also provide the to the valve through a solenoid operated valve. The solencid safety basis for which 2 SRVs are permitted inoperable during i operated valve is actuated by the ADS logic system (for those continuous power operation. With more than 2 SRVsinoperable, SRVs which are included in the ADS) or manually by the operator the margin to the reactor vessel pressure safety' limit is i from a control switch in the main control room or at the remota significantly reduced, therefore, the plant must enter a cold  ;

ADS panel. Operation of the SRVs in the relief mode for the ADS condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> once more than 2 SRVs are  ;

is discussed in the Bases for Specification 3.b.D. determined to be inoperable. . (See reload evaluation for the  :

current cycle).  :

Experiences in safety / relief valve testing have shown that failure  :

or dete;ioration of safety / relief valves can be adequately detected A manual actuation of each SRV is performed to demonstrate i if at least 5 of the 11 valves are bench tested once every 24 that the valves are mechanically functional and that no blockage '

months so that all valves are tested every 48 months. exists in the valve discharge line Valve opening is confi.Tr.eo by Furthermore, safety / relief valve testing experience has monitoring the response of the turbine bypass valves end the demonstrated that safety / relief valves which actuate within 13% SRV acoustic monitors. Adequate reactor steam dems pressure  ;

of the design pressure setpoint are co.'sidered operable (see must be available to avoid damaging the valve. Adequate steam ,

ANSI /ASME OM-1-1981). The safety bases for a single nominal flow is required to ensure that reactor pressure can be  :

valve opening pressure of 4440 1145 psig are described in maintained during the test. Testing is performed in the RUN '

NEDC 31 S97", "Upd ted SRV t ' -.;;c.;; ".:qim;cac. f= $ mode to reduce the risk of a reactor scram in response to small J^ 9'" " 32016P-1, " Power Uprate Safety Analysis for James pressure flueuetiene fluctuadons which may occur while opening l A. FitrPatrick Nuclear Power Plant," inclus5ng Errata and Addenda and reclosing the valves.  ;

Sheet No.1, dated January 1994. The single nominal setpoint is set below the reactor vessel design pressure (1250 psig) per the Low power physics testing and reactor operator training with requirements of Article 9 of the ASME Code - Section lil, Nuclear inoperable components will be conducted ' only when the ,

Vessels. The setting of 4444 1145 psig preserves the safety safety / relief valves are margins associated Amendment No. 12,131,217,219,229, t 152  ;

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JAFNPP 'I 3.7 (cont'd) 4.7 (cont'd)-

(2) During testing which adds heat to the suppression pool,  ;

the water temperature shall not exceed 10*F above the .  !

normal power operation limit specified in (1) above. In i connection with such testing, the pool temperature must i be reduced to below the normal power operation limit -

specified in (1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

(3) The reactor shall be scrammed from any operating condition if the pool temperature reaches 110*F. Power l operation shall not be resumed until the pool temperature :l is reduced below the normal power operation limit  ;

specified in (1) above. l (4) During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig at  ;

normal cooldown rates if the pool temperature reaches i 120*F. f

2. Primary containment integrity shall be maintained at all times ,

when the reactor is critical or when the reactor water 2. a. Perform required visual examination and leakage rate temperature is above 212*F, and fuel is in the reactor vessel, testing of the Primary Containment in accordance with -

except while performing low power physics tests at the Primary Containment Leakage Rate Testing Program. I atmospheric pressure at power levels not to exceed 5 MWt.  :

b. Demonstrate leakage rate through each MSlV is's 11.5  !

scfh when tested at 2 25 psig. - The testing frequency is i in accordance with the Primary Containment Leakage  !

Rate Testing Program.  !

c. Once per 24 months, demonstrate the leakage rate of i 10AOV-68A,B for the Low Pressure Coolant injection .t system and 14AOV-13A,B for the Core Spray system to S be less than 11 scfm per valve when pneumatically  !

tested at 2 45 psig at ambient temperature, or less than l 10 gpm per valve if hydrostatically tested at 2 4000 .[

1,035 psig at ambient temperature. t i

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Amendment No. 44,-334, 166 [

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JAFNPP 3.7 BASES (cont'd)

Using the minimum or maximum torus water level (which are INSERT A based on downcomer submergence levels where 13.88 feet Using : 10 F i= (S::ti:n 5.2 FS??) :- the t=u: =:t= I above the bottom of the torus is 0.005 feet higher than the temp = t=0 :nd m:xim= Sitic! t;mp==:=0 Of 95 F, minimum submergence of 51.5 inches and 14.00 feet above tempact=0 of

• 15 F :: cchieved, which ?: = !! 5:!:= the the bottom of the torus is equivalent to the maximum 170 e ::=p= t=0 which i: u;;d fr ::mp'Ot : .d:n= tion submergence of 53 inches assumed in containment analyses)  ;

containment pressure during the design basis accident is F= n : tic! m:xim=- tru: = t= ::=pr:t=0 Of 95 " nd I approximately 45 psig which is below the design of 56 psig. ==u : :ng the n= :! ::=p!;m:nt Of : ntd. m;nt 00 'ing The minimum downcomer submergence of 51.5 inches results pump: (t=: L"C! ;;mp: :nd :=0 P.H ::rci : =:t= pump )

in a minimum torus water volume of approximately 496400 00ntcin .cnt p::::=0 5 n0 ::guired t m !nt in Od:ge :: not 105,900 feet . The majority of the Bodega tests (9) were run pe:itive cuction 5;:d (""SH) f= the : : Op::y L"C! :nd ""C!

I with a submerged length of 4 feet and with complete pump:.

condensation. Thus, with respect to downcomer submergence, this spacification is adequate. Additional Limiting suppression pool temperature to 430105 F during JAFNPP specific analyses done in connection with the Mark i RCIC, HPCI, or relief valve operation, when decay heat and Containment-Suppression Chamber integrity Program indicate stored energy are removed form from the primary system by ,

the adequacy of the specified range of submergence to ensure discharging reactor steam directly to the torus assures I '

that dynamic forces associated with pool swell do not result in adequate margin for a potential blowdown any time during overstress of the torus or associated structures. Level RCIC, HPCI, or relief valve operation.

instrumentation is provided for operator use to maintain downcomer submergence within the specified range. INSERT B l Enp=:- Onte! det: indic ::: th:t ensc dv ::::= conden ing The maximum temperature at the end of blowdown tested !ced: een bc evoided " the p :k temp =:t= Of the during the Humboldt Bay (10) and Bodega Bay tests was supp :=:en poc! ?: m: int:ined bc! = 1SO r dring ny pried .

170 F, and this is conservatively taken to be the limit for Of : !! f v !ve Op=: tion =?th conic condition et the d!=h=g l comp! tc conden=tica cf the !!mit f= complete condensation exit. Specification: have 50:n p!:00d on the enve!:p Of of the reactor coolant, although condensation would occur for 7000t= cp=ct:ng ::ndition: Oc that th :::ct= :n 50 temperatures above 170 F. dep::::rized in time!y menn= te Ov:!d th reg!== cf potertic!!y high true !:: ding:. I r

t Amendment No. 16, 3S,1 SS,181,107, 188

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l i I i Channes to Technical Snecification Pane 188 i i

INSERT A l

i. Containment analyses predict a 46*F increase in pool water temperature, after complete 1

LOCA blowdown. These analyses assumed an initial suppression pool water temperature of 95 F and a rated reactor power of 2536 MWt. LOCA analyses in Section 14.6 of the j 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 3

170 F temperature seen during the Bodega Bay tests.

j For an initial maximum torus water temperature of 95oF, assuming the worst case

complement of containment cooling pumps (one LPCI pump and two RHR service water I pumps), containment pressure is required to maintain adequate net positive suction head (NPSH) for the core spray and LPCI pumps.

1 i INSERT B t

! Experiments indicate that unacceptably high dynamic containment loads may result from l unstable condensation when suppression pool water temperatures are high near SRV l discharges. Action statements limit the maximum pool temperature to assure stable

! condensation. These actions include: limiting the maximum pool temperature of 95oF during normal operation; initiating a reactor scram if during a transient (such as a stuck 4 open SRV) pool temperature exceeds 110oF; and depressurizing the reacbr if pool

temperature exceeds 120 F. T-quenchers diffuse steam discharged from SRv's and

promote stable condensation. The presence of T-quenchers and compliance with these i

action statements assure th'at stable condensation will occur and containment loads will be acceptable.

JAFNPP ,

I 4.7 BASES A. Primary Containment The d:dgn tod;! :: Of :::!:nt :: !de-'  :: :::!u-ted E-FS^ S :t!:n 'i.S in =;r :n; $c ;-bcry nC :n.m:nt The water in the suppression chamber is used only for mammum-ellewebic ::dd;= !::t :::: Of *.5 pr :=/d y. ,

cooling in the event of an accident; i.e., it is not used for '50 :nc!yi 'cr:d St =?$ $c !::k :::: cri : :ndiv normal operation; therefore, a daily check of the g ::::Cm;= ;y;ter 9! = cffidency of 99 gan= f=

temperature and volume is adequate to assure that M! gene, 99 p= cent f= p--'! u!::: cid ;xx"n; &

adequate heat removal capability is present. f!::!:n ; ; duct ::!: '::t?;n: :::t;d in '!D

• 194'. $c m  !=u- :':! =hd: b;dy p dn; drud d;x 50 :t:3t The primary containment preoperational test pressures are .97 ::= :nd $c m:2.ur ::t:! $y : d d;x 5: t;u' '

based upon the calculated primary containment pressure 70m et $0 05t0 bou-dry Ovr en :=p;;r: i. C: n of two response corresponding to the design basis loss-of-coolant '0=.. '50 :::c' tent $y :!d d::: $c :x'd :::= == 0 i accident. The peak drywell pressure would be about 45 30 dcy pr!:d i 32.5 ::= :: $c be= dry Of & !:=

psig which would rapidly reduce to 27 psig within 30 sec. p:pu' tion :0 0 'LPZ). Thu , 5:00 d:::: =: & -

following the pipe break. Following the pipe break, the men mur St w:u!d 50 :=pected h $0 =!i !y :- nt Of t

suppression chamber pressure rises to 26 psig within 30  : d::!;n hd !::: Of :::!:n' :: ! dent. Design basis sec, equalizes with drywell pressure and thereafter rapidly accidents were evaluated as discussed in Section 14.6 of decays with the drywell pressure decay (14). the FSAR and the power uprate safety evaluation, Reference 18. The whole body and thyroid doses in the The design pressure of the drywell and suppression control room, low population zone (LPZ) and alte boundary chamber is 56 psig(15). The design basis accident meet the requirements of 10 CFR Parts 50 and 100. The 1 leakage rate is 0.5 percent / day at a pressure of 45 psig. technical support center (TSC), not designed to these As pointed out above, the drywell and suppression licensing bases, was also analyzed. The whole body and  !

chamber pressure following an accident would equalize thyroid dose acceptance criteria used for the main control fairly rapidly. Based on the primary containment pressure room are met for the TSC when initial access to the TSC response and the fact that the drywell and suppression and occupancy of certain areas in the TSC is restricted by -

chamber function as a unit, the primary containment will administredve control. The LOCA dose evaluations, I be tested as a unit rather than the individual components Referar.cos 19, 20, and 21, assumed: the primary separately. containment leak rate was 1.5 volume percent per day; source term releases were in accordance with TID-14844 l and Regulatory Guide 1.3, and were consistent with the  !

Standard Review Plan; and the standby gas treatment system filter efficiency was 99% for halogens. These  ;

doses are also based on the-Amendment No. >

193

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JAFNPP

7.0 REFERENCES

(1) E. Janssen, " Multi-Rod Burnout at Low Pressure," ASME Paper (9) C.H. Robbins, " Tests of a Full Scale 1/48 Segment of the 62-HT-26, August 1962. Humbolt Bay Pressure Suppression Containment,"

GEAP-3596, November 17,1960.

(2) K.M. Backer, " Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters," AE-74 (Stockholm, Sweden), May (10) " Nuclear Safety Program Annual Progress Report for Period 1962. Ending December 31,1966, N g:::: R;; n for P :!:d Ending Decembc: 31,10SS, ORNL-4071."

(3) FSAR Section 11.2.2.

(11) Section 5.2 of the FSAR.

(4) FSAR Section 4.4.3.

(12) TID 20583, " Leakage Characteristics of Steel Containment (5) 1.M. Jacobs, " Reliability of Engineered Safety Features as a Vessel and the Analysis of Leakage Rate Determinations."

Function of Testing Frequency," Nuclear Safety, Vol. 9, No. 4, July-August 1968, pp 310-312. (13) Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program", dated September 1995.

(6) Deleted .

(14) Section 14.6 of the FSAR.

(7) 1.M. Jacobs and P.W. Mariott, APED Guidelines for Determining Safe Test Intervals and Repair Times for (15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, Engineered Safeguards - April 1969. Section 111. Maximum allowable internal pressure is 62 psig.

(8) Bodega Bay Preliminary Hazards Report, Appendix 1, Docket (16) 10 CFR Part 50, Appendix J, " Primary Reactor Containment ~

50-205, December 28,1962. Leakage Testing for Water-Cooled Power Reactors, Option B -

Performance Based Requirements", Effective Date October 26, 1995 (17) Deleted INSERT A Amendment No. 190,227,234, 285

1 l

Channes to Technical Specification Paae 285 INSERT A (18) General Electric Report NEDC-32016P-1, " Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant," April 1993 (proprietary), including Errata and Addenda Sheet No.1, dated January 1994.

(19) James A. FitzPatrick Calculation JAF-CALC-RAD-00023, Rev. O, " Power Uprate Program - Technical Support Center Post-Accident Radiological Habitability Study,"

August 1996.

(20) James A. FitzPatrick Calculation JAF-CALC-RAD-00042, Rev. O, " Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration," September 1995.

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I JAFNPP  !

7.0 REFERENCES

(continued)

(21) James A. FitzPatrick Calculation JAF-CALC-RAD-OOO48, Rev. O, " Power Uprate Project - Radiological impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents," May 1996 L

(22) General Electric Report GE-NE-187-45-1191, ,

" Containment Systems Evaluation for the James A. .

FitzPatrick Nuclear Power Plant," November 1991 (proprietary).

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Amendment No.

285a i

k Attachm:nt IV to JPN-96-046 ERRATA AND ADDENDA SHEET NO.1, DATED JANUARY 1994 FOR GENERAL ELECTRIC REPORT NEDC-32016P-1 1

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i New York Power Authority i

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 l

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ERRATA AND ADDENDA SHEET NO.1, JANUARY 1994 1

FOR I

GENERAL ELECTRIC REPORT NEDC-32016P-1 The following corrections apply to Table 5-1 (page 5-8) of General Electric Report NEDC-32016P-1, " Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant,"

dated April 1993.

4 For the parameter listed as " Vessel High Pressure Scram":

1, i 1) The value "1045 psig" should be "1059 psig"

2) The value "1080 psig" should be "1094 psig" l The values 1045 psig and 1080 psig are Technical Specification limits. The corrected values,1059 psig and 1094 psig respectively, are the analyticallimits which were intended to be reported in this table.

The text in this report (Section 5.1.2.1, page 5-3) associated with Table 5-1 is correct as written.

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