JPN-88-023, Proposed Tech Specs,Correcting Typos

Proposed Tech Specs,Correcting Typos

ML20154R048
Person / Time
Site:	FitzPatrick
Issue date:	05/27/1988
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20154R034	List: JPN-88-023, Proposed Tech Specs,Correcting Typos ML20154R030 ML20154R037
References
JPN-88-023, JPN-88-23, NUDOCS 8806070198
Download: ML20154R048 (87)

Text

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ATTACHMENT I TO JPN-88-023 PROPOSED ADMINISTRATIVE CHANGES TO THE TECHNICAL SPECIFICATIONS (JPTS-8 6-004 )

NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT l DOCKET NO. 50-333 i DPR-59 l

8806070198 880527 PDR ADOCK 05000333 l _P DCD

JAFNPP TECHNICAL SPECIFICATIONS TABLE OF CONTENTS PA9A 1.0' Definitions 1 LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 Fuel Cladding Integrity 2.1 7 1.2 Reactor Coolant System 2.2 27 SURVEILLANCE LIMITING CO"DITIONS FOR OPERATION REOUIREMENTS 3.0 General 30 l 3.1 Reactor Protection System 4.1 30f l

3.2 Protective Instrumentation 4.2 49 A. Primary Containment Isolation Functions A. 49 B. Core and Containment Cooling Systems - B. 49 Initiation and Control C. Control Rod Block Actuation C. 50 D. Radiation Monitoring Systems - Isola- D. 50 tion and Initiation Functions E. Drywell Leak Detection E. 54 F. Surveillance Infot. nation Readouts F. 54 G. Recirculation Pump Trip G. 54 3.3 Reactivity Control 4.3 88 l

l A. Reactivity Limitations A. 88

! B. Control Rods B. 91 C. Scram Insertion Times C. 95 D. Reactivity Anomalies D. 96 3.4 Standby Liquid Control System 4.4 105 A. Normal Operation A. 105 B. Operation With Inoperable Components B. 106 C. Sodium Pentaborate Solution C. 107 I

3.5 Core and Containment Cooling Systems 4.5 112 l A. Core Spray and LPCI Systems A. 112 B. Containment Cooling Mode of the RHR B. 115 System C. HPCI System C. 117 D. Automatic Depressurization System D. 119 (ADS)

E. Reactor Core Isolation Cooling (RCIC) E. 121 System AmendmentNo.pd i

JAFNPP LIST OF TABLES (CONT'D)

Table Title Pace 4.6-1 Comparison of the James A. FitzPatrick Nuclear 157 Power Plant Inservice Inspection Program to ASME Inservice Inspection Code Requirements 3.7-1 ' Process Pipeline Penetrating Primary Containment 198 4.6-2 Minimum Test and Calibration Frequency for Drywell 162a

-Continuous Atmosphere Radioactivity Monitoring System 4.7-1 Minimum Test and Calibration Frequency for 210 Containment Monitoring Systems .

4.7-2 Exception to Type C Tests 211 3.12-1 Water Spray / Sprinkler Protected Areas 244j 3.12-2 Carbon Dioxide Protected Areas 244k 3.12-3 Manual Fire Hose Stations 2441 4.12-1 Water Spray / Sprinkler System Tests 244q 4.12-2 Carbon Dioxide System Tests 244r 4.12-3 Manual Fire Hose Station Tests 244s 6.2-1 Minimum Shift Manning Requirements 260a 6.10-1 Component Cyclic or Transient Limits- 261 AmendmentNo.pd,ph,pI,pIl vi

JAFNPP LIST OF FIGURES Floure Title Pace 3.1-1 Manual Flow Control 47a 3.1-2 Operating Limit MCPR versus I 47b 4.1-1 Graphic Lid in the Selection of an Adequate Interval Between Tests 48 4.2-l' 1est~1nterval vs. Probability of System Unavailability 87 3 4-1 Sodium Pentaborate. Solution of System Volume-Concentration Requirements 110-3.4-2 Saturation Temperature of Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Limits of Specifications 3.5.J.1, 3.5.J.2 and 3.5.J.3 134 l

3.5-6 (Deleted) 135d 3.5-7 (Deleted) 135e 3.5-8 (Deleted) 135f 3.5-9 (Deleted) 135g l 3.5-10 'MAPLHGR Versus Planar Average Exposure Reloads 4 & 5, P8DRB299 135h 3.5-11 MAPLHGR Versus Planar Average Exposure Reload 6 & '/,

BP8DRB299, QUAD + 1351 3.5-12 MAPLHGR Versus Average Planar Exposure Reload 7 BD319A 135j 3.6-1 Reactor Vessel Thermal Pressurization Limitations 163 4.6-1 -Chloride Stress Corrosion Test Results at 500*F 164 6.1-1 Management Organization Chart 259 6.2-1 Plant Staff Organization 260

/ / / / / , /

Amendment No. 14, 22, $3, 94, 72,74, 88, 98. 109, 113 vil 1

4 JAFNPP TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS i

l The succeeding fTequently used. terms are explicitly defined so that a uniform interpretation of the Opecifications may be achieved.

A. Reportable Event - A reportable event shall be any of those conditions specified in Section 4

50.73 to 10 CFR Part 50.

B. Core Alteration - The act of moving any component in the. region above the core support plate, below the upper grid and within the shroud.- Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration. Normal movement of in-core instrumentation is not defined as a core alteration.

AmendmentNo.f,[2, 1[O 1

JAFNPP 1.0 (cont'd)

C. Cold Condition - Reactor coolant temperature 3. Instrument Channel - An instrument channel 6.212*F. means an arrangement of a sensor and auxil-iary equipment required to generate and D. Hot Standby Condition - Hot Standby condition transmit to a trip system a single trip means operation with coolant temperature > 212*F, signal related to the plant parameter the Mode Switch in Startup/ Hot Standby and monitored by that instrument channel, reactor pressure <" l,005 psig.

4. Instrument Check - An instrument check is a l E. Immediate - Immediate means that the required qualitative determination of acceptable action will be initiated as soon as practicable operability by observation of instrument considering the safe operation of the unit and behavior during operation. This determina-the importance of the required action. tion shall include, where possible, compar-ison of the instrument with other independent F. Instrume_ntation instruments measuring the same variable.

1. Functional Test - A functional test is the S. Instrument Channel Functional Test - An manual operation or initiation of a system, instrument channel functional test means the subsystem, or component to verify that it injection of a simulated signal into the functions within design tolerances (e.g., instrument primary sensor where possible to the manual start of a core spray pump to verify the proper instrument channel re-verify that it runs and that it pumps the sponse, alarm and/or initiating action.

required volume of water).

6. Logic System Function Test - A logic system

2. Instrument Channel Calibration - An functional test means a test of relays and .

instrument channel calibration means the contacts of a logic circuit from sensor to adjustment of an instrument signal output so activated device to ensure components are that it corresponds, within acceptable operable per design intent. Where practi-range, and accuracy, to a known value(s) of cable, action will go to completion: i.e.,

the parameter which the instrument pumps monitors. Calibration shall encompass the entire instrument channel including actuation, alarm or trip.

AmendmentNo.f 2

JAFNPP 1.0 (cont'd) .

1. Refuel Mode - The reactor is in the refuel- system trips and control rod withdrattal mode when the Mode Switch is in the Refuel interlocks in service.

Mode position. When the Mode Switch is in the Refuel position, the refueling inter- J. Operable - A system, subsystem, train, component locks are in service, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its.specified

2. Run Mode - In this mode the reactor system function (s). Implicit in this definition shall pressure is at or above 850 psig and the be the assumption that'all necessary attendant Reactor Protection System is energized with instrumentation, controls, normal and emergency APRM protection (excluding the 15 percent electrical power sources, cooling or seal water, high flux trip) and the RBM interlocks in lubrication or other auxiliary equipment that are service. required for the system, subsystem, train, com-ponent or device to perform its function (s) are

3. Shutdown Mode - The reactor is in the shut- also capable of performing their related support down mode when the Reactor Mode Switch is in function (s).

the Shutdown Mode position.

K. Operating - Operating means that a system or

a. Hot shutdown means conditions as above component is performing its intended functions in with reactor coolant temperature > 212*F. Its required manner.

b. Cold shutdown means conditions as above L. Operatino Cycle - Interval between the end of one with reactor coolant tamperature refueling outage and tue end of the subsequent f-212*F. and the reactor vessel vented. refueling outage.

4. Startup/ Hot Standby - In this mode the reac- M. Primary Containment Integrity -

I tor protection scram trip initiated by main Primary containment integrity means that the steam line isolclion valve closure is by- drywell and pressure suppression chamber are passed when reactor pressure is less than intact and all of the following conditions are 1,005 psig, the low pressure main steam line satisfied

7 isolation valve closure trip is bypassed, the Reactor Protection System is energized with 1. All manual containment isolation valves on APRM (15 percent) and IRM neutron monitoring lines connected to the Reactor Coolant System or containment which are not required to be open during plant accident conditions are closed. These valves may be Amendment No. p$

4

JAFNPP l 1.0 (cont'd) opened to perform necessary operational is the period of time between the shutdown of the activities. unit prior to refueling and the startup of the l Plant subsequent to that refueling.

2. At least one door in each airlock is closed and sealed. R. Safety Limits - The safety limits are limits within which the reasonable maintenance of the

3. All automatic containment isolation valves fuel cladding integrity and the reactor coolant are operable or de-activated in the isolated system integrity are assured. Violation of such position. -a limit is cause for unit shutdown and review by the Atomic Energy Commission before resumption of

4. All blind flanges and manways are closed. unit operation. Operation beyond such a limit may not in itself result in serious consequences N. Rated Power - Rated power refers to operation at but it indicates an operational deficiency a reactor power of 2,436 MNt. This is also subject to regulatory review.

termed 100 percent power and is the maximum power level authorized by the operating license. Rated S. Secondary Containment Inteority - Secondary steam flow, rated coolant flow, rated nuclear containment integrity means that the reactor system pressure, refer to the values of these building is intact and the following conditions parameters when the reactor is at rated power. are met:

O. Reactor Power Operation - Reactor power operation 1. At least one door in each access opening is is any operation with the Mode Switch in the closed.

Startup/ Hot Standby or Run 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 Pressure - Unless otherwise valves are operable or secured in the indicated, reactor vessel pressures listed in the isolated position.

Technical Specifications are those measured by the reactor vessel steam space sensor. T. Surveillance Frecuenqy - Periodic Q. Refuelino Outage - Refueling outage Amendment No. J4 5

l JAFNPP surveillance tests, checks, calibrations, and V. Electrically Disarmed Control Rod examinations shall be performed within the specified surveillance intervals. These inter- To disarm a rod drive electrically, the four vals may be adjusted 25 percent. The interval amphenol type plug connectors are removed from as pertaining to instrument and electric surveil- the drive insert and withdrawal solenoids ren-lance shall never exceed one operating cycle. In dering the rod incapable of withdrawal. This cases where the elapsed interval has exceeded 100 procedure is equivalent to valving out the drive percent of the specified interval, the next sur- and is preferred. Electrical disarming does not veillance interval shall commence at the end of eliminate position indication.

the original specified interval.

W. High Pressure Water Fire Protection System U. T__h_qrmal Parameters The High Pressure Water Fire Protection System

1. Minimum critical power ratio (MCPR)-Ratio of consists of: a water source and pumps; and that power in a fuel assembly which is distribution system piping with associated post calculated to cause some point in that fuel indicator valves (isolation valves). Such valves assembly to experience boiling transition to include the yard hydrant curb valves and the the actual assembly operating power as first valve ahead of the water flow alarm device calculated by application of the GEXL on each sprinkler or water spray subsystem, correlation (Reference NEDE-10958).

X. Stacoered Test Basis

2. Fraction of Limiting Power Density - The ratio of the linear heat generation rate (LHGR) A Staggered Test Basis shall consist of:

existing at a given location to the design LHGR. The design LHGR is 14.4 KW/ft for a. A test schedule for "n" systems, subsystems, GE8x8EB fuel and 13.4 KW/ft for the remainder.

l trains or other designated components ob-tained by dividing the specified test

3. Maximum Fraction of Limiting Power Density - interval into "n" equal subintervals.

The Maximum Fraction of Limiting Power Density (MFLPD) is the highest value existing in the b. The testing of one system, subsystem, train core of the Fraction of Limiting Power Density or other designated component at the begin-(FLPD). ning of each subinterval.

4. Transition Boiling - Transition boiling means Y. Rated Recirculation Flow the boiling region between nucleate and film boiling. Transition boiling is the region in That drive flow which produces a core flow of which both nucleate and film boiling occur 77.0 x 106 lb/hr.

intermittently with neither type being completely stable.

Amendment No. d f , k, 7[ , 7[,

6

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

A.1.b. APRM Flux Scram Trip Settino (Refuel or Start & Hot Standby Mode)

APRM - The APRM flux scram setting shall be Jbl5 percent of rated neutron flux with the Re'.ctor Mode Switch in Startup/ Hot Standby or Refuel.

c. APRM Flux Scram Trio Settinos (Run Mode)

B. COLE Thermal Power (imit (Reactor Pressure M1785 Riigl (1) Flow Referenced Nestron Flux Scram Trip Setting When the reactor pressure is f$785 psig or core flow is less than or equal to 10% of rated, the When the Mode Switch is in the RUN l

core thermal power shall not exceed 25 percent of position, the APRM flow referenced flux rated thermal power. scram trip setting shall be:

C. Pawir Traqsient SdbO.66W & 54% for two loop operation or:

To ensure that the Safety Limit established in S fb(0.66 W + 54% - 0.66&W) for single Specification 1.1.A and 1.1.B is not exceeded, loop operation each required scram shall be initiated by its where:

expected scram signal. The Safety Limit shall be asstaaed to be exceeded when scram is accomplished S= Setting in percent of rated by a means other than the expected scram signal. thermal power (2436 MNT)

W= Recirculation flow in percent of rated 4LW= Difference between two loop and single loop effective drive flow at the same core flow. (2LW = 0 for two loop operation. 2L W for single loop operation is to be determined upon implementation of single loop operation.)

Amendn.ent No. J d, ,4

JAFNPP 1.2 and 2.2 BASES The reactor coolant pressure boundary integrity is ANSI Code permits pressure transients up to 20 percent-an important barrier in the prevention of uncon- over the design pressure (120% x 1,150 - 1,380 psig).

trolled release of fission products. It is The safety limit pressure of 1,375 psig is referenced essential that the integrity of this boundary be to the lowest elevation of the Reactor Coolant System.

protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The current reload analysis shows that the main steam The pressure safety limit of 1,325 psig as measured isolation valve closure transient, with flux scram, is by the vessel steam space pressure indicator is the most severe event resulting directly in a reactor equivalent to 1,375 psig at the lowest elevation of coolant system pressure increase. The reactor vessel the Reactor Coolant System. The 1,375 psig value pressure code limit of 1,375 psig, given in FSAR is derived from the design pressures of the reactor Section 4.2, is above the peak pressure produced by pressure vessel and reactor coolant system piping. the event above. Thus, the pressure safety limit The respective design pressures are 1250 psig at (1,375 psig) is well above the peak pressure that can 575*F for the reactor vessel, 1148 psig at 568'F result from reasonably expected overpressure tran-for the recirculation suction piping and 1274 psig sients. (See current reload analysis for the curve at 575* for the discharge piping. The pressure produced by this analysis.) Reactor pressure is safety limit was chosen as the lower of the continuously indicated in the control room during pressure transients permitted by the applicable operation.

design codes: 1965 ASME Boiler and Pressure Vessel Code,Section III for pressure vessel and 1969 ANSI A safety limit is applied to the Residual Heat B31.1 Code for the reactor coolant system piping. Removal System (RHRS) when it is operating in the The ASME Boiler and Pressure Vessel Code permits shutdown cooling mode. When operating in the shut-pressure transients up to 10 rercent over design down cooling mode, the RHRS is included in the pressure (110% x 1,250 - 1,375 psig) and the reactor coolant system.

The numerical distribution of safety / relief valve set-points shown in 2.2.1.B (2 @ 1090 psi, 2 @ 1105 psi, I 7 @ 1140 psi) is justified by analyses described in the General Electric report NEDO-24129-1, Supplement 1, and assures that the structural acceptance criteria set forth in the Mark I Containment Short Term Program are satisfied.

?

l Amendment No. , 4 l

29

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JAFNPP _

3.0 BASES A. This specification states the applicability of D. Continued each specification in terms of defined OPERATIONAL CONDITION (mode) and is provided to delineate The intent of this provision is to insure that specifically when each specification is facility operation is not initiated eith either applicable. required equipment or systems inoperable or other limits being exceeded.

B. This specification defines those conditions necessary to constitute compliance with the tenms Exceptions to this provision may be made for a of an individual Limiting Condition for Operation limited number of specifications when startuo and associated ACTION requirement. with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION C. This specification delineates the ACTION to be statements of the approprz. ate specifications, taken for circumstances not directly provided for in the ACTION statements and whose occurrence E. This specification delineates what additional would violate'the inte*,t of the specification. conditions must be satisfied to permit operation Under the terms of Sr ecification 3.0, the facility to continue,: consistent with the ACTION state-is to be placed in 0LD SHUTDOWN within the ments for power sources, when a normal or emer-following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> s. It is assumed that the unit gency power source is not OPERABLE. It specifi-is brought to the required OPERATIONAL CONDITION cally prohibits operation when one division is (mode) within the required times by promptly inoperable because its normal or emergency power l initiating and carrying out the appropriate source is inoperable and a system, subsystem, ACTION statement. train, component or device in another division is inoperable for another reason.

D. This specification provides that entry into an OPERABLE CONDITION (mode) must be made with (a) The provisions of this specification permit the the full complement of required systems, equip- ACTION statements associated with individual ment or components OPERABLE and (b) all other systems, subsystems, trains, components or parameters as specified in the Limiting Condi- devices to be consistent with the ACTION state-tions for Operation beims met without regard for ment of the associated electrical power source.

allowable deviations and out of service pro- It allows operation to be governed by the time visions contained in the ACTION statements.

Amendment No. ph 30b

l JAFNPP 3.1 BASES (cont'd) is discharged from the reactor by a scram can be The IRM high flux and APRM 6: 15% power scrams provide accommodated in the discharge piping. Each scram adequate coverage in the startup and intermediate discharge instrument volume accommodates in excess of range. Thus, the IRM and APRM systems are required 34 gallons of water and is the low point in the to be operable in the refuel and startup/ hot standby piping. No credit was taken for this volume in the modes. The APRM 6120% power and flow referenced design of the discharge piping as concerns the amount scrams provide required protection in the power range of water which must be accommodated during a scram. (referemce FSAR Section 7.5.7). The power range is covered only by the APRMs. Thus, the IRM system is During normal operation the discharge volume is not required in the run mode.

empty; however, should it fill with water, the water discharged to the piping from the reactor could not The high reactor pressure, high drywell pressure, be accommodated, which would result in slow scram reactor low water level and scram discharge volume times or partial control rod insertion. To preclude high level scrams are required for startup and run this occurrence, level detection instruments have modes of plant operation. They are, therefore, been provided in each instrument volume which alarm required to be operational for these modes of reactor and scram the reactor when the volume of water reaches operation.

34.5 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without The requirement to have the scram functions indicated impairment of the scram times or amount of insertion in Table 3.1-1 operable in the refuel mode assures of the control rods. This function shuts the reactor that shifting to the refuel mode during reactor power down while sufficient volume remains to accommodate operation does not diminish the protection provided the discharged water and precludes the situation in by the Reactor Protection System.

which a scram would be required but not be able to perform its function adequately. Turbine stop valve closure occurs at 10 percent of valve closure. Below 217 psig turbine first stage A Source Range Monitor (SRM) System is also provided pressure (30 percent of rated), the scram signal due to supply additional neutron level information during to turbine stop valve closure is bypassed because the startup but has no scram functions (reference para- flux and pressure scrams are adequate to protect the graph 7.5.4 FSAR). reactor.

Amendment No. Jb 34

JAFNPP 4.1 DASES (cont'd)

The bi-stable trip circuit which is a part of the The frequency of calibration of the APRM flow Group (B) devices can sustain unsafe failures biasing network has been established as each which are revealed only on test. Therefore, it refueling outage. The flow biasing network is is necessary to test them periodically. functionally tested at least once/ month and, in addition, cross calibration checks of the flow A study was conducted of the instrumentation input to the flow biasing network can be made channels included in the Group (B) devices to during the functional test by direct meter calculate their unsafe failure rates. The reading. There are several instruments which non-ATTS (Analog Transmitter Trip System) analog must be calibrated and it will take several days devices (sensors and amplifiers) are predicted to to perform the calibration of the entire network.

have an unsafe failure rate of less than 20x10-6 While the calibration is being performed, a zero failures /hr. The non-ATTS bi-stable trip flow signal will be sent to half of the APRM's circuits are predicted to have unsafe failure resulting in a half scram and rol block condition.

rate of less than 2x10-6 failures /hr. The ATTS Thus, if the calibration were performed during analog devices (sensors), bi-stable devices operation, flux shaping would not be possible.

(master and slave trip units) and power supplies Based on experience at other generating stations, have been evaluated for reliability by Mean Time drift of instruments, such as those in the flow Between Failure analysis or state-of-the-art biasing network, is not significant and therefore, qualification type testing meeting the to avoid spurious scrams, a calibration frequency requirements of IEEE 323-1974. Considering the of each refueling outage is established.

2-hour monitoring interval for analog devices as assumed above, the instrument checks and Group (C) devices are active only during a given functional tests as well as the analyses and/or portion of the operational cycle. For example, qualification type testing of the devices, the the IRM is active during startup and inactive design reliability goal for system reliability of during full-power operation. Thus, the only test 0.9999 will be attained with ample margin. that The bi-stable devices are monitored during plant operation to record their failure history and establish a test interval using the curve of Figure 4.1-1. There are numerous identical bi-stable devices used throughout thr; Plant's instrumentation system. Therefore, significant data on the failure rates for the bi-stable devices should be accumulated rapidly.

Amendment No. ,44, 38 I

JAFNPP TABLE 3.1-1 -~

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No. Modes in Which Total Number of Operable Trip Level Function Must be of Instrument Instrument Trip Function Setting Operable Channels Pro . Action-Channels vided by Design (1) per Trip Refuel Startup Run for Both Trip System (1) (6) Systems (16) 1 Mode Switch in X X X 1 Mode Switch A Shutdown (4 Sections) 1 Manual Scram X X X 2 Instrument A Channels 3 IRM High Flux bl20/125of X X 8 Instrument A full scale Channels 3 IRM Inoperative X X 8 Instrument A Channels 2 APRM Neutron Flux- S15% Power X X 6 Instrument A Startup(15) Channels 2 APRM Flow Referenced Sd(0.66W54%)(FFP/MFLPD) X 6 Instrument A or B Neutron Flux (Not to Channels exceed 117%) (12)(13)

(14)(17) 2 APRM Fixed High d120% Power X 6 Instrument A or B Neutron Flux (14) Channels 2 APRM Inoperative (10) X X X 6 Instrument A or B Channels Amendment No. }h,3,b,4$,7 , 8 , 9 41

JAFNPP LABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REOUIREMENT Minimum No. Modes in Which Total of Operable Trip Level Function Must be Number of Instrument Trip Function Setting Operable Instrument Action Channels Channels (1) per Trip Refuel Startup Run Provided System (1) (6) by Design (16) for Both Trip Systems 2 Turbine Control 500(P(850 psig X(4) 4 Instrument A or C Valve Fast Closure Control oil pressure Channels between fast closure solenoid and disc dump valve Amendment No. p[

41b

__ Q

JAFNPP IABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No. Modes in Which Total-of Operable Trip Level Function Must be Number of Instrument Trip Function Setting Operable Instrument Action Channels Channels (1) per Trip Refuel Startup Run Provided System (1) (6) )16) by Design for Both Trip Systems l 4 Turbine Stop 6 10% valve X(4)(5) 8 Instrument A or C Valve Closure closure Chennels NOTES OF TABLE 3.1-1

1. There shall be two operable or tripped trip systems for each function, except as specified 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 shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.

A. Initiate insertion of operable rods and complete insertion 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 30 percent of rated.

2. Permissible to bypass, if Refuel and Shutdown positions of the Reactor Mode Switch.

3. By passed when reactor pressure is less than 1005 psig.

4. Bypassed when turbine first stage pressure is less than 217 psig or less than 30 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. ,

42

JAFNPP 3.2 BASES (cont'd)

High radiation monitors in the main steam line tunnel The trip settings of f=300 percent of design flow for have been provided to detect gross fuel failure as in high flow or 40*F above maximum ambient for high l the control rod drop accident. With the established temperature are such that uncovering the core is pre-setting of 3 times normal background, and main steam vented and fission product release is within limits.

line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not The RCIC high flow and temperature instrumentation exceeded for this accident. Reference Section are arranged the same as that for the HPCI. The trip 14.6.1.2 FSAR. During the Hydrogen Addition Test, settings of f-300 percent for high flow or 40*F above l the normal background Main Steam Line Radiation Level maximum ambient for temperature are based on the same is expected to increase by approximately a facter of criteria as the HPCI.

5 at the peak hydrogen concentration as indicated in note 16, Table 3.1-1. With the hydrogen addition, The reactor water cleanup system high temperature the fission product release would still be well l instrumentation are arranged similar to that for the within the 10 CFR 100 guidelines in the event of a HPCI. The trip settings are such that uncovering the control rod drop accident. core is prevented and fission product release is within limits.

Pressure instrumentation is provided to close the main steam isolation valves in the run mode when the main The instrumentation which initiates ECCS action is l steam line pressure drops below 625 psig. The reactor arranged in a dnal bus system. As for other vital pressure vessel thermal transient due to an inadver- instrumentation erranged in this fashion, the speci-tent cpening of the turbine bypass valves when not in fication preserves the effectiveness of the system the run mode is less severe than the loss of feed- even during periods when maintenance or testing is water analyzed in Section 14.5 of the FSAR, therefore, being performed. An exception to this is when logic closure of the main steam isolation valves for thermal functional testing is being performed.

transient protection when not in the run mode is not required. The control rod block functions are provided to pre-I vent excessive control rod withdrawal so that MCPR The HPCI high flow and temperature instrumentation are does not decrease to the Safety Limit. The trip provided to detect a break in the HPCI steam piping.

Tripping of this instrumentation results in actuation of HPCl isolation valves. Tripping logic for the high flow is a 1 out of 2 logic.

1 Amendment No. J[, ,

, p8' 57

JAFNPP 3.2 BASES (cont'd) logic for this function is 1 out of n: e.g., any trip The scaling arrangement is such that trip setting is on one of six APRM's, eight IRM's, or four SRM's will less than a f actor of 10 above the indicated level.

result in a rod block.

A downscale indication on an APRM or IRM is an indi-The minimum instrument channel requirements assure cation the instrument has failed or the instrument is sufficient instrumentation to assure the single not sensitive enough. In either case the instrument failure criteria is met. The minimum instrument wil? not respond to changes in control rod motion and channel requirements for the RBM may be reduced by- tnus, control rod motion is prevented. The downscale one for maintenance, testing, or calibration. This trips are set at 2.5 indicated on scale.

time period is only three percent of the operating time in a month and does not significantly increase The flow comparator and scram discharge volume high the risk of preventing an inadvertent control rod level components have only one logic channel and arc l withdrawal. not required for safety. The flow comparator must be bypassed when operating with one recirculation water The APRM provides gross core protection; i.e., limits pump.

the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The refueling interlocks also operate one logic channel, and are required for safety only when the The RBM rod block function provides local protection Mode Switch is in the Refueling position.

of the core; i.e., the prevention of boiling transi-tion in a local region of the core, for a single rod For effective emergency core cooling for small pipe withdrawal error from a limiting control rod pattern. breaks, the HPCI system must function since reactor The trips are set so that MCPR is maintained greater pressure does not decrease rapidly enough to allow than the Safety Limit. either core spray or LPCI to operate in time. The Automatic pressure relief function is provided as a The IRM rod block function provides local as well as backup to the HPCI in the event the HPCI does not gross core protection. operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in AmendmentNo.jh 58 n

Ngr i

JAFNPP 4.2 HASES The instrumentation listed in Table 4.2-1 through To test the trip relays requires that the channel be 4.2-6 will be functionally tested and calibrated at bypassed, the test made, and the system returned to regularly scheduled intervals. The same design re- its initial state. It is assumed this task requires liability goal as the Reactor Protection System is an estimated 30 min. to complete in a thorough and generally applied. Sensors, trip devices and power workmanlike manner and that the relays have a failure supplies are tested, calibrated and checked at the rate of 10-6 failures per hr. Using this data and game frequency as comparable devices in the Reactor the above operation, the optimum test interval is:

Protection System.

i= 2(0.51 - 1 x 103 hr.

Those instruments which, when tripped, result in a 10-6 = 40 days rod block have their contacts arranged in a 1 out of n logic, and all are capable of being bypassed. For For additional maroin a test interval of once/ month such a tripping arrangement with bypass capability will be used initially.

provided, there is an optimum test interval that should be maintained in order to maximize the relia- The sensors and electronic apparatus have not been bility of a given channel (7). This takes account of included here as these are analog devices with read-the fact that testing degrades reliability and the outs in the control room and the sensors and elec-optimum interval between tests is approximately given tronic apparatus can be checked by comparison ~with by: other like instrumet's. The checks which are made on a daily basis are adequate to assure operability of i= 2t the sensors and electronic apparatus, and the test r interval given above provides for optimum testing of the relay circuits.

Where: i= the optimum interval between tests.

t= the time the trip contacts are disabled from performing their function while the test is in progress.

r= the expected failure rate of the relays.

Amendment No. ,pd 61

JAFNPP TABLE 3.2-2 (cont'd)

INSTRUME]EATION THAT INITTATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No. Total of Operable Number of Instru-Instrument ment' Channels Pro-Item Channels Per vided by Design for No. Trio System (1) Trio Furfion Trip Level SettinP;> 50 psig 2 Inst. Channels Close Isolation valves

Low Pressure in RCIC Subsystem 31 1 HPCI Turbine Steam fh106 in H 2O dp 2 Inst. Channels Close Isolation Valves j Line High Flow in HPCI Subsystem J

q 32 1 RCIC Turbine High fh10 psig 2 Inst. Channels Close Isolation Valves Exhaust Diaphragm in RCIC Subsystem Pressure 4

i 33 1 HPCI Turbine High fh10 psig 2 Inst. Channels Close Isolation Valves i Exhaust Diaphragm in HPCI Subsystem Pressure 34 1 LPCI Cross-Connect NA 1 Inst. Channels Initiates annunciation Position when valve is not closed 35 1 HPCI Steam Line 100)>P)> 50 psig 2 Inst. Channels- Close Isolation Valve Low Pressure. in HPCI Subsystems 36 1 HPCI Steam Line/ dk40*F 2 Inst. Channels Close Isolation Valve Area Temperature above mar. ambient in HPCI Subsystem Amendment No. , [8, 70b

NOTES FOR TABLE 3.2-6 (CONTINUED)

2. In the event that all indications of this parameter is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.

3. Three (3) indicators from level instrument channel A, B, & C. Channel A or B are utilized for feedwater control, reactor water high and low level alarms, recirculation pump runback. High level trip of main turbine and feedwater pump turbine utilizes channel A, B, & C.

4. One (1) recorder utilized the same level instrument channel as selected for feedwater control.

5. Three (3) indicators from reactor pressure instrument channel A, B, & C. Channel A or B are utilized for feedwater control and reactor pressure high alarm.

6. One (1) recorder. Utilizes the same reactor pressure instrument channel as selected for feedwater control.

7. The position of each of the 137 control rods is monitored by the Rod Position Information System. For control rods in which the position is unknoin, refer to Paragraph 3.3.A.

8. Neutron monitoring operability requirements are specified by Table 3.1-1 and Paragraph 3.3.B.4.

9. A minimum of 3 IRM or 2 APRM channels respectively must be operable (or tripped) in each safety system.

10. Each Safety Relief Valve is equipped with two acoustical detectors of which one is in service and a backup thermocouple detector. In the event that a thermocouple is inoperable SRV performance shall be monitored daily with the associated acoustical detector.

11. From and after the date that none of the acoustical detectors is operable but the thermocouple is operable, continued operation is permissible until the next outage in which a primary containment entry is made. Both acoustical detectors shall be made operable prior to restart.

12. In the event that both primary and secondary indications of this parameter for any one valve are disabled l and neither indication can be restored in forty-eight (48) hours, an orderly shutdown shall be initiated -l and the reactor shall be in a Hot Shutdown condition in twelve (12) hours and in a Cold Shutdown within the next twenty-four (24) hours.

13. From and after the date that the minimum number of operable instrument channels is one less than the

~

l l minimum number specified for each parameter, continued operation is permissible during the succeeding 7 days unless the minimum number specified is made operable sooner.

Amendment No. , , 6,7 76c

JAFNPP TABLE 3.2-7 INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIE

~

Minimum Number of Total Number of I-=trument Operable Instrument Channels Provided by Channels per trip Decign for Both System (1) Instrument Trio Level Settina Channels Action 1 Reactor High Pressure d$1120 psig 4' (2) 1 Reactor Low-Low Db-38 in, indicated 4 (2)

Water Level 17veh126.5in.above thetopofactivefuel)

Notes for Table 3.2-7

1. Whenever the reactor is in the run mode, there shall be one operable trip system for each parameter for each operating recirculation pump. From and after the time it is found that this cannot be met, the indicated action shall be taken.

2. Reduce power and place the Mode Selector Switch in a Mode other than the Run Mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No.df ,fr8 77

\

JAFNPP 3.3 (cont *d) 4.3 (cont'd)

b. The control rod directional control e. When it is initially determined that a valves for inoperable control rods control rod is incapable of normal shall be disarmed electrically. insertion, an attempt to fully insert the control rod shall be made. If the i

c. Control rods with scram times greater control rod cannot be fully inserted, a than those permitted by Specification shutdown margin test shall be made to 3.3.C.3 are inoperable, but if they can demonstrate under this condition that be inserted with control rod drive the core can be made subcritical for any pressure they need not be disarmed reactivity condition during the electrically. remainder of the operating cycle with the analytically determined, highest

d. Control rods with a failed "Full-in" or worth control rod capable of with-

"Full-out" position switch may be drawal, fully withdrawn, and all other bypassed in the Rod Sequence Control control rods capable of insertion fully System and considered operable if the inserted. If Specification 3.3.A.1 and actual rod position is known. These 4.3.A.1 are met, reactor startup may rods must be moved ir. sequence to their proceed.

correct positions (full in on insertion and full out on withdrawal). f. The scram discharge volume drain and vent valves shall be full-travel cycled

e. Control rods with inoperable at least once per quarter to verify that accumulators or those whose position the valves close in less than 30 seconds cannot be positively determined shall and to assure proper valve stroke and be inoperable. operation.

f. Inoperable control rods shall be 9 At least once per operating cycle, the positioned such that Specification operability of the entire scram 3.3.A.1 is met. In addition, during discharge system as an integrated whole reator: power operation, no more than shall be demonstrated by a scram of one control rod in any 5 X 5 array may control rods from a normal control rod be inoperable (at least 4 operable configuration of less than or equal to control rods must separate any 2 50% rod density by verifying that the inoperable ones). If this specifi- drain and vent valves:

cation O nest be met the reactor shall not be started, or if at power, the 1. Close upon receipt of a signal for reactor shall be brought to a cold control rods to scram and:

condition within 24 hr.

Amendment No. J$I, f2$ J 6' 89a

JAFNPP ,

. .a This requirement may be satisfied as part of'any scram originating from the rod density conditions specified above, provided that Specification.

4.3.A.2.f is independently satisfied during the quarter in which the scram occurs.

i Amendmcnt No. g i

90

- _ - - _ - - . _ _ . -w e- - - - . -- + - v..--. . ,e- , , , , M

JAFNPP 3.3 (cont'd) 4.3 (cont *d)

B. Control Rods B. Control Rods a

1. Each control rod shall be cor. pled to 1. The coupling integrity shall be l its drive or completely inserted and verified for each withdrawn control rod the control rod directional control as follows:

valves disarmed electrically. This requirement does not apply in the a. When a rod is withdrawn the first refuel condition when the reactor is time after each refueling outage vented. Two control rod drives may be or after maintenance, observe removed as long as Specification discernible response of the 3.3.A.1 is met. -

nuclear instrumentation. Howeier, for initial. rods when response is not discernible, subsequent exercising of these rods after the reactor is above 20 percent power shall be performed to verify instrumentation response.

i b. When the rod is fully withdrawn the first time after each refueling outage or after maintenance, observe that the drive does not go to the overtravel position.

c. During each refueling outage and-after each control rod maintenance, observe that the drive does not go to the overtravel position.

2. The control rod drive housing support 2. The control rod drive housing support system shall be in place during reactor system shall be inspected after power operation or reassembly and the results of the inspection recorded.

1

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l Amendment No.jd 91 '

o , e - --

JAFNPP 3.3 and 4.3 BASES (cont'd)

At power levels below 20% of rated, abnormal rod drop accident consequences are acceptable.

control rod patterns could produce rod worths Control rod pattern constraints above 20% of high enough to be of concern relative to the 280 rated power are imposed by power distribution calories per gram drop limit. In this range, the requirements as defined in Section 3.3.3.5 of R'JM and RSCS constrain the control rod sequence these Technical Specifications. Power level for and patterns to those which involve only automatic cutout of the RSCS function is sensed acceptable rod worths. by first stage turbine pressure. Because the instrument has an instrument error of 1 2% of The Rod Worth Minimizer and the Rod Sequence full power, the nominal instrument setting is 22%

Control System provide automatic supervision to of rated power. Power level for automatic cutout assure that out-of-sequence control rods will not of the RWM function is sensed by steam flow and l be withdrawn or Inserted; i.e., it limits is set manually at 30% of rated power to be operator deviance from planned withdrawal consistent with the RCSC setting.

sequences. They serve as a backup to procedural control of control rod sequences which limit the Functional testing of the RWM prior to the start maximal reactivity worth of control rods, in the of control rod withdrawal at startup, and prior event that the Rod Worth Minimizer is out of to attaining 20% rated thermal power during rod service, when required, a second licensed insertion while shutting down, will ensure operator or other qualified technical plant reliable operation and minimize the probability employee of the rod drop accident.

can manually fulfill the control rod pattern The RSCS can be functionally tested prior to conformance functions of this system. In this control rod withdrawal for reactor startup. By case, the RSCS is backed up by independent selecti ng, for example, A12 and attempting to procedural control to assure conformance. withdraw, by one notch, a rod or all rods in each other group, it can be detern.Ined that the A12 The functions of the RWM and RSCS make it group is exclusive. By bypassing to full-out all unnecessary to specify a license limit on rod A 12 rods, selecting A34 and attempting to worth to preclude unacceptable consequences in withdraw, by one notch, a rod or all rods in the event of a control rod drop. At low powers, group B, the A 34 group is determined below 20%, these devices force adherence to exclusive. The same procedure can be repeated acceptable rod patterns. Above 20% of rated for the B groups. After 50% of the control power, no constraint on rod pattern is required to assure that Amendment No. J<f, p3' 101

JAFNPP 4.4 (Cont'd) pump solution in the recirculation path.

Explode one of three primer assemblies manufactured in same batch to verify proper function. Then install the two remaining-primer assemblies of the same batch in the explosive valves.

Demineralized water shall be injected into the reactor vessel to test that valves (except explosive valves) not checked by the recirculation test are not clogged.

Test that the setting of the system pressure relief valves is between 1,400 and 1,490 psig.

f 3. Disassemble and inspect one explosive valve I

so that it can be established that the valve is not clogged. Both valves shall be inspected in the course of two operating f l

l cycles.

l B. Operation with Inoperable Components B. Operation with Inoperable Components Fr > :. 2d af ter the date that a redundant When a component becomes inoperable its redundant co 3,onent is made or found to be inoperable, component shall be demonstrated to be operable j I

Specification 3.4.A shall be considered immediately and daily thereafter.

fulfilled, and continued operation permitted, provided that:

1. The component is returned to an operable condition within 7 days. ,

Amendment No.,)[

106

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

2. From and after the date that one of the Core Spray 2. When it is determined that one Core Spray Systew Systems is made or found inoperable for anny is inoperable, the operable Core Spray System, reason, continued reactor operation is permissible and the LPCI System, shall be demonstrated to be during the succeeding 7 days unless the system is operable immediately. The remaining Core Spray made operable earlier, provided that during the 7 System shall be demonstrated to be operable daily days all active components of the other Core Spray thereafter.

l System and the LPCI System shall be operable.

3. The LPCI mode of the RHR System shall be operable 3. LPCI System testing shall be as specified in whenever irradiated fuel is in the reactor and 4.5.A.1.a, b, c, d, f and g except that three RHR prior to reactor startup from a cold condition, pumps shall deliver at least 23,100 gpa against a except as specified below. system head corresponding to a reactor vessel pressure of 20 psig.

a. From the time that one of the RHR pumps is made or found to be inoperable for any a. When it is determined that one of the RHR reason, continued reactor operation is pumps is inoperable, the remaining active permissible during the succeeding 7 days components of the LPCI, containment spray unless the pump is made operable earlier sui,3ystem and both Core Spray Systems provided that during such 7 days the required for operation shall be demonstrated remaining active components of the LPCI, to be operable immediately, and the containment spray mode, and all active remaining RHR pumps shall be demonstrated to components of both Core Spray Systems are be operable daily thereafter.

operable.

I 1

l l

l 1

1 1

Amendment No. ) [ fd' g l 114 l

l L______ - ______ _ .

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

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5. All recirculation pump discharge valves 5. All recirculation pump _ discharge valves shall be operable prior to reactor startup shall be tested for operability any time the (or closed if permitted elsewhere in these reactor is in the cold condition exceeding specifications). 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have not been performed during the preceding 31 days.

6. If the requirements of 3.5.A cannot be met, the reactor shall be placed in the cold condition within 24 hrs.

B. Cgnt;phnt Coolino Subsystem Mode (of the PHg B. Containment Coolino Subsystem Mode (of the PHR System) System)

1. Both subsystems of the containment cooling 1. Subsystems of the containment cooling mode mode, each including two RHR, one ESW pump are tested in conjunction with the test and two RHRSW pumps shall be operable when- performed on the LPCI System and given in ever there is irradiated fuel in tne reactor 4.5.A.I.a, b, c, and d. Residual heat vessel, prior to startup from a cold removal service water pumps, each loop condition, and reactor coolant temperature consisting of two pumps operating in Eh 212*F except as specified below: parallel, will be included in testing, supplying 8,000 gpm. The Emergency Service Water System, each loop of which consists of a single operating emergency service water pump will be tested in accordance with l Section 4.llD.

During each five-year period, an air test shall be performed on the containment spray headers and nozzles.

2. Continued reactor operation is permissible 2. When it is determined that one RER pump for 30 days with one spray loop inoperable and/or one RHRSW pump of the components and with reactor water temperature greater required in 3.5.B.1 above are inoperable, than 212*F. the remaining redundant active components of the containment cooling mode subsystems shall be demonstrated to be operable immediately and daily thereafter.

Amendment No.dJ , %

115a

1 4

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

a. From and after the date that the HPCI System a. When it is determined that the HPCI' subsystem is made or found to be inoperable for any is inoperable the RCIC, the LPCI subsystem, j reason, continued reactor operation is per- both core spray subsystems, and the ADS-4 missible only during the succeeding 7 days

~

subsystem actuation logic shall be unless such system is sooner made operable, -demonstrated to be operable immediately.

l provided that during such 7 days all active The RCIC system and ADS subsystem logic l components of the Automatic Depressurization shall be demonstrated to be operable daily

! System, the Core Spray System, LPCI System, thereafter.

f and Reactor Core Isolation Cooling System are operable.

l b. If the requirements of 3.5.C.1 cannot be met, the reactor shall be placed in the cold condition and pressure less than 150 psig within 24 hrs.

4 2. Low power physics testing and reactor operator

training shall be permitted with reactor coolant j temperature f(212*F with an inoperable com-

! pcnent(s) as specified in 3.5.C.1 above.

l a

4 Amendment No. f[, 1,ph 118

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

D. Automatic Depressurization System (ADS) D. Automatic Depressurization System (ADS)

1. The ADS shall be operable whenever the reac- 1. Surveillance of the Automatic Depressuriza-tor pressure is greater than 100 psig, and tion System shall be performed during each.

1 irradiated fuel is in the reactor vessel and operating cycle as follows:

l prior to reactor startup from a cold condi-tion, except as specified below: a. A simulated automatic initiation which opens all pilot valves.

a. From and after the date that one of the l seven safety / relief valves of the ADS is b. Manually open each safety / relief valve made or found to be inoperable for any ~ while bypassing steam to the condenser reason while it is required, continued and observe a JL10% closure of the turbine reactor operation is permissible only bypass valves, to verify that the safety /

during the succeeding 30 days unless relief valve has opened.

repairs are made and provided that during such time the HPCI System is operable. c. A simulated automatic initiation which is inhibited by the override switches.

b. From the time that more than one of the l seven safety / relief valves of the ADS are made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding 24 hrs.

unless repairs are made and providea, that

/

AmendmentNo.)(,S4 p 119

JAFNPP 3.5 (Cont'd) 4.5 (Cont'd) l F. Minimum _Emeroency Core and Containment Coolino F. Minimum Emeroency Core and Containment Coolina System Availability System Availability

1. Any combination of inoperable components in Not Applicable.

the Core and Containment Cooling Systems shall not defeat the capability of the remaining operable components to fulfill the core and containment cooling functions.

2. When the irradiated fuel is in the reactor vessel and the reactor is in the cold condition all LPCI, core spray, and containment cooling subsystems may be inoperable provided no work is being done which has the potential for draining the reactor vessel.

G. Maintenance of Filled Dischnige Pipe G. Maintenance of Filled Discharoe Pigg Whenever core spray subsystems, LPCI sub- The following surveillance requirements shall be systems, HPCI, or RCIC are required to be adhered to, in order to assure that the discharge operable, the discharge piping from the pump piping of the core spray subsystem, LPCI subsys-discharge of these systems to the last block tem, HPCI, and RCIC are filled:

valve shall be filled.

1. Every month prior to the testing of the LPCI

a. From and after the time that the pump subsystem and core spray subsystem, the discharge piping of the HPCI, RCIC, LPCI, or discharge piping of these systems shall be Core Spray Systems cannot be maintained in a vented from the high point, and water flow filled observed.

Amendment No.jA '

122

JAFNPP ,

3.5 (cont *d) 4.5 (cont'd) condition, that pump shall be considered inoper- 2. Following any period where the LPCI subsys-able for purposes satisfying Specifications tems or core spray subsystems have not been 3.5.A, 3.5.C, and 3.5.E. required to be operable, the discharge piping of the inoperable system shall be H. Avert e Planar Linear Heat Generation Rate vented from the high point prior to the IAPLHGR) return of the system to service.

During power operation, the APLHGR for each type 3. Whenever the HPCI, RCIC, or Core Spray System of fuel as a function of axlal location and is lined up to take suction from the conden-average planar exposure shall be within limits sate storage tank, the discharge piping of based on applicable APLHGR lirait values which the HPCI, RCIC, and Core Spray shall be have been approved for the r espective fuel and vented from the high point of the system, lattice types. Whes hand calculations are and water flow observed on a monthly basis.

required, the APLHGR for each type of fuel as a function of average planar exposure shall not 4. The level switches located on the Core Spray exceed the limiting value for the most limiting and RHR System discharge piping high points lattice (excluding natural uranium) shown in which monitor these lines to insure they are Figures 3.5-10 through 3.5-12 during two full shall be functionally tested each month.

recirculation loop operation. During single loop operation, the APLHGR for each fuel type shall H. Average Planar Linear Heat Generation Rate not exceed the above values multiplied by 0.84 (APLHGR)

(see Bases 3.5.K, Reference 1). If at anytime l during reactor power operation greater than 25% The APLHGR for each type of fuel as a function of of rated power it is determined that the limiting average planar exposure shall be determined daily value for APLHGR is being exceeded, action shall during reactor operation at g 25% rated thermal then be initiated within 15 minutes to restore power.

operation to within the prescribed limits. If the APLHGR is not returned to within the prescribed limits within two (2) hours, an orderly reactor power reduction shall be commenced immediately. The reactor power shal?

be reduced to less than 25% of rated power within the next four hours, or until the APLHGR is returned to within the prescribed limits.

AmendmentNo.pd,p, [

8 9[,

123

==

. ~s m y .

. ;% ,. i JAFNPP -

3.6 (cont'd) .4.6 (cont *d). .]

m E. Safety and Safetv/ Relief Valves E. Safety and Safetv/ Relief Valves

1. During reactor power operating conditions and 1. At least one half of all safety / relief valves.

~

prior to startup from a cold condition, or shall be bench checked or replaced with bench ~

whenever reactor coolant pressure is greater checked valves'once each operating cycle. - The than atmosphere and temperature greater than safety / relief valve' settings shall be set as 212*F, the safety mode of all safety / relief required in Specification 2.2.B.-'All valves valves shall be operable, except as specified shall be tested every two ' operating cycles. :

by Specification 3.6.E.2. The Automatic De-pressurization System valves shall be operable as required by specification 3.5.D.

i i

1 4

Amendment No. [ , % % [

142a 1 -

. _ _ _ _ _ _ _ _ _ _ _ - . _ . - __ __ _ , . . . , _ . . . . . . , _ , ,. ._ . . _ . , _ ._m.. .

6 1

7his page intentionally blank.

1

\

4 142b Amendment No %

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

2. a. From and after the date that the safety valve 2. At least one safety / relief valve shall be disas-function of one safety / relief valve is made sembled and inspected once/ operating cycle.

or found to be inoperable, continued opera-tion is permissible only during the succeeding 3. The integrity of the safety / relief valve ballows 30 days unless such valve is sooner made shall be continuously monitored.

operable.

a. The bellows monitoring pressure switches

b. From and after the time that the safety valve shall be removed and bench checked once/

function on two safety / relief valves is made operating cycle. Modified safety / relief or found to be inoperable, continued reactor valves with two-stage assemblies do not have operation is permissible only during the a bellows arrangement and are, therefore, not succeeding 7 days unless such valves are subject to this requirement.

sooner made operable.

4. The integrity of the nitrogen system and compon- .j

~

3. If Specification 3.6.E.1 and 3.6.E.2 are not met, ents which provide manual and ADS actuation of j the reactor shall be placed in a cold condition the safety / relief valves shall be demonstrated at within 24 hr. least once every 3 months.

4. Low power physics testing and reactor operator ,

training shall be permitted with inoperable components as specified in Item B.2 above, provided that reactor coolant temperature is fk212*F and the reactor vessel is vented or the reactor vessel head is removed.

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l l

Amendment No.I f ,J[

143

l i 1 This Page Intentionally Blank 143a AmendmentNo.[

This Page Intentionally Blank l

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l Amendment No, Jtf

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

F. Structural Integrity F. Structural Intecrity-The structural integrity of the Reactor Coolant 1. Nondestructive inspections shall be performed System shall be maintained at the level required on the ASME Boller and Pressure Vessel Code:

by the original acceptance standards throughout Class 1, 2 and 3 components and supports in the life of the Plant. accordance with the requirements of the weld and support inservice inspection program.

This inservice inspection program is based on an NRC approved edition of, and addenda to,-

Section XI of the ASME Boller and Pressure.

Vessel Code which is in effect 12 months or.

less prior to the beginning of the inspection interval.

2. An augmented inservice inspection program is required for those high stressed circum-ferential piping joints in the main steam and feedwater lines larger than 4 inches in diameter, where no restraint against pipe whip is provided. The augmented inservice inspection program shall consist of 100 percent inspection of these welds per inspection interval.

G. Jet Pumps G. Jet Pumps-Whenever the reactor is in the startup/ hot Whenever there is recirculation flow ~with the standby or run modes, all jet pumps shall be reactor in the startup/ hot standby or run modes, operable. If it is determined that a jet pump is jet pump operability shall be checked daily by l inoperable, the reactor shall be placed in a cold verifying that the-following conditions do not condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. occur simultaneously:

Amendment No.Jhd 144 1

-l 1

JAFNPP 3.6 and 4.6 BASES (cont'd)

E. EDfety and Safetv/ Relief Valves The modified version of the safety / relief valves function with a direct-acting pilot arrangement with Experiences in safety valve operation show that no integral bellows.

the testing of 50 percent of the safety valves por refueling outage is adequate to detect It is realized that there is no way to repair or failures or deterioration. The tolerance value replace the bellows during operation, and the plant is specified in Section III of the ASME Boiler must be shutdown to do.this. The 30-day and 7-day and Pressure Vessel Code an 2 1 percent of periods to do this allow the operator flexibility to design pressure. An analysis has been performed choose his time for shutdown; meanwhile, because of which shows that with all safety valves set 1 the redundancy present in the design and the con-percent higher, the reactor coolant pressure tinuing monitoring of the integrity of the other safety limit of 1,375 psig is not exceeded. valves, the overpressure pressure protection has not been compromised in either case. The auto-relief The safety / relief valves have two functions; function would not be impaired by a failure of the l

i.e., power relief or self-actuated by high bellows. However, the self-actuated overpressure pressure. Power relief is a solenoid actuated safety function would be impaired by such a failure, function (Automatic Depressurization System) in There is no provision for testing the bellows leakage which external instrumentation signals of pressure switch during plant operation. The bellows low-low-low water level initiate the valves to leakage pressure switches will be removed and bench open. This function is discussed in checked once/ operating cycle. These bench checks specification B.3.5.D. In addition, the valves provide adequate assurance of bellows integrity. For can be operated manually. those modified safety / relief valves with the direct-acting pilot arrangement, bellows failures and The safety function is performed by the same bellows related calibrations do not apply.

l safety / relief valve with self-actuated integral bellows and pilot valve causing main valve Low power physics testing and reactor operator operation. Article 9 of the ASME Pressure training with inoperable components will be conducted Vessel Code Section III-Nuclear Vessels, only when the safety / relief and safety valves are l requires that these bellows be monitored for failure, since this would defeat the safety function of the safety / relief valve.

Amendment No. ,+T' 152

1 JAFNPP 3.6 and 4.6 BASES (cont'd) not required to be operable (reactor coolant Several locations on the main steam lines and temperature less than or equal to 212*F and the

] reactor vessel vented or the reactor vessel head feedwater lines are not restrained to prevent pipe whip in the event of pipe failure at these locations.

removed). Permitting physics testing and The physical layout within the drywell precludes operator training under these conditions would restraints at these points. Unrestrained high stress not place the plant in an unsafe condition. areas have been identified in these lines where breaks could result in pipe whip such that the pipe could-F. Structural Integrity impact the primary containment wall. Augmented inservice inspection of these weld locations shall be A pre-service inspection of the ASME Code Class 1 performed during each inspection period.

components was performed after site erection to assure the system was free of gross defects. An In addition, visual inspection in accordance with the initial inspection program as detailed in approved ASME code will be made during periodic Appendix F of the FSAR was developed and based on pressure and hydrostatic tests of critical systems.

an approved edition of the ASME Code. The inspection program specified encompasses the major areas of the vessel and piping system within The program has been expanded to include the the drywell. The inspection period is based on the requirements of later, approved ASME Code observed rate of defect growth from fatigue studies editions and addenda as far as practicable. The sponsored by the AEC.

Importance of these inspections is recognized, and efforts to develop practical new alternative These studies show that thousand of stress cycles, at methods of assuring plant inservice integrity stresses beyond any expected to occur in a Reactor will continue. This inspection program should Coolant System, were required to propagate a crack.

assure the detection of problem areas well before The test they represent a significant impact on safety.

Amendment TC.,p6' 153

JAFNPP

4. 7 (cont'd)

c. Type C tests (1.) Type C tests shall be performed by local pressurization. The pressure shall be applied in the same direction as that when the valve would be required to perform its safety function, except as listed in Table 4.7-2 unless it can be determined that l the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results. Each valve to be tested shall be closed by normal operation and without any preliminary exercising or adjustments.

(2.) Valves, unless pressurized with fluid from a seal system, shall be pressurized with air or nitrogen at a pressure of Pa, and the gas flow to maintain Pa shall be measured.

(3.) Valves, which are sealed with fluid from a seal system, such as the liquid in the suppression chamber shall not be tested.

Amendment No. pd 171

JAFNPP 4.7 (cont'd) w (4.) See table 4.7-2 for exceptions.

l' (S.) Acceptance criterion - The combined leakage rate for-all penetrations and valves subject to type B and C tests shall be less than 0.60 La. Leakage from containment isolation valves that are sealed with fluid from a seal system may be excluded when determining the, combined leakage rate provided that-the installed isclation valve seal-water system fluid inventory is sufficient to assure the sealing function for at least 30 days.

d. Other leak rate tests (1) The leakage rate for containment isola-tion valves 10-A0V-68A, B (penetration I-13A, B) for Low Pressure Coolant Injection system and 14-A0V-13A, B (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 ambient temper-ature) or 10 gallons per minute per valve (hydrostatically) tested at 1000 psig with ambient temperature.

Amendment No. jd$

172

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

JAFNPP 4

4.7 (cont *d)

(5) Type C test.

Type C tests shall be performed during each reactor shutdown for refueling but in no case at intervals greater than two year.-

(6) Other leak rate tests specified in Section 4.7d shall be performed during each reactor shutdown for refueling but

! in no case at intervals greater than two years.

1 3 f. Containment modificatioa Any major modification, replacement of a-l component which is part of the primary reactor containment boundary, or resealing a seal-welded door, performed after the pre-operational leakage rate test shall be followed by either a Type A, Type B, or Type C test, as applicable, for the area affected s by the modification. The measured leakage from this test shall be included in the test report. The acceptance criteria as oppro-priate, shall be met. Minor modifications, replacements, or resealing of seral-welded l doors, performed directly prior to the conduct of a scheduled Type A test do not require a separate test.

Amendment No. ,$ d ',)MI 174

)

JAFNPP-

.i 3.7 (cont'd) 4.7 (cont'd)

When the primary containment is inerted, it shall' be. continuously monitored for gross leakage by review of the inerting system makeup requiressents.:

The monitoring system may be taken out of service for maintenance, but shall be returned to service.

as soon as possible.

4. Pressure Suppression Chamber Reactor Building 4. Pressure Suppression Chamber - Reactor Building Vacuum Breakers Vacuum Breakers t

a. Except as specified in 3.7.A.4.b below, two a. The pressure suppression chammher reactor Pressure Suppression Chanaber Reactor Building building vacuume breakers and associated Vacuum Breakers shall be operable at all instrumentations including setpoint shall be times when the primary containment integrity checked for proper operation every three - ,

is required. The setpoint of the differen- months.

tial pressure instrumentation which actuates the pressure suppression chamber reactor building vacuum breakers shall be less than or equal to 0.5 psid.

a"

b. From and after the date that one of the pressure suppression chamber reactor building vacuum breakers is made or found to 'j be inoperable for any reason, reactor operation is permissible only during the succeeding 7 days, unless such vacuum l

l l

Amendment No.

177

JAFNPP 3.7 (cont'd) 4.7 (cont'd) breaker is sooner made operable, provided that the repair proceddre does not violate primary containment integrity.

'5. Pressure Suppression Chamber - Drywell Vacuum 5. Pressure Suppression Chamber - Drywell Breakers Vacuum Breakers l a. Whan primary containment integrity is a. Each drywell suppression chamber vacuum required, all drywell suppression chamber breaker shall be exercised through an vacuum breakers shall be operable and opening - closing cycle monthly.

positioned in the fully closed position except during testing and aa specified in 3.7.A.S.b b. When it is determined that one vacuum below. breaker is inoperable.for fully closing when operability is required, the-

b. One drywell suppression chamber vacuum breaker operable breakers shall be exercised v.ay be non-fully closed so long as it is deter- immediately, and every 15 days thereafter mined to be not more than l' open as indicated until the inoperable valve has been by the position lights. returned to normal service.

c. One drywell suppression ciamber vacuum breaker c. Once each operating cycle, each vacuum may be determined to be inoperable for opening. breaker valve shall be visually inspected to insure proper maintenance and

d. If specifications 3.7.A.S.a, b, and c cannot operation.

be met, an orderly shutdown will be initiated, and the reactor shall be placed in a cold d. A leak test of the drywell to suppression condition. chamber structure shall be conducted once per operating cycle; the acceptable leak rate is 0.25 in, water / min, over a 10 min period, with the drywell at 1 psid.

Amendment No.

178

JAFNFP 3.7 (cont *d) 4.7 (cont'd)

e. Leakage between the drywe'.1 and suppression e. Not applicable chamber shall not exceed a rate of 71 scfm as monitored via the suppression chamber 10 min pressure transient of 0.25 in, water / min.

f. The T;_f actuated vacuum breakers shall open f. Nct applicable when subjected to a force equivalent to 0.5 psid acting on the valve disc.

g. From and after the date that one of the g. During each refueling oatage each vacuum pressure suppression chamber /drywell vacuum breaker shall be tested to determine that the l force required to open the vacuum breaker does breakers is made or found to be inoperable for any reason, the vacuum breaker shall be locked not exceed the force specified in Specifica-closed and reactor operation is permissible tion 3.7.A.S.f and each vacuum breaker shall only during the succeeding sever days unless be insnected and verified to meet design such vacuum breaker is sooner made operable, requirements.

l provided that the repsir procedure does not l

violate primary containment integrity.

1 l

l Amendment No. )[ 179

9

.e JAENPP 3.7 (cont'd) 4.7 (cont *d)

(2.) With the reactor at reduced power level, trip main _ steam isolation valves and verify closure time.

d. At least twice per week, the main steam line power-operated ~ isolation valves shall be exercised by partial closure and subsequent reopening.

2. In the event any isolation valve specified in 2. Whenever an isolation valve listed in Table 3.7-1 Table 3.7-1 becomes inoperable, reactor power is inoperable, the position of at least one other operation may continue, provided at least one valve in each line having an inoperable valve valve in each line having an inoperable valve is shall be recorded daily.

in the mode corresponding to the isolated condition.

3. If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

{

i Amendment No.

186 l

l

JAFNPP 4.7 BASES (cont'd) assumption of no holdup in the secondary contain- As most leakage and deterioration of integrity is ment, resulting in a direct release of fission expected to occur through penetrations, especially products from the primary containment through the those with resilient seals, a periodic leak rate filters and stack to the environs. Therefore, test program of such penetrations is conducted at the specified primary containmant leak rate and the peisk pressure of 45 psig to insure not only filter efficiency are conservative and provide that the leakage remains acceptably low but also additional margin between expected offsite doses that the sealing materials can withstand the and 10CFR100 guidelines. accident pressure. For airlock leak test, a seal test at the peak pressure could be substituted The maximum allowable test leak rate at the peak for the complete airlock test, if no maintenance pressure of 45 psig (Pa) is 0.5 weight percent work is done which could affect the sealing per day (Lam). The maximum allowable test leak capability of the airlock.

rate at the reduced pressure of 23 psig (P t) will be verified to be conservative by actual The leak rate testing program was originally primary containment leak rate measurements at based on Commission guidelines for development of l both 45 psig and 23 psig upon completion of the leak rate testing and surveillance schedules for containment stru:ture. reactor containment vessels, (16) and discussed in Question 5.4 of the FSAR. With the exceptions To allow a margin for possible leakage deterior- listed in Table 4.7-2, the system conforms to the ation between intervals, the maximum allowable latest Commission guidelines (17). The exceptions l leak rate (Ltm), which will be met to remain on stated iu Table 4.7-2 are necessary since the normal test schedule, is 0.75 L t. In additional requirements were added after the addition, it is intended to operate the primary system was designed.

containment structure at a slight positive pressure to continuously monitor primary contain- B. Standbv Gas Treatmant System and ment leakage. C. Secondary Containment Initiating reactor building isolation and opera-tion of the Standby Gas Treatment System to maintain at least a 1/4 in of water va. "tm within the secondary containment provides an adequate test of the operation of the reactor Amendment No. ,97 194

JAFNPP Table 3.7-1 (Cont'd)

PROCESS PIPELINE PENETRATIr3 PRIMARY CONTAINMENT (Numbers in parentheses are keyed to numvers on following pages: signal codes are listed on following pages)

Power Loca tion Power Closing Drywell Valve Type to Open Ref. to to Close Isolation Time Normal h aarks and Line Isolated Penetration (6) (5) (6) Group Dryweli (5) (6) Signal (7) Status Exceptions Core Sper, Minimum Pump F1 e X-2iOA,B M0 Gate Ac B Outside Ac RM Not applicable Closed Core Spray to Reictor X-16 A,B M0 Gate Ac A Outside Ac RM Not applicable Open Note t10)

Core Spray to Reactor X-16 A,B M0 Gate AC A Outside Ac RM Not applicable Closed Note (10)

Core Spray to Testable Check Reactor X-16A,B A0 Check (3) A Inside Note (3) Rev. flow Not applicable Closed Yalve Note (3,16)

Core Spray Test to Suppression Pool X-210A,B M0 Globe Ac B Outside Ac G RM 45 Sec Closed Core Spray Pump Suction X-227A,B M0 Gate Ac B Outside Ac RM Not Applicable Open Dryw11 Equipment Drain Samp Discharge X-19 M0 Plug Ac B Inside Ac A,F,RM 30 Sec Open Drywell Equipment Drain Step Discharge X-19 A0 Plug Air /Ac B Outside Spring A,F,RM Not Applicable Closed (17)

Drywell Floor Drain Sump Discharge X-18 H0 Plug Ac B Inside Ac A,F,RM 30 Sec Open Drywell Floor Drain Sump Discharge X-18 A0 Plug Afr/Ac B Outside Spring A.F,RM Not Applicable Open Traveling Incore Explosive One valve Probe X-35A,B.C D Shear Dc A Outside Dc RM Not Applicable Open on each line One valve on Traveling Incore each ifne Probe X-35A,0,C.D SO Ball Ac A Outside Ac A.F,RM Not Applicable Open Note (14)

Traveling Incore Probe Purge X-35E SO Valve Ac A Outside Spring A,F,RM Not Applicable Closed Traveling Incore Probe Purge X-35E Check Fwd. Flow A Inside Process Rev. Flow Not Applicable Closed HPCI - Turbine Steam Supply X-11 M0 Gate Ac A Inside Ac L,RM 20 Sec Open ) Signal "G" opens HPCI - Turbire ) valve.

1 Signal "L" Steam Supply X-11 M0 Gate Dc A Outside Dc L,kM 20 Sec Closed ) overrides and Amendment No. FI, f8, Jd8

.i

~

JAFNPP [

TABLE 4.7-2 EXCEPTION TO TYPE C TESTS Certain Type C test will be performed or omitted.as follows:

Penetration' System Valve Local Leak Rate Test Performed X-7A, B, C, Main Steam 29-AOV-80A, B. These valves are air-operated _giobe valves --

and D C, and D pressurized in reverse direction and measurement c?

29-AOV-86A, B leakage will be equivalent to results from^ pressure C, and D applied in the same direction as when the valves would be required to' perform its satety function.

Therefore, pressure will be. applied between the isolation valves and leakage measured.' A water seal of 25 psig will be used on Che inboard valve to determine the outboard valve's leak rate.- (limit-11.5 scfh at 25 psig (a)

X-10 RCIC 13-MOV-15 See X-25-(27-AOV-131A,'B)

X-11 HPCI 23-MOV-15 See X-25 (27-AOV-131A, B)

X-25 Dry Well Inerting 27-AOV-112 This valve is a butterfly valve - pressurization-CAD and Purge in reverse direction and measurement of leakage will be equivalent to results from pressure applied in the same direction as that when the valve would be required to perform its safety function.

X-25 Dry Well Inerting 27-AOV-1311. These valves will be tested in the reverse CAD and Purge 27-AOV-131B direction, since the system was not designed for pressure'to be applied in the same direction as that when the valve would be required to performs its safety function.

Basis - The pressurizaticr tirection was not a requirement at the time of plant designs; to redesign the system to permit this is not-feasible as it would delay plant operation.

X-26 A/B Dry Well Inerting 27-AOV-113 See X-25 (27-AOV-112)

CAD and Purge 27-MOV-113 This globe valve will be tested in the reverse direction. See X-25 (27-AOV-131A, B)

(a) During cycle 3 the plant may operate with valve 29-AOV-86A type C leakage not to exceed'300 SCFD and valves 29-AOV-86A, B, C and D total leakage not to exceed 1104 SCFD.

AmendmentNo.pd,p 211

_I'\ f JAFNPP TAa1.z 4 .7-2 ccowr*m l

$yptge valve local leak pate Test Pwrforme1' Penetrat19n 27-SOV-120s See X-2% (27-A0V-131A, 3) 27-Sov-1218 27-SOV-1228 X-31 Bd Dry well Inerting 27-SOV-125s See X-25 (27-A0V-131A3 CAD and Purge 10-t'tN-31A his valve will trt pressurized in the X-39A Cont. Spray reverse direction and leakage seamared.

See X-25 (27-50V-131 A, B)

X-39a O)nt. Spray 10-f10V-31A See,X-39A ,

21JtT VSM-1007 See X-25 (27-A0V-131A, 8)

X-45 See X-25 (27-AOV-131A, 9)

X-59 Dry well Inerting 27-50V-12 3 A Cad and Purge X-202 Torus Vacuum AOV-101A/B See X-25 (2 7-A0V-112)

Breakers See X-25 (27-AOV-131A, 3)

X-203A Dry well Inerting 27-SOV-119s O D a M Purge ,

X-2033 Dry W:1'. Inerting 27-SOV-124A See X-25 (27-AOV- 131A1 CAD an1 Purge X-205 Dry well Inerting 27-AOV-117 See X-25 (27-Aov- 1123 CKJ and Purge 27-HOV-117 See X-25 (27-41ov-113) l

( RCIC, RHH E!91 not be tested as 11nsa are water

[

X-210 A/B ed by suppression chastber water l '- 25 (27-A0V-131A, b) l RHR 10-POV-29A . h. 31ve will be tested in the reverse l

X-211A dits tion. See X-25 (27-A0V-131 A, 21 X-2113 RHR 10-TOV-30s This valve will be tested in the reverse direction.

X-212 RCIC 13 W -134 See X-25 (27-Auv-131 A/B)

X-218 JLRT VSM-100T See I-25 (27-Auv-131A/u)

X-220 Dry well Inerting 27-A0V-116 See X-25 (27-AOV-112)

CAD and Purge 27-SOV-132A See X-25 (27-A0v-131A/s) 27-SOV-132s see X-210 A/B X-222 MPC1 RHR See X-210 A/D X-224 X-225 RHR See X-210 A/n 212 Amendment fd

.a s

JAFNPP . . . .

TABLE 4.7-2 (CONT'b)

I-Penetration System yalve Local Leak Rate Test Performed ~

s X-226 i HPCI See'X-210 A/B ~

X-227 Core Spray. .See X-210 A/B-

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t j X-228 Condensate See X-210 A/B 1

4 1

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4 4

1 3

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Amendment No.

213

JAFNPP 3.8 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES 4.8 SURVEILLANCE REQUIREMENTS-Source Leakace Test Tests for leakage and/or contamination shall be performed by~ the licensee or by other persons Specification:- specifically authorized by the Commission or an agreement State, as follows:

Radioactive sources shall be leak tested for con-tamination. The leakage test shall be capable of 1. Each sealed source, except startup sources subject detecting the presence of 0.005 microcurie of radio- to core flux, containing radioactive material, active material on the test sample. If the test other than Hydrogen 3, with a half-life greater reveals the presence of 0.005 microcurie or more of than thirty days and in any form other than gas removable contamination, the source shall be decon- shall be tested for leakage and/or contamination taminated, and repaired, or be disposed of in at intervals not to exceed six months.

accordance with Commission regulations.

2. The periodic leak test required does not apply to Those quantities of by-product material that exceed sealed sources that are stored and not being the quantities listed in 10 CFR 30.71 Schedule B are used. The sources excepted from this test shall to be leak tested in accordance with the schedule be tested for leakage prior to any use or transfer shown in Surveillanca Requirements. All other sources to another user unless they have been leak tested (including alpha emitters) containing greater than within six months prior to the date or use or 0.1 microcuries are also to be leak tested in accor- transfer. In the absence of a certificate from a l dance with the Surveillance Requirements. transferor indicating that a test has been within six months prior to the transfer, sealed source BSSeS shall not be put into use until tested.

Ingestion or inhalation of source material may give 3. Startup sources shall be leak tested prior to and rise to total body or organ irradiation. This following any repair or. maintenance and before specification assures that leakage from radioactive beir:g subjected to core flux.

material sources does not exceed allowable limits.

In the unlikely event that those quantitles of radio-active by-product materials of interest to this specificrtion which are exempt from leakage testr J are ingested or inhaled, they represent less than one maximum permissible body burden for total body irradiation. The limits for all other sources i (including alpha emitters) are based upon 10 CFR 70.39(c) limits for plutonium.

, Amendment No.

214 1

JAFFPP 3.9 Continued 4.9 Continued-s

2. The Diesel Fuel Oil Transfer System shall be 2. During the monthly diesel generator' testing, operable whenever the diesel generator it' the diesel fuel oil transfer systems shall be supplies is required to be opdrable, except as checked for proper operation.

specified below:

a. From and after the time that one fuel oil l transfer pump per Diesel Generator System is made or found to be inoperable for any reason, continued reactor operation is permissible for a period not to exceed 60 days; provided that the remaining fuel oil transfer pumps be demonstrated to be opereble immediately and weekly thereafter.

b. From and after the time that only two fuel-oil transfer pumps per Diesel Generator System are operable, continued reactor opertion is permissible for a period not to exceed 30 days total per pair of diesels, provided that the remaining fuel oil transfer pumps are demonstrated to be operable and daily thereafter.

Amendment No. JKI 219

JAFNPP 4.9 BASES (cont'd)

D. Battery System Measurements and electrical tests are conducted at specified intervals to provide indication of cell condition and to determine the discharge capability of the batteries.

E. LPCI MOV Indeoendent Power Supolv Measurement and electrical tests are conducted at specified intervals to provide indication of cell condition, to determine the discharge capability of the battery.

F. Reactor Protection Power Supplies l Functional tests of the electrical protection assemblies are conducted once each six (6) months utilizing a built-in test device and once per operating cycle by performing an instrument calibration which verifies operation within the limits of Section 4.9.G.

Amendment No. ,[6 226

JAFNPP 3.10 (cont'd) control rod after the fuel assemblies in the cell containing.(controlled by) that control rod have been removed from the reactor core.

All other refueling interlocks shall be operable.

7. In the "refuel" mode, there are interlocks which -

prevent the refueling bridge (if loaded) from moving toward the core unless all control rods are fully inserted. Those interlocks may be bypassed during spiral loading except for those control cells which contain fuel or-that control cell which is being loaded. Interlocks for all cells containing fuel, or for any cell about to be loaded, shall be operable.

Amendment No. ,f 230

J.

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A L

B Y

L L

A N

P O I

P a N T 0 F N 3 A E 2 J T N

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G A

P S

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i JAFNPP 3.10 BASES (cont'd)

Switch is in the Refuel position only one control rod can be withdrawn except as noted in Specifica-tions 3.10.A3 D and E. The refueling interlocks, in combination with core nuclear design and refueling procedures limit the probability of an inadvertent criticality. The nuclear character-istics of the core assure that the reactor is subcritical even when the highest worth control rod is fully withdrawn. The combination of re-fueling interlocks for control rods and the refueling platform provide redundant methods of preventing inadvertent critically even after procedural violations. The interlocks on hoists provide yet another method of avoiding inadvertent criticality.

For a new core the dropping of a fuel assembly into the vacant fuel location adjacent to.a with-l drawn control rod does not result in an excursion or a critical configuration, than adequate margin is provided.

A spiral unloading pattern in one in which the fuel in the outer-most cells (four fuel bundles surrounding a control blade) is removed first.

Unloading continues by removing the remaining outermost fuel by cell no that the center cell with be removed last. Spiral loading is the reverse of unloading. Spiral loading and unloading preclude the formation of flux traps (moderator. filled cavities surrounded on all sides by fuel). It is not necessary to accomplish a full core offload or-onload in order to utilize the spiral movement procedure as long as the partial unloading /re-loading.

Amendment No. pdI 235

_m_____

JAFNPP l 3.10 BASES (cont'd)

The maintenance is performed with the Mode Switch This Specification provides assurance that inad-in the Refuel position to provide the refueling vertent criticality does not occur during such interlocks normally available during Part A of operation.

these Bases. In order to withdraw a second con-trol rod after withdrawal of the first rod, it is This operation is performed with the Mode Switch necessary to bypass the refueling interlock on in the Refuel position to provide the refueling the first control rod, which prevents more than interlocks normally available during refueling as one control rod from being withdrawn at the same explained in Part A above. In order to withdraw time. The requirement that an adequate shutdown more than one control rod, it is necessary to margin be demonstrated with the control rods bypass the refueling interlock on each withdrawn remaining in-service insures that inadvertent control rod which prevents more than one control criticality cannot occur during this rod from being withdrawn at a time. The require-maintenance. The shutdown margin is verified by ment that the fuel assemblies in the cell con-demonstrating that the core is shutdown even if trolled by the control rod be removed from the the strongest control rod remaining in-service is reactor core before the interlock can be bypassed fully withdrawn. Disarming the directional ensures that withdrawal of another control rod control valves does not inhibit control rod scram does not' result in inadvertent criticality. Each capability. control rod essentially provides reactivity con-trol for the fuel assemblies in the cell associated The requirement for SRM operability during the with that control rod. Thus, removal of an entire maintenance is covered in Part B above. cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.

The intent of this Specification is to permit the unloading of a significant portion of the reactor The requirement for SRM operability during these-core for such purposes as in-service inspection operations is covered in Part B above.

requirements, examination of the core support plate, etc.

Amendment No.

236

JAFNPP 3.11 (cont'd) 4.11 (cont'd)

B. Crescent Area Ventilation B. Crescent Area Ventilation Crescent area ventilat'on and cooling equipment Unit coolers serving ECCS components shall be shall be operable on a cor?.nuous basis whenever checked for operability once/3 months.

specification 3.5.A, 3.5.B, and 3.5.C are required to be satisfied. 1. When it is determined that two unit coolers serving ECCS components in the same compart-

1. From and after the date that more than one ment are made or found inoperable, reactor unit cooler serving ECCS compartment are operation may continue for 7 days unless made or found to be inoperable, all ECCS one is made operable earlier.

components in that compartment shall be considered to be inoperable for purposes of 2. Temperature indicator controllers shall be specification 3.5.A, 3.5.B, and 3.5.C. calibrated once/ operating cycle.

l

3. If 3.11.B.1 cannot be met, the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. Battery Room Ventilation C. _Batterv Room Ventilation Battery room ventilation shall be operable on a Battery room ventilation equipment shall be continuous basis whenever specification 3.9.E checked for operability once/ week.

is required to be satisfied.

1. When it is determined that one battery room

1. From and after the date that one of the ventilation system is inoperable, the re-battery room ventilation systems is made or maining ventilation system shall be checked l found to be inoperable, its associated for operability and daily thereafter.

battery shall be considered to be inoper-able for purposes of specification 3.9.E. 2. Temperature transmitters and differential pressure switches shall be calibrated once/

operating cycle.

Amendment No.df , p 239

JAFNPP

.c 3.11 (cont'd) 4.11 (Cont'd)

D. Emeroency Service Water System D. Emeroency Service Water System 1.- To ensure adequate equipment and area 1.- Surveillance of the ESN system shall be cooling, both ESN: systems shall be. operable performed as follows:-

when the requirements of specification 3.5.A and 3.5.B must be satis- Item Frecuency fled, except as specified below in specifi-cation 3.11.D.2. a. Simulated Each Automatic operating Actuation cycle Test

b. Flow Rate Once/

Test - ESW pumps 3 months shall deliver at least 3,250 gpm against a system head corresponding to a total pump head of SE80 psi, as determined from the pump certifi-cation curve by measuring the pump shutoff head which shall bedb117 psi.

~

c. Pump Operability Once/ month

d. Motor Operated Once/s_ nth Valses AmendmentNo.JV[

240

JAFNPP 3.11 (cont'd) 4.11 (cont'd)-

E. Intake Deicing Heaters E. Intake Deicina Heaters-Intake heaters are required to be operable when 1. The six heater feeder ammeters shall.be-intake. water temperature is 4637'F. A minimum ' checked weekly whenever the intake water of 18 out of 88 heaters are required to be temperature is dh37'F.

4 operable to maintain the required flow for the

. ESW and RHRSW System. 2. The individual heaters shall be monitored once/6 months for rated heater current or

1. -It specification 3.ll.E.1 cannot be met'the as required by large deviations in the

, reactor shall be placed in a cold condition feeder checks-in 4.ll.E.1 above. l within 24 aours.

3. Resistance to ground shall be checked once/

operating cycle.

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4 4

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i Amendment No.

242

JAFNPP LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS.

3.12 FIRE PROTECTION SYSTEMS 4.12 EJR6 PROTECTION SYSTEMS '

Applicability: Apolicability:

Applies to the Operational Status of the Fire Applies to the Surveillance of the Fire Protection Systems. < Protection System.

Obiective: Objective:

To assure operability of the Fire Protection To verify the operability of the Fire Systems. Protection Systems.

Specification: Specification:

A. Hioh Pressure Water Fire Protection System A. High Pressure Water Fire Protection System 1.a. Both high pressure water fire protection- 1. High pressure water fire protection system pumps and associated automatic and manual testing:

initiation logic shall be operable and aligned to the high pressure water fire Item Frequency header.

a. High pressure water Once/ week

b. The high pressure water fire protection firs protection sys-system shall be operable with an operable tem pressure check.

flow path capable of taking suction from the lake and transferring the water through b. Each pump, on a STAG- Once/ month l~

distribution piping with operable section- GERED TEST BASIS, by alizing control or isolation valves to the starting and operating yard hydrant curb valves and the first it for at least 20 valve ahead of the water flow alarm device minutes on recirculat-on each sprinkler, hose standpipe or spray ing flow system riser required to be operable per specifications 3.12.B and 3.12.D. c. Valve operational test Once/12 months

d. System flush Once/6 months

e. Functional test including: Once/18 Amendment No. )d, pd months 244a

';d ej i :i JAFNPP C. Carbon Dioxide Systems (Cont'd)

2. If the CO2 Protection for the areas listed in Table 3.12.2 cannot be restored to an operable status within 14 days a written report.to the Commission outlining the. action taden, the cause of inoperability, and plans aad schedule to restore the system to an operable status shall be prepared and submitted within 30 days.

D. Manual Fire Hose Stations D. Manual Fire Hose Stations

1. a. The manual fire hc;e stations listed in 1. The manual fire hose stations are inspected as l Table 3.12.3 shall be operable except as listed in Table 4.12.3.

specified below:

b. From and after the date that any of the manual fire hose stations listed in Table 3.12.3 is made or found to be inoperable, l

additional hose lengths shall be added to adjacent operable manual hose stations such that the entire area of protection is maintained within one hour.

Amendment No. )H[

244f

JAFNPP 3.12 and 4.12 BASES (continued)

C. The carbon dioxide systems provide total flood D. Manual hose stations provide backup fire pro-protection for eight different safety related tection throughout the Plant. Those hose areas of the plant from either a 3 ton or 10 stations that are in or near areas with safety ton storage unit as indicated in Table 3.12.2. related equipment are listed in Table 3.12.3. l Both CO2 storage units are equipped with Hose station location and hose length selection mechanical refrigeration units to maintain the provides the capability of reaching any fire in storage tank content at O'F with a resultant a safety related area with the hose stream.

pressure of 300 psig. Automatic smoke and heat When any of the hose stations listed in Table detectors are provided in the CO2 Protected 3.12.3 is inoperable, providing additional hose areas and initiation is automatic and/or manual lengths from other operable hose stations as indicated in Table 3.12.2. For any area in assures maintenance of this capability.

which the CO2 Protection is made or found to Periodic inspection and tests are in accordance be inoperable, continuous fire detection is with NFPA Code guidelines and assures preven-f available and one or more large wheeled CO2 tion, detect! .1, and correction of hose, nozzle, fire extinguisher is also available for each valve and/or gasket damage or deterioration to area in which protection was lost. maintain high levels of operability.

weuxly checks of storage tank pressure and E. Early fire detection and fire fighting activity level verify proper operation of the tank is essential to ensuring that any fire will refrigeration units and availability of result in minimum damage to safety related sufficient volume of CO2 to extinguish a equipment. Since each area monitored utilizes fire in any of the protected areas. a number of smoke and/or heat detectors when more than one detector is inoperable, early Performance of the periodic tests and inspec- fire detection is assured by establishing a tions listed in Table 4.12.2 are in accordance patrolling fire watch which check the area with NFPA-12, 1973, will verify the integrity where the detectors are inoperable at least of system nozzles and distribution headers as hourly.

well as detect and remove any accumulation of rust or scale. The use of "puff test" rather Testing of smoke and heat detectors and associ-than full flow tests will demonstrate proper ated circuitry every 6 months, in accordance valve operation without the attendant potential with manufacturers and NFPA 72E-1974 recommen-equipment and personnel hazards associated with dations ensures a high level of operability.

full flow tests.

F. Fire barrier penetration seals are designed to give 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or more protection and to meet the requirements of IEEE - 383, "Fire Test of Building Construction and Materials". Visual inspection and leak testing ensure that seals are intact. Leak testing with open flame or combustion generated smoke is prohibited.

244i

6.3 PLANT STAFF QUALIFICATIONS 6.3.1 The minimum qualifications with regard to educational background and experience for plant staff positions shown in Fig. 6.2-1 shall meet or exceed the minimum qualifications of ANSI NI8.1-1971 for comparable positions; except for the Radiation and Environmental Services Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 6.3.2 The Shift Technical Advisor (STA) shall meet or 6xceed the minimum requirements of either Option 1 (Combined SRO/STA Position) or Option 2 (Continued use of STA Position), as defined in the Commission Policy Statement on Engineering Expertise on Shift, published in the October 28, 1985 Federal Register ( 50 FR 43621) . When invoking Option 1, the STA role may be filled by the Shift Supervisor or Assistant Shift Supervisor (1).

6.3.3 Any deviations will be justified to the NRC prior to an individual's filling of one of these positions.

NOTE:

(1) The 13 individuals who hold SRO licenses, and have completed the Fit : Patrick Advanced Technical Training Program prior to the issuance of Lhis license amendment, shall be considered qualified as dual-role SRO STAS.

6.4 RETRAINING AND REPLACEMENT TRAINING A training program shall be maintained under the direction of the Training Superintendent to assure overall proficiency of the plant staff l organization. It shall consist of both retraining and replacement training and shall meet or exceed the minimum requirements of Section 5.5 of ANSI NI8.1-1971.

The retraining program shall not exceed periods two years in length with a curriculum designed to meet or exceed the requalification requirements of 10 CPR 55, Appendix A. In addition fire brigade training shall meet or exceed the requirements of NFPA 27-1975, except for Fire Brigade training sessions which shall be held at least quarterly. The effective date for implementation of fire brigade training is March 17, 1978.

6.5 REVIEW AND AUDIT Two separate groups for plant operations have been constituted. One of these, the Plant Operating Review Committee (PORC), is an onsite review rroup. The other is an independent review and audit group, the offsite Safety Review Committee (SRC).

Amendment No pI, )[, , 9I, 1 1 248

m , - .A_

ATTACHMENT II TO JPN-88-023 PROPOSED ADMINISTRATIVE CHANGES TO THE TECHNICAL SPECIFICATIONS (JPTS-86-004)

NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

Attachment II to JPN-88-023 SAFETY EVALUATION Page 1 of 16 Section I DESCRIPTION OF THE PROPOSED CHANGE The proposed changes to the James A. FitzPatrick Technical Specifications affect several'pages and specifications.

Specifically, the changes are:

Section Pace Description (a) Table of Contents 1 Add section 3.0 titled ' General'.

Change page number for Specifica-tion 3.1 from page '30' to page

$30f'.

[b] List of Tables vi Insert title for Table 4.6-2.

Insert title for Tables 3.12-1, 3.12-2, 3.12-3, 4.12-1, 4.12-2, and 4.12-3.

[c] List of Figures vii Insert 'f' for Figure 3.1-2 title.

Add 'and 3.5.J.3' to title of Figure 3.5-1.

Change Figure 3.5-9 title to

' Deleted'.

l Insert title for Figure 3.5-12.

(d) Spec. 1.0 1 Correct the spelling of

' explicitly'.

l l [e] Spec. 1.0.F.4 2 Insert 'a' in first sentence l defining instrument check.

(f) Spec. 1.0.I.4 4 Add ' Amendment No. 83' to bottom of page.

Replace ' trips' with ' trip' in first sentence of spec. 1.0.4.

(g) Spec. 1.0.Q 5 Correct the abbreviation of

' cont'd'.

Delete 'a' from definition of refueling outage.

Attachment II to JPN-88-023 SAFETY EVALUATION Page 2 of 16 Section Pace Description

[h] Spec. 1.0.X.a 6 Replace 'a' from definition of staggered test basis with '"n"'.

Place n inside quotes (last line),

[i] Spec. 1.1.B 8 Replace 'less than' with 'less than or equal to' prior to 10%.

[j] Spec. 1.2 & 2.2 BASES 29 Correct spelling of 'resulting'.

[k] Spec. 3.0.E BASES 30b Correct spelling of ' inoperable'.

[1] Spec. 3.1 BASES 34 Rewrite the top paragraph, right column, to clarify nuclear instrumentation coverage for the reactor modes of operation.

[m] Spec. 4.1 BASES 38 Insert '44' into Amendment No.

[n] Table 3.1-1 41 Insert 'X' to require Manual Scram Operability in Run Mode.

41b Insert two '<' symbols in Trip Level Setting for Turbine Control Valve Fast Closure.

42 Insert a 'h' symbol before 10%

valve closure.

Insert 'less than' prior to 1005 psig in Note 3 to correct the condition for which the MSIV closure scram is bypassed.

[o] Spec. 3.2 BASES 57 Correct five grammatical errors:

1) replace ' drop' with ' drops'

2) remove 'this'

3) replace 'of' with 'or'

4) replace 'and' with 'or'

5) replace ' setting' with

' settings' Delete ' flow' from first sentence of the third paragraph in the right-hand column.

58 Correct spelling of ' channel'.

[p] Spec. 4.2 BASES 61 Insert 't' into formula for optimum interval between tests.

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l Attachment II to JPN-88-023 l SAFETY EVALUATION I Page 3 of 16  !

I Section Pace Description (q) Table 3.2-2 68 Insert 'THIS ITEM INTENTIONALLY BLANK' for item no. 10.

70a Replace 'psid' with 'dp' in Item No. 28.

Insert 'THIS ITEM INTENTIONALLY BLANK' for item nos. 22, 23 & 24.

70b Insert 'dp' in Item No. 31.

(r) Notes for Table 76c Correct spelling of ' permissible'.

3.2-6 (s] Table 3.2-7 77 Replece '>' with 'd' under the trip level _ setting column for the reactor low-low water level instrument prior to -38.

(t) Specs. 3.3.A.2 & 89a Retype such that paragraph 4.3.A.2 90 4.3.A.2.e resides in the right side column and paragraph 3.3.A.2.d continues with text currently on page 90.

(u) Spec. 3.3.B.1 91 Replace ' rods' with ' rod' in first sentence.

(v) Spec. 4.3.B.3 101 Delete 'feedwater and' from BASES the fourth complete sentence in the first paragraph in the right-hand column.

(w) Spec. 3.4.B 106 Insert 'B' for paragraph number.

(x) Spec. 3.5.A.2 114 Insert change bars for Amendment 95.

(y) Spec. 4.5.B.1 115a Delete 'of 3,700 gpm' from third sentence.

(z) Spec. 3.5.C.b 118 Change referenced paragraph from

'3.5.C' to '3.5.C.1'.

(aa) Specs. 3.5.D & 119 Replace ' relief / safety' with 4.5.D ' safety / relief'.

(bb) Spec. 3.5.F 122 Insert 'F' for paragraph identification.

Attachment II to JPN-88-023 SAFETY EVALUATION Page 4 of 16 Section Egga Description

[cc] Spec. 3.5.H 123 Insert 'at' into fourth sentence describing APLHGR.

[dd] Spec. 3.6.E & 4.6.E 142a Delete note for effective date.

142b Delete entire expired page.

143 Delete note for effective date.

Change referenced specification from '3.6.B.1' to $3.6.E.1' and from '3.6.B.2' to $3.6.E.2'.

143a,b Delete both expired pages.

[ee] Spec 3.6.G 144 Replace ' operable' with

' inoperable' in second sentence of paragraph G (Jet Pumps).

[ff] Specs. 3.6.E & 152 Delete $ coincident high drywell 4.6.E BASES pressure and' (second sentence, second paragraph) and replace

' low-low' with ' low-low-low' from the same sentence and paragraph in the left hand column.

Replace ' relief / safety' with

'cafety/ relief'.

153 Insert 'less than or equal to' into the second lineg(left-hand column) prior to 212 F.

[gg] Spec. 3.6.F & 153 Correct spelling of 'will'.

4.6.F BASES

! [hh] Spec. 4.7.A.2.c.1 171 Change referenced Table '3.7-l' I to '4.7-2'.

( [11] Spec. 4.7.A.2.c.4 172 Change referenced Table '3.7-2' l

to '4.7-2'.

[jj] Spec. 4.7.A.2.e,f 174 Insert '40' in Amendment No.

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f

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l Attachment II to JPN-88-023 SAFETY EVALUATION Page 5 of 16 Section Phge Dascription Replace 'on' with Sof' in fourth sentence of spec. 4.7.A.2.f.

[kk) Spec. 3.7.4.a 177 Insert 'less than or equal to '

into the last line prior to 0.5.

(ll) Spec. 3.7.A.S.a 178 Replace 'The' with 'When' in first line.

(mm) Spec. 3.7.A.S.g 179 Replace 'chamberreactor building' with ' chamber /drywell' in first sentence.

(nn) Spec. 4.7.D.1.c.2 186 Insert ' ' after ' level' and delete ',' after ' trip'.

(oo) Spec. 4.7.A BASES 194 Replace 'AEC' with ' Commission' in first and second sentences.

Reposition of a comma in first sentence.

[pp) Table 3.7-1 201 Replace Drywell Penetration for the TIP Purge from 'X-35B' to

'X-35E'.

(qq) Table 3.7-2 211 - Replace '3.7-2' to '4.7-2'.

213

[rr) Spec. 3.8 & 214 Insert ' contamination, the source 4.8.2 shall be decontaminated,' into the third sentence of the first paragraph.

Insert an 'a' into Specification 4.8.2 (third sentence).

[ss) Spec. 3.9.C.2.a 219 Insert 'one' into first sentence of spec. 3.9.C.2.a

[tt] Spec. 4.9.F BASES 226 Correct spelling of ' tests'.

[uu) Specification 3.10.8 230 Insert '59' into Amendment No.

Delete expired spec. 3.10.8.

230a Delete entire expired page.

1 Attachment II tx) JPN-88-023 l SAFETY EVALUATION Page 6 of 16 Rection P&qe Description (vv) Spec. 3.10 BASES 235 Correct spelling of 'not'.

(ww] Speq. 3.10 BASES 236 Correct spelling of ' BASES' at the top of the page.

[xx) Spec. 3.11.B.1 239 Replace '3.5.C' with '3.5.B' and

'3.5.D' with '3.5.C' in last two lines.

(yy] Spec. 4.11.C.1 239 Correct spelling of ' ventilation '.

(zzi Spe . 3.11.D.1 240 Insert 'both ESW systems shall be' into f!'st secconce.

Delete 'the' from first line.

(aA] Spec. 4.11.E.2 242 Replace '3.11.E.1' with

'4.11.E.1'.

(bB) Spec. 4.12.A.1.b 244a Replace 'once/ week' with

'once/ month'.

(cC) Sp.c. 3.12.D.1.a & b 244f Replace '3.]2.2' with '3.12.3'.

(db] Spec. 3.12.D BASES 2441 Replace '3.12.2' with '3.12.3' in second sentence.

(eE) Spec. 6.4 248 Replace 'Cnordinator' with

' Superintendent'.

Section II PURPOSE OF THE PROPOEED CHANGE The purpose of the proposed changes are to correct typographical and other errors in the Technical Specifications. These errors were discovered during the process of obtaining previous license amendments. The tyoes of errors within the Technical Specifications include type ph'- - errors (e.g. , misspelled words) , grammatical errors (e.g nt-v asary e r ds) , and expired pages. These changes are adminisura" ..nd wi .frove the clarity of the Technical Specificatione The prnp A e (a)) to pagu i adds the title for Specificeti 6 r '. ) and changes the page number for Specification ' . 't .oth of these changes are due to the v

_. w m _ _

Attachment II to JPN-88-023 SAFETY EVALUATION page 7 of 16

addition of Specification 3.0 which was approved and issued as Amendment 83.

The proposed changes (item (b)) to page vi adds the title for Tables 4.6-2, 3.12-1, 3.12-2, 3.12-3, 4.12-1, 4.12-2, and 4.12-3.

Table 4.6-2 was added to the Specifications by Amendment 28 but never entered into the List of Tables. Tables 3.'.2 3.12 ~ and h 4.12 4.12-3 were added to the Specifications by Amendment 34 but never entered into the List of Tables.

s The proposed changes (item (c]) to page vii updates the List of Figures. The titles for Figures 3.1-2 and 3.5-1 are incomplete and need to be completed. Figure 3.5-9 was deleted by Amendment 109 but its entry in the List of Figures was never removed. Figure 3.5-12

[ was added to the Specifications by Amendment 109 but never entered into the List of Figures.

The proposed change (item (d)) to page 1 corrects the spelling of ' explicitly'.

The proposed change (item (e]) to page 2 inserts an 'a' into the first sentence defining instrument check. This change will make the sentence grammatically correct.

The proposed change (item (fj) to page 4 adds ' Amendment No. 83' to the bottom of the page. This page was updated by Amendment 83 but never noted on the page. Also, a grammatical change is made to the first sentence of the startup/ hot standby mode of operation description.

- The proposed change (item [g]) to page 5 corrects the spelling cf the abbreviation of ' cont'd' and deletes 'a' from the definition of refueling outage. This correction makes the definition grammatically correct.

The proposed change (item [h]) to page 6 replaces 'a' with m '"n"' in Specification 1.0.X.a. In this case, '"n"' means number of systems. Alco, the n that exists in the last line of this Specificat'nn is placed inside quotes. ,

The proposed change (item (i]) to page 8 inserts an inequality expression to account for core flow less than or equal to 10% of rated.

The proposed change (item (j)) to page 29 corrects the spelling of 'resulting'.

The proposed change (item (k)) to pagc 30b corrects the spelling of ' operable'.

e _ __ _ - - - _ - - - - - - - - - - - - - - -_ - - - - - - - - - - - - - - - - - - - - _ - - - _ - -__

Attachment II to JPN-88-023 SAFETY EVALUATION Page 8 of 16 The proposed change (item (1]) to the top paragraph, right-hand column on page 34 is intended to clarify the existing paragraph for nuclear instrumentation coverage for the reactor modes of operation.

The rewritten paragraph does not change the meaning of the paragraph but makes it more understandable.

The proposed change (item (m]) to page 38 adds $44' into Amendment No. to signify this page was previously updated by this amendment.

The proposed change (item rn)) to Table 3.1-1 (page 41) adds a coquirement that a manual scram trip function be operable in the run mode. This requirement was inadvertently deleted when the table was updated for a submittal which was subsequently approved and issued as Amendment 98 to the FitzPatrick Technical Specifications.

The proposed change (item (n)) to Table 3.1-1 (page 41b) adds l two '<' symbols for the turbine control valve fast clocure trip level setting. The symbols were inadvertently omitted when the table was .

last updated.

The proposed changes (item (n)) to Table 3.1-1 (page 42) inserts

'S' prior to 10% valve closure and inserts 'less than' to Note 3.

Both omissions occurred while updating the page for a previous submittal.

The proposed changes (item (o)) to page 57 correct five grammatical errors. Also, ' flow' is deleted from the paragraph  !

describing reactor water cleanup instrumentation since this system is independent of flow.

The proposed change (item (o)) to page 58 corrects the spelling of ' channel'.

The proposed change (item [p]) to page 61 inserts a 't' into the mathematical equation for the optimum interval between tests.

The proposed change (item (q)) to Table 3.2-2 (page 68) inserts

'THIS ITEM INTENTIONALLY BLANK'. This line item description should have been included in the submittal for Amendment 84.

The proposed change (item (q)) to Table 3.2-2 (page 70a) replaces 'psid' with 'dp' in item no. 28. This is the appropriate expression for differential pressure in this instance. Also, insert a line item description that the instrumentation for item nes. 22

- 24 has been deleted. The instrumentation was removed by a plant modification which was approved by Amendment 48.

The proposed change (item (q)) to Table 3.2-2 (page 70b) inserts

'dp' in item no. 31 to complete the trip level setting for the HPCI turbine steam line high flow trip function.

I

Attachment II to JPN-88-023 SAFETY EVALUATION Page 9 of 16 The proposed change (item (r]) to page 76c corrects the spelling of ' permissible'.

The proposed change (item (s)) to page 77 corrects a wrong inequality symbol.

The proposed change (item (t]) to pages 89a and 90 rearrange the pages so that the appropriate paragraphs fall under the appropriate specifications (Specs. 3.3.A.2 and 4.3.A.2).

The proposed change (item (u]) to page 91 (first sentence of Spec. 3.3.B.1) corrects a typographical error.

The proposed change (item (v]) to page 101 eliminates the statement indicating that the power level for automatic cutout of the rod worth minimizer is sensed by feedwater flow. Automatic cutout of the RWM is sensed only by steam flow.

The proposed change (item (w]) to page 106 inserts 'B.' to identify Specification 3.4.B.

The proposed change (item [x]) to page 114 adds the change bars from Amendment 95. The change bars are necessary to indicate what text has been previously changed.

l The proposed change (item (y]) to page 115a deletes 'of 3700 l gpm' from Specification 4.5.B.1. This deletion should have been l included in the submittal for Amendment 71. Amendment 71 pertained l

to the emergency service water pump surveillance requiroment.

l The proposed change (item (z!) to page 118 change <> a reference l specification. The Specification currently reference 4 (3.5.C) does l not exist and is replaced with the correct Specification (3.5.C.1).

l l The proposed change (item (aa)) to page 119 replaces i $ relief / safety' with ' safety / relief'. This change is necessary for i the specifications to be consistent.  !

The proposed change (item (bb]) to page 122 inserts 'F' to l

identify the specification.

The. proposed change (item (cc)) to page 123 grammatically corrects the fourth sentence of Specification 3.5.H.

The proposed change (item (dd)) to page 342u deletes the note for the effective date.

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Attachment II to JPN-88-023 SAFETY EVALUATION Page 10 of 16 The proposed changes (item [dd)) to pages 142b, 143a, and 143b delete the pages in their entirety. These pages are no longer applicable.

The proposed change (item (dd)) to page 143 removes the note for effective date. This note is no longer necessary.

The proposed change (item (ee]) to page 144 changes operable to inoperable in the second sentence of Specification 3.6.G. The existing specification reads:

"Whenever the reactor is in the start-up/ hot standby or run modes, all jet pumps shall be operable. If it is determined that a jet pump is operable, the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."

The second sentence contradicts the first sentence but is in agreement and correct if operable is changed to inoperable. The specification with the proposed change reads:

> "Whenever the reactor is in the start-up/ hot standby or run modes, all jet pumps shall be operable. If it is determined that a jet pump is inoperable, the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."

The proposed change (item (ff]) to page 152 eliminates

... coincident high drywell pressure..." from the second paragraph of Specification 3.6.E BASES. This same change was made to pages 66, 68, and 69 by Amendment 84. This change is a result of a plant modification (F1-83-034) which removed the high drywell pressure permissive for ADS actuation from the ADS logic. The change to this page was inadvertently omitted from the submittal for Amendment 84.

Also, on this same page and paragraph, "low-low" is being replaced by "low-low-low". This error has existed since the initial issuance of the Technical Specifications. As evidenced in Table 3.2-2, ADS actuation results due to a trip caused by reactor low-low-low water level. This change will make the specification consistent with the correct condition listed in tha table.

Also on page 152, ' relief / safety' is replaced with

' safety / relief' to maintain tonsistency in the Technical Specifications.

The proposed change (item (ff]) to page 153 inserts an inequality expression regarding safety / relief valves. The expression is replacing its corresponding inequair y sign.

t

Attachment II to JPN-88-023 SAFETY EVALUATION Page 11 of 16 The proposed change (item (gg)) to page 153 corrects the spelling of 'will'.

The proposed change (item [hh]) to page 171 correctly identifies a referenced table.

The proposed change (item (ii)) to page 172 co7 reenly identifies a referenced table.

The proposed change (item (jj]) to page 174 corrects two errors. A missing amendment no, is added and a grammatical error is corrected.

The proposed change (item (kk]) to page 177 inserts a missing inequality expression.

The proposed change (item (11]) to page 178 corrects a grammatical error.

The proposed change (item (mm]) to page 1*/9 replaces

'chamberreactor' with ' chamber /drywell' . This is a necessary change in order to achieve consistency with Specification 3.7.A.4.

The proposed change (item (nn)) to page 186 repositions a punctuation mark to make the sentence grammatically correct.

The proposed change (item (oo]) to page 194 replaces 'ACO' with

' Commission'. Also, a punctuation mark is repositioned.

The proposed change (item [pp)) to page 201 replaces drywell penetration 'X-35B' with 'X-35E'. This change will correctly identify the drywell penetration for the TIP purge.

! The proposed change (item (qq)) to pages 211 - 213 changes Table l 3.7-2 to Table 4.7-2. This was the original number of the table l and was renumbered by Amendment 40. This revised table number is I lacorrect and is being changed back to the original number.

(

l The proposed change (item (rr)) to page 214 inserts text into the third sentence (first paragraph) of Specification 3.8. This text has been missing since the initial issuance of the Technical Specifications. Also, an 'a' is inserted into the third 'ntence of Specification 4.8.2 to make it grammatically correct.

The proposed change (item (ss]) to page 219 inserts 'one' into the first sentence of specification 3.9.C.2.a. This word was l inadvertently omitted when the page was retyped for the submittal of Amendment 83.

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Attachment II to JPN-88-023 l

SAFETY EVALUATION Page 12 of 16 The proposed change (item (tt)) to page 226 corrects the spelling of ' tests'.

The proposed change (item (uu]) to page 230 deletes the expired Specification 3.10.8. The entire page 230a is deleted which contains the remainder of this specification. Page 230a will be replaced with page stating that the page has been intentionally deleted.

The proposed change (item (vv)) to page 235 corrects the spelling of 'not'.

The proposed change (item (ww]) to page 236 corrects the spelling of ' BASES'.

The proposed change (item [xx)) to page 239 correctly identifies two referenced specifications.

The proposed change (item (yy)) to page 239 corrects the spelling of ' ventilation'.

The proposed change (item (zz]) to page 240 inserts text missing since the initial issuance of the Technical Specifications to the first sentence of Specification 3.11.D.1. Also, a grammatical change is made to this same specification.

The proposed change (item (aA)) to page 242 correctly identifies a referenced specification.

The proposed change (item (bB)) to page 244a replaces the "once/ week" surveillance frequency for the high pressure water fire protection system pumps to "once/ month". This error was introduced in the submittal approved and issued as Amendment 80. This change to the surveillance frequency was never analyzed by the Authority or NRC. As a result, this change restores the surveillance frequency to its pre-Amendment 80 value.

The proposed change (item (cC)) to page 244f corrects the referenced Tables in Specifications 3.12.D.1.a and b.

The proposed change (item (dD]) to page 2441 correctly identifies the referenced Table in Specification 3.12.D BASES.

The proposed change (item (eE]) to page 248, Specification 6.4, replaces ' Coordinator' to ' Superintendent' to be consistent with Figure 6.2-1 (organizational chart).

Section III IMPACT OF THE PROPOSED CHANGE The proposed changes to the FitzPatrick Technical Specifications will not impact plant safety. All of the changes are administrative

. or editorial in nature. The proposed changes involve no limiting

E Attachmet:t II to JPN-88-023 SAFETY EVALUATION Page 13 of 16 conditions for operations, surveillance requirement, setpoint or safety limit changes, nor do they affect the environmental monitoring

. program. These changes clarify and improve the quality of the Technical Specifications by correcting editorial errors.

The majority of the proposed changes are the correction of typographical errors. These errors consist primarily of misspelled words but also include inadvertent deletion or misplacement of text.

In the proccas of updating a page for a previous amendment, portions of text were inadvertently deleted. In the case of misplaced text, which occurred once, a paragraph of one specification was moved into the adjacent specification. Resolution of these types of errors restore the Technical Specifications to their proper format. These proposed changes do not effect plant operations since they are administrative in nature.

Grammatical errors exist as well in the Technical Specifications. These errors include missing words and improper punctuation. Correction of these errors make the Technical Specifications easier to read. As a result, plant operations or l safety is not impacted.

Other errors are strictly administrative. These include expired pages and specifications. Removal of the expired pages and specifications eliminate the possibility of referring to an inapplicable specification. These proposed changes do not impact plant operations or safety.

, The proposed change to Table 3.1-2 (page 41) adds a requirement l that a manual scram trip function be operable in the run mode. This

! requirement has existed since the initial issuance of'the Technical l Specifications but was removed inadvertently when updating the table for a previous submittal. This proposed change returns the table to its original condition and does not impact plant operations or safety.

The proposed change to page 144 will correct a condition for l plant shutdown resulting from jet pump operability. The existing specification states that the reactor shall be placed in a cold condition if a jet pump is operable. This is in direct contrast to the preceding sentence which states that whenever the reactor is in a mode other than refueling that all jet pumps be operable. This proposed change will correct the specification to state that whenever a jet pump is inoperable that the reactor be placed in a cold condition. This proposed change will not affect plant operations.

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g Attachment II to JPN-88-023 SAFETY EVALUATION Page 14 of 16 The proposed changes to page 152 are needed to achieve consistency throughout the Technical Specifications. The elimination of "... coincident drywell pressure..." from specification 3.6.E BASES is consistent with the changes brought about as a result of a plant modification which eliminated the high drywell pressure permissive for ADS actuation from the ADS logic.

This change has previcusly been approved and issued by Amendment 84.

The inclusion of this change in the submittal for this amendment was inadvertently omitted. The proposed change has no impact on plant safety or operations.

Replacement of "low-low" in the same specification with "low-low-low" is necessary in order for the specification to agree with Table 3.2-2. This error has existed since the initial issuance of the Technical Specifications. The proposed change will have no impact on plant operations or safety.

The proposed change to page 244a replaces the "once/ week" surveillance frequency for the high pressure water fire protection system pumps to "once/ month". This change was erroneously introduced in the process of updating the page for a submittal which was approved and issued as Amendment 80. This submittal did not include a change to this surveillance frequency. This surveillance frequency of once/ month is a condition that has always existed.

This change will return the surveillance interval to its original value and does not impact plant operations or safety.

The proposed changes to the Technical Specifications do not change any system or subsystem and will not alter the conclusions of either the FS?.R or SER accident analysis.

Section IV EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIOM l

l The proposed changes to the James A. FitzPatrick Technical Specifications involve no significant hazards considerations.

They are all administrative or editorial in nature and include:

typographical errors; grammatical errors; and clarification of a specification. Operation of the FitzPatrick Plant in accordance with the proposed amendment would not involve a significant hazards consideration as stated in 10 CFR 50.92 since it would not:

l (1) involve significant increase in the probability or consequences of an accident previously evaluated. The intent of the proposed changes are to clarify and correct the Technical Specifications.

The changes are administrative and include: correction of misspelled words; deletion of expired pages; and correction of grammatical errors. There are no setpoint changes, safety limit changes, surveillance requirement changes, or limiting l

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Attachment II to JPN-88-023 SAFETY EVALUATION Page 15 of 16 conditions for operation. These changes have no impact on plant safety or plant operations. The changes will have no impact on previously evaluated accidents.

(2) create the possibility of a new of different kind of accident previously evaluated. The proposed changes are purely administrative in nature and involve only the correction of typographical and other errors. These proposed changes are intended to clarify and improve the quality of the Technical Specifications. This cannot create the possibility of a new or different kind of accident.

(3) involve a significant reduction in the margin of safety. The proposed changes correct errors which currently exist in the Technical Specifications. The changes are all administrative in nature and will clarify the specifications by eliminating errors such as typographical errors. These changes do not change any setpoint or safety limit changes regarding isolation or a3arns.

The proposed changes do not affect the environmental monitoring program. These changes do no affect the plant's safety systems and do not reduce any safety margin.

In the April 6, 1983 Federal Register (48FR14870), NRC published examples of license amendments that are not likely to involve a significant hazards consideration. Example (i) of that list is applicable to this change and states:

"A purely administrative change to the Technical Specifications:

for example, ... correction of an error...".

The Authority considers that the proposed changes can be classified as not likely to involve significant hazards considerations, since the changes are administrative or editorial in nature and do not involve hardware changes nor any changes to the plant's safety related structures, systems, or components. The proposed changes are designed to improve the quality of the Technical Specifications.

Section V Implementation of the Proposed Chance The proposed change will not adversely impact the ALARA, Security or Fire Protection programs at the FitzPatrick plant, nor will it impact the environment.

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

Attachment II to JPN-88-023 SAFETY EVALUATION Page 16 of 16

a. will not increase the probability or the consequences of an accident or nalfunction of equipment important to safety as previously evaluated in the Safety Analysis Report:

b. will not increase the possibility of an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report; c, will not reduce the margin of safety as defined in the basis for any technical specification;

d. does not constitute an unreviewed safety question; and

e. involves no significant hazards consideration, as defined in 10 CFR 50.92.

Section VII REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR).

2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).

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