ML20100M597

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Proposed Tech Spec Changes to Achieve Conformance W/ NUREG-0737 & Generic Ltrs 83-02 & 83-36
ML20100M597
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 12/06/1984
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20100M580 List:
References
RTR-NUREG-0737, RTR-NUREG-737 GL-83-02, GL-83-2, GL-83-36, NUDOCS 8412120380
Download: ML20100M597 (45)
v • d • e


Text

-Q ATTACHMENT-I PROPOSED TECHNICAL SPECIFICATION CHANGES RELATED TO NUREG-0737 ITEMS NEW YORK POWER AUTHORITY JAMES A.'FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 P

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.1 TEG NICAL SPECIFICATIONS TABE OF _00tmWIS F Pg 1.0 Definitions 1 LIMITING SAFETY SAFE 1Y LIMITS SYSTEM SEITINGS

, 1.' 1 Ebel Cladding Integrity _

2.1 7:

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1.2- Reactor Coolant System '2.2 27

. SURVEILIANCE .'

LIMITING C(2iDITIONS' FOR OPERATION REQUIREMENTS 3.0' General . :4.0'

, s 30

<3.1- Reactor Protection System 4.1 30f 3.2 ' Instrmentatiion 4.2 49 r .

A. Primary Containment .Isolatilan Functions A. 49 B.' Core and Contaiment Cooling Systems B. 49 Initiation and Control r C. Control Rod Block Actuation C. 50 D. Radiation Monitoring Systems - Isolation' D. 50 and Initiation Ebnctions E.~ Drywell'Imak Detection- E. '54 F. Surveillance Information Readouts- F. 54 G. Recirculation Pump Trip G. 54 H. Accident Monitoring Instrmentation H. 54 l 3.'3 Reactivity Control 4.3 88-A.. Reactivity Limitations- A. 88 B. Control Rods' B. 91 C. Scram Insertion Time C.- 95 D. Reactivity Anamalies D.  %

E Stan&y Liquid Control System 4.4 105 3.4 A. Normal Operation A. 105 B. Operation With Inoperable Couponents B. 106

~ C. Sodium Pentaborate Solution . C. 107 Amendment No. i s

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JAFNPP TABLE OF CONIENIS - (cont'd) 1 3.5 Core and Contairunent Cooling Systems ~ 4.5' 112 A. 112

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A. . Core Spray and LPCI Systems B. Containment Cooling Mode of the RHR -

B. 115 System.

C. HPCI System _

C. 117

, D. Automatic Depressurization System (ADS) D. 119 E. Reactor Core Isolation Cooling (RCIC), E. 121

' System .

F. Miniman Bnergency Core Cooling System F.' , 122 ,

. Availability G.' Maintenance of Filled Discharge Pipe G. 122 H. Average Planar Linear Heat Generation H. 123-Rate ,(APIRGR)

I. Linear Heat Generation Rate-(IBGR) I. 124 J.. '1hennal Hydraulic Stability l J. '

124a SURVEIIIANCE LIMITING CONDITIONS RR OPERATIONS REQUIRl!NENIS 3.6, Reactor Coolant System 4.6 136 l' A. Thennal Limitations A. 136 B. Pressurization Tenperature. B. 137 C. Coolant Chemistry C. 139-D. Coolant Imakage _

D. 141 E. Safety and Safety / Relief Valves E. 142a F. Structural Integrity-F. 144

-G.. Jet-Pteps G. 144-H. ' Jet _ Ptmp Flow Mismatch H., 145

1. Shock Suppressors (Snubbers) I. 145a i

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3.7 Conrair-nt Systens- 4.7f 165.

A.

A. Primary Contairunent 165 B. Standby Gas Treatment System B. 181 C. bemiary Cnnea% ment C. 184 i D. Primary Containment Isolation Valves D. 185-

- 3.8 Miscellaneous Radioactive Material Sources 4.8. 214 Amendnent No. ii

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- , : Q, ' s EJAMEP TABG OF 00tmWIS (cont'd)

,3.9 Auxiliary Electrical Systems 4.9 215

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. A. Normal and Reserve' A-C Power Systens - A. 215 B.-Bnergency A-C Power System- ' B. L216 X C. Diesel Riel , . ,

C. 218 D. Diesel Generator Operability _ D. . 220 E. : Station Batteries - ..

E. 221 F. LPCI MN Independent Power Supplies F. .222a G. Reactor Protection System Electrical. G. -222c l I

Protection Assenblies

- 3.10' Core Alterations 4.10 227

. A. : Refueling Interlocks .A. ~227 B. Core Monitoring B.- 230 C. Spent Fuel Storage Pool Water Level. C. 231 D. Control Rod and Control Rod Drive D. 231 Maintenance

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!3.11 ~ Additional Safety Related Plant Capabilities 4.11 237 A. Main Control Room Ventilation A. :237 7 B.: Crescent Area Ventilation B. 239' sC.' Battery Room Ventilation .

C.. '239

' D. Bnerpncy Service Water System D. 240 E. Inta w de-icing Heaters E. 242 3.12. - Fire Protection Systems 4.12 244a

'E. ~A. High Pressure Water Fire Protection A. 244a- .

Systen .

B.' Water Spray and Sprinkler Systems ' B. . 1 . 244e '

C. Carbon Dioxide Systens C. 244e D.-Manual Fire Hose Stations D. 244f

- E. Fire Protection Systems Snoke and Heat- E. > ..

y 244g Detectors-F. Fire Barrier Penetration Seals F.' 244g 15.0. Design Features 245~

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~5.1. Site 245-

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5.2 Reactor. 245 e..-_ _

p b-JAFNPP-TABE OF CNIENTS (cont'd)

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5.3 Reactor Pressure Vessel- 245 5.4l Contaiment 245

'5. 5 Ebel Storage .

245 5.6 Seismic Design 246 6.0- Administrative Controls 247

- 6.1 Responsibility 247 6.2 Plant Staff Organization 247 6.3 Plant Staff Qualifications 248 6.4 Retraining and Replacement Training 248 6.5 Review and Audit 248 I 6.5.1 Plant Operating Review Cocmittee (PORC) 249 6.5.2 Safety Review Comnittee (SRC) 250 6.6 Reportable Occurrence Action 253 6.7 Safety Limit Action 253 6.8 Procedures 253 6.9 Reporting Requirements 254a 6.10 Record Retention , 254g 6.11 Radiation Protection Program 255 6.12 Industrial Securit,.f Program 258 6.13 Bnergency Plan 258 6.14 Fire Protection Program 258 6.15 Enviromental Qualification - 258a l 7.0 References 285 iv Amm%tNo.[

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JAENPP e

LIST OF-TABLES TABLE TITLE PAGE

, 3.1-l' Reactor Protection System' (Scram) - 41 Instrunentation Requirement Reactor Protection System (Scram) 4.1-1 66 Instrunent fimetion Tests 4.1-2 Reactor Protection ~ System (Scram) 46 Instrument Calibration

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Instrumentation that Initiates Prhnery 9 3.2-1 64 Containment Isolation

. 3. 2-2 ..- Instrunentation that Initiates or Controls -

66 :

the Core and Conai==rit Cooling Systems 3.2-3 Instrunentation that Initates Control Rod Blocks 72 /

3.2-4 Radiation Monitoring Systems that Initiate ' 74 and/or Isolate Systems 3.2-5 Instrumentation that Monitors leakage Detection 75' Inside the Drywell-3.2-6 Surveillance Instrunentation 76 o

3.2-7 - Instrunentation that Initiates Recirculation Ptep Trip 77 3.2-8 Accident Monitoring Instrunentation - , 77a' j 4.2-1 - Mini == Test and Calibration Frequency for PCIS 78 4.2-2 Minimian Test and Calibration Frequency for Core 79 and Containment Cooling Systems

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4.2-3 Mininnan Test and Calibration Frequency for 81 Control Rod Blocks Actuation

- 4.2-4 Mini == Test and Calibration Frequency for Radiation 82 Manitoring Systems Jaendment No. v

_- .--a-_--__---__ - - _ - - - - - - - . - _ . - - _ - - - - --__,----.----------.----,----..---_-------_.--.w_--.--_ - - _ - _ _ _ _ - - - - - _ _ _ - . - - - - _ _ _ _ _ - - - - , - -

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JAFNPP LIST OF TABLES (Cont'd)

'4.2-5 Mininun Test and Calibration Frequency for Drywell 83 Imak Detection

.4.2-6 Mininun Test and Calibration Frequency for 84 Surveillance Instrm entation

-4.2-7 Mininun Test and Calibration Frequency for 86 Recirculation Ptmp Trip. ,

4.2-8 Mininun Test and Calibration Frequency for 86a l Accident Monitoring Instrunentation 3.6-1 Safety Related Shock Suppressors (Snubbers) 156b 4.6-1 Couparison of the James A. FitzPatrick Nuclear 157 Power Plant Inservice Inspection Program to ASME Inservice Inspection Code Requir ments 3.7-1 Process Pipeling Penetrating Primary Contaiment 198 4.7 Minimm Test and Calibration Frequency for 210 Containment Ibnitoring 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 Systen Tests 244q 4.12-2 Carbon Dioxide System Tests 244r 4.12-3 Manual Fire Hose Station Tests 244s 6.10-1 C aponent cyclic or Transient Limits 261 Amendment No. vi w _-_ _ _ _ _

JAFNPP LIST OF FIGURES FIGURE TITLE PAGE 3.1-1 Manual Flow Control 47a.

3.1-2 . Operating Limit FCPR versus T 47b 4.1-1 Graphical Aid in the Selection of an Adequate Interval 48.

Between Tests-

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4.2-l' Test Interval vs. Probability of System Unavailab'ility 87

^3.4-1 Sodim Pentaborate Solution Volixne-Concentration 110 Requirements 3.4-2 Saturation Tenperature of Sodium Pentaborate Solution 111 3.5-6 1RPUiGR Versus Planar Average Exposure Reload 2, 135d 8DRB283 3.5-7 MAPUlGR Versus Planar Average Exposure Reload 3, 135e P8DRB265L 3.5-8 MAPUlGR Versus Planar Average Exposure Reload 3, 135f P8DRB283 3.5-9 MAPulGR Versus Planar Average Exposure Reloed 4, 135g P8DRB284H l 3.5-10 MAPUiGR Versus Planar Average Exposure Reloads 4 & 5 135h P8DRB299 3.6-1 Reactor Vessel '1hermal Pressurization Limitations 163 4.6-1 Gloride Stress Corrosion Test Results at 500'F. 164 6.1-1 Management Organization chart 259 6.2-1 Plant Staff Organization 260 Amendment No. vii~

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Tablo 4.1-l'

' REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS <

INSTRUMENT CHANNEL GROUP FUNCTIONAL TEST MINIMUM FREQUENCY (3)

Mode Switch in Shutdown A Place Mode Switch in Shutdown Each refueling' outage. _

' Manual Scram A Trip Channel and Alarm: Every 3 months..

RPS Channel Test Switch JA Trip Channel and. Alarm Every refueling outage or after channel maintenance IRM :High Flux C Trip Channel and Alarm (4) Once per week during' refueling..or'startup and before each startup.

Inoperative. C Trip Channel and Alarm (4) Once per week during refueling or.startup-and before each startup.

APRM High Flux B Trip output Relays.(4) Once/ week..

Inoperative B' Trip output Relays (4) Once/ week ,

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Downscale B Trip output Relays (4) Once/ week:

Flow Bias- B Calibrate Flow Bias Signal (4) Once/ Month (1)

High Flux in Startup C Trip. Output Relays (4) Once per week during refueling or startup '

or Refuel and before each startup.: -

High Reactor Pressure B Trip Channel and Alarm (4) Once/ month. (1) Instrument check once per day :

High Drywell Pressure A Trip Channel and-Alarm Once/ month (1)

Raactor Low-Water Level (5) A Trip Channel and Alarm Once/ month (1)

High Water Level in Scram. A Trip Channel Once/ month (7)

D arge Instrument Volume High Water Level in Scram- B. . Trip Channel and Alarm (4) Once/ month Discharge Instrument Volume Main Steas Line High Radiation 'B Trip. Channel and. Alarm (4) Once/weekL 3

-Amendment 7 44 P

. JAFl@P 3.2 LIMITING CONDITIONS FOR OPERATION 4.2 SURVEILLANCE REQUIREDENTS "3.2 INSTRUMENTATION 4.2. INSTRUMElffATION

Applicability
Applicability: ,

Applies to the plant instrumentation which Applies to the surveillance. requirement of the i cither'(1) initiates and controls a protective instrumentation which either!(1) initiates and I- function, or (2) provides'information to aid the controls a. protective function,'or.(2).provides operator in monitoring and assessing plant- information to~ aid the operator in monitoring and status during normal and accident conditions. . assessing plant status during normal and accident 4 conditions, i

@jective: ~&jective:

i i To assure the operability,of the aforementioned 1: To specify the type and frequency of surveillance instrumentation.- a to be applied.to the aforementioned instrumen-l

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,tation.

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Specifications:

  • Specifications:

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j A. Primary Containment Isolation Functions A .' Primary Containment Isolation Functions i

! men primary. containment integrity ~is required, Instrumentation shall be functionally tested j the limiting conditions of operation for the and calibrated as indicated in Table 4.2-1.

instrumentation that initiates primary i containment isolation are given in Table _ . System logic shall be functionally tested as B.3.2-1. -

indicated in Table 4.2-1..

B. Core and Containment Cooling Systems - -:B. Core and Containment Cooling Systems'-

i Initiation fu Control Initiation and Control i The limiting conditions for operation for Instrumentation shall be functionally tested, j the instrumentation that initiates or controls calibrated, and checked as. indicated in Table

-4.2-2.

the Core and Containment Cooling Systems are.

A given in Table 3.2-2. 'Ihis instrumentation must -

be operable when the system (s) it initiates or Amencinent No.

- 49 i

- . _, . 4 . - - . _ . ._ _ . . m._.-. . , - m M;j N

  1. - m r JAFl@P

, 3.2 (cont'd) ,4'2.(cont'd) 2 E. Drywell' Leak Detection E. :Drywell Isak Detection ~

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. %e limiting conditions .of operation for. the . Instrumentation shall be calibrated and:

instrumentation that monitors drywell-leak. checked as indicated in Table 4.2-5.

detection are given;in Table 3.2-5.

F. Surveillance Information Readouts F. Surveillance Information Readouts:

.-W e limiting ~ conditions for the instrumentation Instrumentation.shall be calibrated and that' provide (s) surveillance information checked as indicated in-Table 4.2-6.

readouts:are given in Table 3.2-6.

G. Recirculation Pump Trip G. Recirculation Punp Trip

%e limiting conditions for operation for the Instrumentation shall be' functionally tested instrumentation that trip (s) the recirculation and calibrated as' indicated in' Table 4.2-7.

pumps e a means of limiting the consequences .

of 6 % ure to scram'during'an anticipated System-logic shall be functionally tested as transi..S are give.n in Table 3.2-7. indicated in Table 4.2-7.

H. Accid; st.toring Instrumentation H. Accident Monitoring Instrumentation

% e limiting conditions for operation of the

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Instrumentation shall be demonstrated instrumentation that provides accident ~ operable by performance of a channel check.

. monitoring are given in Table-3.2-8. and channel calibration as indicated in Table 4.2-8.

. Amendment.No. 54

'3.2 Bases (cont'd) 'JAFNPP The recirculation pump trip has been added at the suggestion of ACRS as'a means of limiting the consequences of the unlikely. occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event' falls within the

  • envelope of study events given in General Electric Company Topical Report, NEDO-10349, dated March, 1971.

Accident monitoring instrumentation provides additional 9

- information which:is helpful to the operator in assessing plant conditions following an accident by (1) providing.information needed to permit the operators to

, take preplanned manual actions to accomplish safe plant shutdown; (2) determining whether systems are performing their intended functions; (3) providing information to the operators that will enable them to determine the potential for a breach of the barriers to radioactivity release and if a barrier has been breached; (4) furnishing data for deciding on.the need to take unplanned action if an automatic or manually initiated safety system is not functioning properly or the plant is not responding properly to the safety systems in operations and (5) allowing for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of any problem. This instrumentation has been upgraded to conform with the acceptance criteria of NUREG-0737 and NRC Generic Letter 83-36.

1 Amendment No.

60

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JAFNPP TABLE 3.2-2 (Cont'd)

INSTRUMENTATION THAT INITIATES 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. Trip System (1) Trip Function Trip Level Setting Both Trip Systems Remarks 22 2 Condensate Storage 2 59.5 inches above 2 Inst. Channels Transfers RCIC pump Tank Low Level tank bottom suction to suppres-(= 15,600 gal, avail) sion chamber 23 24 25 1 Core Spray Sparger f.0.5 psid 2 Inst. Channels Alarm to detect to. Reactor Pressure core spray sparger vessel d/p pipe break 26 2 Condensate Storage .E 59.5 in. above 2 Inst. Channels Transfers HPCI pump Tank Low Level tank bottom suction to sup-(= 15,600 gal avail) pression chamber.

27 2 Suppression Chamber 6,'6 in. above normal 2 Inst. Channels Transfers HPCI pump High Level level suction to suppres-sion chamber.

28 1 RCIC Turbine Steam .6 282 in. H 2O psid 2 Inst. Channels Close Isolation-Line High Flow Valves in RCIC Subsystem.

Amendment No. f

/ 70a

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Y JAFNPP TAELE 3.2-8 ACCIDENT MONITORING INSTRUMENTATION .

NO. OF . MINIMUM NO. OF .

U' CHANNELS PROVIDED OPERABLE CHANNELS MEASUREMENT INSTRUMENT BY DESIGN ' REQUIRED ACTION RANGE

1. Stack High Range Effluent 2 1 'A 10-1 to 107 mR/hr.

Monitor

2. -Turbine Building Vent High. 2 1 A- 10-l'to 107 mR/hr. I Range Effluent Monitor
3. Radwaste Building Vent High ~2 1 A 10-1 to 107 mR/hr.

Range Effluent Monitor

4. Containment High Range 2 1 A 1 to 108 rads /hr.

Radiation Monitor *

5. Containment Pressure 2 wide range 1 A' 0 to 250 psig Transmitter .2 narrow range 1 A -5 + 5 psig
6. Drywell Level Transmitter 2 1 A 22 to 100 ft. (H2O)
7. Suppression Pool Level 2 1 A 1.7 to 27.5 ft...(H 2O)

Transmitter

8. Reactor Vessel Pressure 2 1 A 0 to 1500 psig Transmitter
9. Drywell Hydrogen 2 1 A 0 to 30% H2 Concentration Monitor
  • AtsG450 R/hr, closes vent and purge valves 77a Amendment No.

TABLE 3.2-8(cont'd)

ACCIDENT MONI'IORING INSTRUMENPATION NO. OF MINIMJM NO. OF CHANNELS PRWIDED OPERABLE CHANNEIS MEASUREMENT INSTRUMENT BY DESIGN -REQUIRED ACTION RANGE

10. Post-Accident Containment and Reactor Coolant Radioactivity Sanpling Conponents a) Gaseous Radioactivity 1 1 B 10-5 uCi/cc Sampling Tray Conponents to 106 uCi/cc b) Liquid Radioactivity 1 1 B 10 uCi/ml Sampling Tray Cmponents to 10 Ci/ml IUTES FOR TABLE 3.2-8 A. With the nunber of operable channels less than the required minimum, either restore the inoperable channels to operable status within 30 days, or: (1) initiate an alternate method of monitoring the appropriate parameter (s), or (2) be in a cold condition within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. 1. When the ability to provide a liquid and/or gaseous sample to the Post-Accident Sample Station has been lost, restore the capability within 7 days or be in at least a cold condition within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. When the ability to sample and analyze a liquid and/or gaseous sanple has been lost, within 7 days confirm that alternate arrangements for sampling and analysis of the required samples can be made available within 24 hrurs of the need to perform sampling and analysis. If alternate arrangements for sampling and analysis can not be confirmed within 7 days,be in the cold condition within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3. When operating in accordance with 2 above, restore conplete sanpling and analysis capability within 90 days. If complete sampling and analysis capability cannot be restored within 90 days, be in the cold condition within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No.

77b

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TABLE 4,2-2

' MINIMUM TEST AND CALIBRATICE FREQUENCY FCR CORE AND CONTAINM!NP COOLING SYSTEMS -

Instrument Gannel I} t ument Functional Test ' Calibration Frequency- Instrument Check ,

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1) Reactor Winter Level (1)- Once/3 months. Chee/ day-
2) .Drywell Pressure - (1) . .Once/3 months None
3) , Reactor Pressure "(1) Chce/3 months. None
4) Auto _ Sequencing. Timers WA Once/ operating cycle' None.-
5) ADS - LPCI or CS Pump Disch. .(1) Once/3 months None Pressure Interlock
6) Trip System Bus Power Monitors- (1) WA None
8) Core Spray Sparger d/p .(1) Once/3 months" Chce/ day
9) Steam Line High Flow (HPCI & RCIC) (1) Once/3 months . None
10) Steam Line/ Area High Tenp. (HPCI & RCIC) (1) Once/ operating' Cycle 'Once/ day
12) HPCI E. RCIC Steam Line Inw Pressure . (1) - Once/3 months' None
13) HPCI & RCIC Suction Source Levels -(1) Chce/3 months ' - None ~. g' 14)4kV Emergency Power IMder-Voltage Once/ operating cycle .Once/ operating cycle None Relays and timers
15) HPCI & RCIC Exhaust Diaphragm Pressure . (1)- Once/3 months None High . _
17) LPCI/ Cross Connect Valve Position .Chce/ operating cycle NA NA Note: See' listing of notes following Table 4.2-6 for the notes referred to herein.

Amendment'.No. -

79

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JAFNPP: -" '

TABLE 4.2 '

L iMINIMUM TEST AND CALIBRATION FREQUENCY FOR CONTROL ROD BLOCKS'ACTUATIONt -

INSTRUMENT CHANNEL INSTRUMEN* FUNCTIONAL TEST CALIBRATION INSTRUMENT CHECK (9) U:

1).APRM Downscale' (1) (3) Once/3 months Once/ Day'.

2) APRM - Upscale -(1) (3). ~0nce/3 months 1 Once/ Day'.
  • 3) IRM -_ Upscale (2) (3) (2). (2)

+.

4).IRM - Downscale (2) (3) (2) (2) 2

5) RBM - Downscale (1) ~(3) Once/3 months Once/ Day. 4
6) RBM - Upscale (1). (3)

Once/3 months once/ Day-

7) SRM - Upscale. .(2) (3) (2) '(2) ,
8) SRM'- Detector Not in Startup Position- ( 2)l (3) (2).

] 9)cIRM- Pectector Notdin Startup Position (2) -(3) (2)

10) Scram Discharge Instrument. Volume - Once/ month- (2) (3) .Once/ operating cycle (2). N/A

'High Water' Level LOGIC SYSTEM FUNCTIONAL TEST (4) (6) FREQUENCY ,

.1) System Logic Check ,0nce/6 months NOTE: See listing of notes following Table.4'2-6 . for the notes referred to herein.

AmendmentNo.), , 81 i

em n

i TABLE 4.2-8 liINIMUM TEST AND CALIBRATION PREQUENCY FOR ACCIDENT MONI'IORIIG INSTRUMENTATION INSTRUMENT INSTRUMENT FUNCTIONAL TEST CALIBRATION FREQUENCY INSTRUMENT CHECK

1. Stack High Range Effluent Once/ Operating Cycle Once/ Operating Cycle Once/ day Monitor
2. Turbine Building Vent High Once/ Operating Cycle Once/ Operating Cycle Once/ day Range Effluent Monitor
3. Radwaste Building Vent High 01ce/ Operating Cycle Once/ Operating Cycle Once/ day Range Effluent Monitor
4. Containment High Range Once/ Operating Cycle Once/ Operating Cycle Once/ day Radiation Monitor
5. Containment Pressure WA Once/ Operating Cycle 01ce/ day Transmitter
6. Drywell Level Transmitter WA 01ce/ Operating Cycle Once/ day
7. Suppression Pool Level WA O1ce/ Operating Cycle 01ce/ day Transmitter
8. Reactor Vessel Pressure Channel WA Once/ Operating Cycle Once/ day
9. Drywell Hydrogen WA Once/ Operating Cycle 01ce/ day Concentration Analyzer
10. Post-Accident Containment and Reactor Coolant Radioactivity Sampling Components a) Gaseous Radioactivity 01ce/ Operating Cycle Once/ Operating Cycle Onca/ quarter Sampling Tray Components b) Liquid Radioactivity Once/ Operating Cycle 01ce/ Operating Cycle Once/ quarter Sampling Tray Ccxnponents Amendment !b.

86a

3.5 (Cont'd) JAFNPP 4.5 (Cont'd) l E. Reactor Core Isolation Cooling E. Reactor Core Isolation Cooling j (RCIC) System (RCIC) System 1.- The RCIC System shall be operable 1. RCIC System testing shall be performed whenever.there is irradiated fuel as follows provided a . reactor steam in the reactor vessel and the reactor supply is available. If steam is not ,

l pressure is greater than 150 psig and available at the time the surveillance i

prior to a reactor startup from a cold test is scheduled to be performed, the condition, except from the time that test shall be performed within ten _ days the RCIC System is made or found to of continuous operation from the time be inoperable for any reason, continued steam becomes available, i reactor power operation is permissible I during the succeeding 7 days unless the Item Frequency system is made operable earlier provided j that during these 7 days the HPCI System a. Simulated Once/ operating l is operable., Automatic cycle Actuation

2. If the requirements of 3.5.E cannot be (and Restart *) l met, the reactor shall be placed in the Test cold condition and pressure less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. b. Pump Oper- Once/ month

! 3. Low power physics testing and reactor operator training shall be permitted c. Motor Oper- Once/ month with inoperable components as specified ated Valve in 3.5.E.2 above, provided that reactor Operability l coolant temperature is :$212*F.

d. Flow Rate Once/3 months
e. Testable Tested for Check Valves operability any time the reactor is in the cold condition exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if oper-ability tests have not been performed during the pre-ceding 31 days.
  • Automatic restart on a low water level signal which is subsequent to a high water level trip. l Amendment No. 4g 121

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

h-u C

E. Safety and Safety / Relief Valves E. Safety and Safety / Relief Valves-

1. During reactor power operating 1. At least one half of all conditions and prior to startup safety / relief valves'shall'be bench' from a cold condition, or whenever.

checked or replaced with bench checked reactor coolant pressure is greater valves once each operating cycle. The than atmosphere and temperature safety / relief valve settings greater than 212*F, shall be set as required in Specification

.the. safety mode of all '2.2.B. All valves shall be tested every.

safety / relief valves shall be two operating cycles.

operable, except as specified by Specification 3.6.E.2. The Automatic Depressurization System ~

Valves shall be operable as required-by Specification 3.5.D.

Amendment No. 13, ,- ,7 142a-

" ~ ,

w h PNPP - 4.6 (ctnt'd) s 3.6 (cont'd) '

I E. Safety and Safdty/ Relief.V_alves E. Safety and Safety / Relief. Valves' .

i
1. During reactor power operating ..'

.l. .At least one half of all k

conditions and prior to startup safety / relief valves shall l

f rom a cold condition, or whenever- be bench checked or re- -

j reactor coolant pressure is placed with bench checked valves'once each operating greater than atmosphere and g cycle. .The safety / relief temperature. greater than 212 P, valve settings shall be set 5 . the safety mode of all safety / relief ~ as required in Specification valves shall be. operable, except i 2.2.B. All valves shall be j as specified by Specification tested every'two operating

3. 6. E. 2.- The Automatic Depressurization cycles.

l j

system valves shall be operable as required 1 by Specification 3.5.D. . l l

2. Reactor operation may continue with one i safety / relief valve inoperable. '

From and after the date that two safety / relief valves are made or found inoperable, continued reactor operation .1-j l is permissible only during the l

succeeding ~J days,unless one valve i is made operable. .

1

  • I \

e i /

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l I I i (

I l

l Amendment No. 70' 142b This page is effective until -

j the October 1982 outage.

l -l-DE LE TE 4

i

~

i

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

2. a. From and after the date 2. At least one safety / relief valve shall be ,

that the safety valve disassembled and inspected once/ operating function of one safety / cycle.

relief valve is made or ,.

found to be inoperable, continued operation is permissible only during the succeeding 30 days unless such valve is made operable sooner.

b. From and after the time that the safety valve functicn on two safety /

relief valves is made or found to be inoperable, 3. The integrity of the nitrogen system and continued reactor operation components which provide manual and ADS is permissible only during actuation of the safety / relief valves the succeeding 7 days unless shall be demonstrated at least once every such valves are sooner made 3 months.

operable.

4. An annual report of safety / relief valve
3. If Specification 3.6.B.1 and 3.6.B.2 failures and challenges will be sent to the are not met, the reactor shall be placed NRC in accordance with Section 6.9.A.2.b 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 />.
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 isd 212*F and the reactor vessel is vented or the reactor vessel head is removed.

AmendmentNo.4,7[ 143

= _

m _ m .

oAFNPP 4.6 (c nt'd) 3.6 (ct:nt'd) .

2. 'At least one safety / relief ,
3. If Specification 3. 6.E. l and 3. 6.E. 2 valve shall be disassembled are not met the reactor shall be and inspected once/ operating placed in a cold condition within - cycle. . [

24 hr..

3. Deleted
4. Low power physics testing and ' reactor  !

operator training shall be permitted 4. The integrity of the nitrogen i with inoperable components as specified system and components which in 3.6.E.2, and provided that reactor provide manual and ADS coolant temperature is 4 2120F and actuation of the safety / relief the reactor vessel is vented or the ,

valves shall be demonstrated i vessel head is removed. at least once every 3 months.

I l

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Amendment No. 70 143a This page is effective until the October 1982 outage.

DE LETE; ,

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_l.

JAFNPP,b }. ,_./- ,

]

3.6' (cont'd) .

I - [ ,

j E. If, for a period of longer than' 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, ttd -

i temperature of any safety / relief discharge

  • j pipe is more than 400F above its steady state '

value, or the acoustical monitor reading of ,

any safety / relief valve discharge pipe is more' .-

than 3 times. greater than. its steady state value, the following actions shall be taken:

I a. a report shall beLissued in accordance with 6.9.A.4.1 which addresses the actions that have been taken or "a schedule of I actions to be taken.

I i

b. an engineering evaluation shall be per- -

formed justifying continued operation

  • for the corresponding increase in tem-
  • perature or acoustical monitor reading.
c. the affected safety / relief valve shall

' be removed at the next cold shutdown -

of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, tested in the as- '

found condition, and recalibrated as -

necessary prior to reinstallation.

I 3

d. NRC approval of.the engineering ,

evaluation specified in 3.6.E.5'.b above -

shall be obtained prior to continuing power operation for more than 90 days after the initial discovery of the 40 F increase in temperature or the factor of 3 increase ,

in acoustical monitor reading.

The steady state values of temperature and acoustical monitor readings shall be as

. measured af ter 5 days of steady state power operation.

143b This page is ef fective until the Amendment No. 70 October 1982 outage.

. 0EE LE' TEE - -

_m b%

3.7-(cont *C) JAFMPP '4.7 (cont'd)J U

]

,When the primary contairument 'is:

iinerted,-it shallLbe continuously. ~

monitored for gross leakages by review

-of the.inerting system makeup .

requirements. The monitoring system.

may be Ltaken .out .of service for _ <

maintenance, but shall-be returned - -

to service as soon.as possible.

.(--

4. . Pressure Suppression Chamber- 4. Pressure Suppression Chamber -

Beactor Building vacuus: Breakers - Reactor Building Vacuum Breakers

a. Except as specified in -
a. The pressure suppression 3.7.A.4.b below, two chamber-reactor building Pressure Suppression Chamber-Reactor vacuum breakers and associated Building Vacuum. Breakers shall-be instrumentations. including operable at all times when,the setpoint shall be checked for primary containment integrity is _ proper operation'every three required. The setpoint of the months.  : ,- '

differential pressure,instruentation

. which actuates the pressure suppression chamber reactor -building vacuum s l

breakers shall be A 0.5 psi external'

! pressure. -

b. From and after the'date that one-of the pressure suppression chamber -

l reactor building vacuum breakers is made or found to be inoperable for any reason, reactor operation =is permissible only during the -

succeeding 7 days,-unless Amen h nt -177,

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

i I

l JAFNPP 1

' TABLE 4.7-2 .

1, EXCEPTIOtt TO TYPE C TESTS ,

certain Type C tests will be perfo w d 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 globe valves -- '

and D C, and D pressurized in reverse direction and meas-29-AOV-86A, B, urement of. leakage will be' equivalent to C, and D results from pressure appliv1-in the same. -

direction as when the valves would be' required to perform its safety function.

Therefore, pressure will be applied between

-the' isolation valves and leakage measured.'

A water. seal of 25 psig will be used on the. ,

inboard valve to determine the outboard valvels Lleak rate. (limit-11.5 SCFH at.25

~

psig) '

X-10 BCIC 13-MOV-15 'See X-25 (27-A0V-131A, B)

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

s X-25 Dry Well Inerting 27-AOV-112 This valve is a butterfly valve pressur-l CAD and Purge ization in reverse direction and measurement I 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.

t

! .L I

l Amenchment No.

l 211 i

. , _ _ , -s_

r- ,

i JAFNPP I

l l TABLE 4.7-2

( EXCEPTION TO TYPE C TESTS (CONTINUED) l Penetration System Valve Local Leak Rate Test Performed ,

X-25 Dry Well Inerting 27-AOV-131A These valves will be tested in the reverse.

CAD and. Purge 27 AOV-131B direction, since the system was not designed for test pressure to be applied in the same l-direction as that when the valve would be required to perform its safety function.

Basis - The pressurization direction was not a requirement at the time of plant design.

g.

X-26 A/B Dry Well Inerting 27-AOV-ll3 See X-25 (27-AOV-ll2)

CAD and Purge 27-MOV-122 This globe valve will be tested in the reverse direction. See X-25 (27-AOV-131A, B)

T Amendment No.

211a I

JArwPP Tams w 4.7-2 tCow?'D1 l penetration System valve Ineal trak Rate Test Performed ,

27-50V-120ls See X-2% (27-ADV-131A, 3) 27-50V-1213

. 27-SOV-1223 .

X-31 ad Dry well Enerting 27-Sov-1258 See X-25 (27-ADV-131A)

CAD and rurge (bat. Spray 10-tNN-31A this valve will be pressurised in the T-39A reverse direction and leakage aura.wred.

See X-25 (27-SOV-131A, 3) ,

2-393 Omt. Spray 10-MOV-31A see X-39A ,

2LRT VSM-1007 See X-25 (27-A0V-131A, al X-4 5 See X-25 (27-AOV-131 A, p)

X-59 Dry teell Inerting 27-50V-123A Cad amt Purge X-202 Turus vacuum AOV- 101 A/B See X-25 (27-A0V-112)

Dreakers See X-25 (27-A0V-131A, 9)

X-203A Dry Well Inerting 27-50V-1995 CAD aml Purge ,

X-203s Dry Well Inerting 27-30V-1244 See X-25 (27-A0V-131A)

CAD and Furge X-205 Dry Welt Knerting 27-A0V-117 See X-25 (27-Auv-1123 CAD and Purge 21-MOV-117 See X-25 (27-Nov-113)

X-210 A/S RCIC, peut E111 not be tested as lines are water

- sealed by suppression c>=hae water {

See X-25 (27-ADV-131A, b)

X-211A RNR 10-P!OV-38A .This valve will be tested in the reverse direction. See X-25 (27-ADV-131A, 9)

X-211a ana 10-eOV-3em This valve will be tested in the reverse direction.

X-212 RCIC 13-MOV-134 See X-25 (27-AuV-13%A/B)

X-21s 11Jt? VSM-1997 See X-45 (27-AuV-131 A/W)

X-220 Dry well Inerting 27-A0V-116 See X-25 (27-A0v-112)

CM and Purge 27-50V-132A See X-25 (21-A0V-131A/B) 27-50V-132a X-222 NPCI See X-210 A/B RNR See X-218 A/18 X-224 X-225 aun See X-210 A/B 212 A=enament/

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6.0 ADMINISTRATIVE CONTROLS Administrative Controls are the means by which plant opera-tions are subject to management control. Measures specified in this section provide for the assignment of responsibilities, plant organization, staffing qualifications and related re-quirements, review and audit mechanisms, procedural controls and reporting requirements. Each of these measures is necessary to l ensure safe and efficient facility operation.

6.1 RESPONSIBILITY The Resident Manager is responsible for safe operation of the plant. During periods when the Resident Manager is un-available, the Superintendent of Power will assume his respon-sibilities. In the event both are unavailable, the Resident Manager may delegate this responsibility to other qualified supervisory personnel. The Resident Manager reports directly to the Executive Vice President Nuclear Generation, as shown in Fig. 6.1-1.

6.2 PLANT STAFF ORGANIZATION The plant staff organization is shown graphically in Fig.

6.2-1 and functions as follows:

1. A licensed senior reactor operator shall be on site at all times when there is fuel in the reactor.
2. In addition to item 1 above, a licensed reactor operator shall be in the control room at all times when there is fuel in the reactor.
3. In addition to items 1 & 2 above, a licensed reactor operator and a licensed senior reactor operator shall be readily available on site whenever the reactor is in other than cold condition.
4. Two licensed reactor operators shall be in the control room during start-ups and scheduled shutdowns.
5. A licensed senior reactor operator shall be respon-sible for all movement of new and irradiated fuel within the site boundary. A licensed reactor operator will be required to manipulate or directly supervise the manipulation of the controls of all fuel moving equipment, except the reactor building crane. All fuel movements by the reactor building crane, except new fuel movements from receipt through dry storage, shall be under the direct supervision of a licensed reactor operator. All fuel movements within the core shall be direct 1 monitored by a member of the reactor analyst group.(a Amendment No. d', 6 , d 247 SEE NEXT PAGE FOR FO(YrNOTE
6. .

In cdditicn to ittma 1, 2 & 3 cb:vo, two cdditional op;rctcra ChOll bo rCCdily cvailcblo Cn cito wh;n:ver tho roactor 10 in other than cold shutdown. During cold shutdown, an additional operator shall be readily available on site.

7. An individual who is qualified in radiation protection shall be on site when fuel is in the reactor.
8. In the event of illness or absenteeism up to two (2) hours is 4 allowed to restore the shift crew or fire-brigade to normal complement.
9. A Fire Brigade of five (5) or more members shall be maintained on site'at all times. This excludes two (2) members of the minimum shift crew necessary for safe shutdown and any personnel required for other essential functions during a fire emergency.
10. A Shift Technical Advisor shall be on site and readily available to the control room except during the cold shutdown or refuel mode.
11. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and maintenance personnel who are working on safety-related systems.

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed

a. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.
b. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hout period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period, all excluding shift turnover time.
c. A break of at least eight hours should be allowed between work periods, including shift turnover time.
d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the Resident Manager or the Superintendent of Power, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting Amendment No. , , -248-

'tha dzviction.- Centrols chall bs includsd'in tha procedurso cuch that individual overtima-choll b3 review d monthly by1ths R2sidsnt1Manngar or hio docignse to accure that excessive hours have not'been assigned. Routine deviation from the above guidelines is not authorized.

6.3 PLANT STAFF-QUALIFICATIONS 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 N18.1-1971 for comparable positions; except for the Radiation l and Environmental Services Superintendent who shall meet or

. exceed the qualifications of Regulatory Guide 1.8, September 1975 and the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents. Any deviations will be justified to the NRC prior to an individual's filling of one of these positions.

G.4  : RETRAINING AND REPLACEMENT TRAINING A training program shall be maintained under the direction of the Training coordinator to assure overall proficiency of the plant staff organization. It shall consist of both retraining and replacement training and'shall meet or exceed the minimum requirements of Section 5.5 of ANSI N18.1-1971. l The retraining program shall not exceed periods two years in length with a curriculum designed to meet or exceed the

,, requalification reqdirements of 10 CFR 55, Appendix A. In addition fir'e brigade training shall meet or exceed the require-ments 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 review groups for the review and audit of plant j operations have been. constituted. One of these, the Plant Operating Review Committee (PORC), is an onsite group. The other is an independent review and audit group, the offsite Safety Review Committee (SRC).

6.5.1 PLANT OPERATING REVIEW COMMITTEE (PORC) f (A) Membership t

L The PORC is comprised of the Resident Manager (Chairman),

Superintendent of Power (Vice Chairman), Operations Superintendent, Maintenance Superintendent, Technical Services l

l Superintendent, Instrument and Control Superintendent,

! Radiological and Environmental Services Superintendent and i Reactor Analyst. Special consultant to provide expert advice

! may be utilized when the nature of a particular problem dictates.

, Amendment No. 248a I

i I

(A) ,

ROUTINE REPORTS AND REPORTAFLE OCCURRENCES (Continu0d)

1. STARTUP REPORT (Continued)
b. Startup Reports shall be submitted within (1) 90 days following completion of the startup test program, or (2) 1 90 days following' resumption or commencement of  ;

commercial power operation, or whichever is earliest. I If the Startup Report does not cover both events, i.e., j completion of startup test program and resumption or j commencement of commercial power operation, supplementary reports shall be submitted at least every three months until both events are completed.

i

2. ANNUAL REPORTS
a. Annual Occupational Exposure Tabulation A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposuresgreaterthan100 mrem /yrandtheirassoc{qtedman rem exposure according to work and job functions, _/e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe i maintenance), waste processing, and refueling. The dose l assignment to various duty functions may be estimates based i on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of  ;

the total whole body dose received from external sources, l shall be assigned to specific major work functions.

b. Annual Report of S/RV Failures and Challenges j An annual report of safety / relief valve failures and ,

challenges will be submitted prior to March 1 of each year.  !

An S/RV failure is defined to be any one of the following:

(1) Failure of a valve to actuate when automatically or i manually signaled to do so, for any reason, including testing, (2) Failure of a valve to close when the actuation  !

signal is removed, (3) Spontaneous operation of a valve in l the absence of an actuation signal, and (4) Determination l that a valve is, or was inoperable in any operating mode except cold shutdown or refueling. An S/RV challenge is  :

defined as an automatic or manual signaling of the valve to ,

actuate for the purpose of controlling physical parameters of '

the primary coolant system. Test actuations of S/RV'n are not considered challenges and will not be reported unless the test results in a failure, i

3. MONTHLY OPERATING REPORT l A report providing a narrative summary of facility operating experience, major safety-related maintenance, and other pertinent information, should be submitted not later than the 15th of each month following the calendar month covered to the USNRC Director, Office of Management Information and Program Control.

1/ This tabulation supplements the requirements of S20.407 of 10 CFR Part 20 AmendmentNo.jF 254-b

RE51MNT VICE PRESIMNT

  • 1 MANAGER 45ALITY A558SAllCE

{ AN8 STAFF (WPS) l m .

PLANT SPERATING REVIEW CeletalTTEE I SUPERINTERMNT eF PoWEa j AND STAFF .

l

.I I

IIAINTENANCE I&C SPERATiteIS '" "*

TECNNICAL SECutlTY/5AFETY 58PT. & 58PT. & SUPT.(5R0) EntmENTAL L SERVICE 5 Fitt P90TECTl001 STAFF STAFF AND STAFF & STAFF M L & STAFFe T STAFF o

BEACTOR ANALT5T $NIFT TECNNICAL ADV1508 SWIFT I (580)

T9AINING SUPERINTEletENT SFFICE I ANS STAFF IBANAGER A5515 TANT SutFT StPERVI5et (580)

I SENiet NGCLEAS IIEW YetE P9WER AUTilealTY Sat- SENlet MACTSE SPERAftt GPERATea (ge) 3AaeES A.FITZPATRICE NUCLEAR POWER PLAIIT B4 - BEACTSE SPERATOS y PLA81T STAFF 00GANIZAil001 NUCLEAR C006TBGL e Respessihelity Ser ,.i;.a - and moeiterleg SPERATOS (RS) of the fire protecties program.

I I I AWEILARY BSCLEAR SPERAT085 SPERAT995 FIGURE 6.2-1

g .

1 ATTACHMENT II i

Safety Evaluation r-for Technical Specifications Related to NUREG-0737 Items (JPTS-84-011)

(JPTS-84-015) r i

1 New York Power Authority James A. Fitzpatrick Nuclear Power Plant Docket No. 50-333 t t

t.

I. Description of the Proposed Changes The proposed changes to the FitzPatrick Technical Specifications relate to items identified in NUREG-0737, ,

" Clarification of TMI Action Plan Requirements."

Specifically, 'the following changes are being proposed:

The Table of Contents on pages i to iv has been retyped to restore margins. In addition, the following changes are incorporated. Page numbers refer to the retyped revision of the Table. On page 1, the following line is added:

3.0 General 4.0 3.0 On page 1, the page number for Section 3.1 is changed from "30" to "30f."

On page 1, " Protective Instrumentation" is changed to

" Instrumentation."

On page 1, the following-line is added:

H. Accident Monitoring Instrumentation H. 54 On page 11, page number "136" is added for Section 3.6.

On page lii, add the following lines G. Reactor Protection System Electrical G. 222c Protection Assemblies on page lii, add the following:

3.12 Fire Protection Systems 4.12 244a A. High Pressure Water A. 244a B. Water Spray /and Sprinkler B. 244e Systems C. Carbon Dioxide Systems C. 244e D. Manual Fire Hose Stations D. 244f E. Fire Protection Systems E. 244g Smoke & Heat Detectors F. Fire Barrier Penetration F. 244g Seals on page iv, the page number for Section 6.5 is changed from

) "249" to "248."

On page iv, the page number for section 6.9 is changed from "254A" to "254a."

l on page iv, the page number for Section 6.10 is changed from "254H" to "254g."

l 1 L

On page iv, the following line is added:

6.15 Environmental Qualification 258a On page v, the List of Tables is revised to include Table 3.2-8 On page vi, the List of Tablas is revised to include Tables 4.2-8, 3.12-1, 3.12-3, 4.12-1, 4.12-2 and 4.12-3, and to delete Table 6.11-1.

On page vii, a typographic. error in the fuel type listed for-Figure 3.5-9 is corrected by changing "P8DRB284L" to P8DRB284H."

On page 44, add Note number. (4) to " Trip Channel and Alarm" for High Water Level in Scram Discharge Instrument Volume,"

Group "B."'

On page 49, in Sections 3.2 and 4.2, change titles from

" Protective Instrumentation" to " Instrumentation." Revise

" Applicability" paragraphs to add instrumentation which provides information to aid in monitoring and assessing plant status. In " Objective" paragraphs change " protective" to

" aforementioned."

On page 54, the following is added to Section 3.2:

H._ Accident Monitoring Instrumentation The limiting conditions for operation of the instrumentation that provides accident monitoring are given in Table 3.2-8.

On page 54, the following is added to Section 4.2:

H. Accident Monitoring Instrumentation Instrumentation shall be demonstrated operable by performance of a channel check and channel calibration as indicated in Table 4.2-8.

On page 60, the following is added to Section 3.2:

Accident monitoring instrumentation provides additional information which is helpful to the operator in assessing plant conditions following an accident by (1) providing information needed to permit the operators to take preplanned manual actions to accomplish safe plant shutdown; (2) determining whether systems are performing their intended functions; (3) providing information to the operators that will enable them to determine the 2

potential for'a_ breach of the barriers to'

. radioactivity release and if a barrier has been breached;~(4) furnishing data for-deciding on the need to take unplanned action if an automatic or

, manually initiated safety system is not functioning properly or the plant is not responding properly to- ,

the safety systems in operation; and (5) allowing for  !

early indication of the need to initiate. action necessary to protect'the public and for an estimate of the magnitude of any problem. This instrumentation has been upgraded to conform with the acceptance eriteria of NUREG-0737'and NaC Generic-Letter-83-36. ,

~0n page 70a Table 3.2-2, " Instrumentation that Initiates or Controls the Core andiContainment Cooling System" .is revised by adding a new Item 22 condensate storage tank low level.

For Item No. 22, the following information'is added: minimum

-nunber of operable instrument channels per trip system, trip function, trip level setting, and total number of instrument channels provided by design for both trip systems. . Remarks

-for Item 26 are revised to read, " Transfers HPCI pung suction to suppression chamber."

New pages'77a and 77b contain a newly created Table 3.2-8,

" Accident Monitoring Instrumentation".

On page 79 Table 4.2-2, " Minimum Test and Calibration Frequency for Core and Containment Cooling System," is revised to include RCIC Suction Source Levels. Item 13 is, therefore, changed to read "HPCI & HCIC Suct. ion Source Levels".

On page 81 "3" is added to item 10..

New page 86a contains newly created Table 4.2-8, " Accident.

Monitoring Instrumentation".

On page 121, Section 4.5.E.1.a is revised to read " Simulated

.~ Automatic Actuation (and Restart) Test *".

on page 121, the following is added: "* Automatic restart on a low water level signal which is subsequent to a high water

-level trip."

v on page 142a the note in the lower right corner concerning effective'date is deleted. Page 142b is deleted.

On page 143, in Section 4.6.E, item 3 and the note in the lower right corner are deleted, item 4 is renumbered, and the following new item is added:

4. An annual report of safety / relief valve failures and w challenges will be sent to the NBC as specified in Section 6.9.A.2.b.

3

l.

Pages 143a and 143b are deleted.

On page 177 change " vacuum breakers shall be 0.5 paid" to-

-- " vacuum breakers shall be d 0.5 psi external pressure."

'On page 211 the.last phrase in the remarks for penetrations

-X7A, B,' C.E.D is revised to read, " Limit 11.5 SCFH at 25 psig)."

. On.page 211, the.' note at the' bottom of ,the page concerning cycle.3 and the superscript referring to it are deleted.

10n page 211, " Table 3.7-2" is changed to " Table 4.7-2". A portion of the description of penetration X-25 and all of the

. description of penetration X-26 are moved

  • A new page 211a.

~

on page 211a, the word " test" is added to remark for X-25 and two lines which are no longer applicable are deleted.

On page 211a, for penetration-X-26 A/B, CAD and Purge system valve 27-MOV-ll3 is changed to 27-MOV-122.

'On page 247,; Section 6.0,' "are" is changed to ."is" in the sixth line.

,.The.following item is added to Section 6.2 (page 248):

11.. Administrative procedures shall be developed and implemented to limit the working hours of unit staff-who perform' safety-related functions; e.g., senior D reactor operators, reactor operators, health physicists,. auxiliary operators, and key maintenance personnel.

Adequate shift coverage'shall'be maintained without routine heavy use of overtime.1 The objective shall be to .have operating personnel work a normal' 8-hour day,

-/ ' '40-hour week while the plant-is operating.. However,.in

the event that unforeseen problems require substantial amounts of overtime to be used or during' extended Jperiods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed
a. An individual should not be permitted to work more than.16_ hours straight,. excluding shift turnover time.

'b.'An individual should not be permitted to work more ,

than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> iri any- seven day period, all excluding shif t turnover time.

4

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c. A break of at least eight hours should be allowed between work periods, including shift turnover time.
d. Except during extended shutdown periods, the use of

~ overtime should be considered on an individual basis and not for the entire staff on a shif t.

Any deviation from the above guidelines shall be authorized by the Resident Manager or the Superintendent of Power, or higher-levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Resident Manager or his designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

In Section 6.3 on page 248a correct " ANSI NI8.1-1971" to " ANSI N18.1-1971".

In Section 6.4 on page 248a correct " ANSI NI8.1-1971" to " ANSI U18.1-1971".

In Section 6.5 on page 248a correct "seperate" to " separate".

On page 254-b, item 2 is reformatted to read as follows:

'2. ANNUAL REPORTS

a. Annual occupational Exposure Tabulation A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions,1/e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than'20% of the individual total dose need not be accounted for. In the aggregate, at

~* . least 80% of the total whole body dose received from external sources, shall be assigned to specific major work functions.

b. Annual Report of S/RV Failures and Challenges An annual report of safety / relief valve failures and challenges will be submitted prior to March 1 of each 5

e c_

a 1

.  :.  ?

year. . An S/RV failure-is defined to be any one of the

.following: (1) Failure of a valve to actuate when automatically or manually signaled to do so,.for any

. reason, including testing,.(2) Failure of a valve to close when the actuation signal is removed, (3)

. Spontaneous operation of a valve in the absence of an

' actuation signal, and (4) Determination that a valve is, or was inoperable in any operating mode except cold shutdown or refueling. An S/RV challenge is-defined as'an automatic or manual signaling of the valve to actuate for_the purpose of. controlling _

i, physical parameters of the primary coolant system.

Test actuations of S/RV's are not. considered l ' challenges and will not be reported unless the' test cresults in a failure.'

W e wording in paragraph 2a is not changed. The wording in paragraph 2b is new.

On page 260,'" Asst. Shift' Supervisor" is added to Fig. 6.2-1, and "Vice President Quality Assurance (NYO)".is changed to "Vice' President Quality Assurance (WO)."

+

The proposed changes on pages i through 44 are editorial in nature, y

a II. Purpose of the Proposed Changes The proposed changes on page 49, 54, 60, 77a, 77b and 86a are necessary as a result of the following NUREG-0737 items.-

II.B.3 Postaccident Sanpling Capability -

II.F.1.1 Noble Gas Effluent Monitor--

i. II.F.1.3 Containment High-Range Radiation Monitor [

i- II.F.1.4 Containment' Pressure Monitor II.F.1.5 Containment Water Level Monitor

  • II.F.1.6 Containment Hydrogen Monitor The proposed changes on pages 70a and 79 are necessary to be consistent with NUREG-0737 Item II.K.3.22.~

The proposed changes on pages 121 and 121a are intended to l' ' incorporate in Technical Specifications the requirements of

! NUREG-0737-Item II.K.3.13 and Reference 1 for automatic restart .

L of the RCIC System on low-low reactor water level following L trip of the system on high reactor water level. .

L The proposed deletion of item 3 of Section 4.6.E is necessary since the safety / relief valves have been modified and do not have a bellows arrangement.

We renunbering of item 4 is editorial in nature.

l l-

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- , ,. ._ , , _ _ . r- - . , _ ,- .. -

Th9 proposed cddition to paga.14'3 is intendsd to incorporato

-in-Tschnical1Spacifications the itsms diccucacd in NUREG-0737,-  ;

Item II.K.3.3, and Reference 1 concerningLsafety relief valvo  ;

challenges.and/or failures.

The proposed. change to the: title of Table 4.7-2 and the.

^

deletion of note (a).and its superscript are editorial in

. nature.

The proposed change to the content of Table 4.7-2 incorporates in Technical Specifications the items discussed in

. 1NUREG-0737, Item II.E.4.1, and Reference 1 concerning the ..

addition of containment isolation valves.

-The proposed change to Section 6.0 is editorial in nature.

The. proposed changes to Section 6.2_are intended to incorporate in Technical Specifications the. items. discussed in-NUREG-0737, Item I.A.l.3.2, and Reference 1 concerning shift c . manning. 1 1 The proposed changes to Section 6.2 on page 248 relate to ,

overtime policy. This policy has been in effect at the JAP NPP as per Plant Standing Order #26. It is now being incorporated into the Technical Specifications as required by Reference 1.

The proposed ' changes to Sections 6.3, 6.4 and 6.5, are

editorial.in nature. j III. Impact of the Proposed Changes
i. The additional. instrumentation described,in these proposed j Technical Specification changes should improve safety at the ,

FitzPatrick plant by providing' additional information which ik ~

helpful to the operator in assessing. plant conditions following an accident. The proposed changes related to overtime restrictions should improve safety by limiting-overtime for safety related functions. .The proposed changes-related-to additional reporting,should improve safety by increasing reporting of safety and relief valve challenges. .

The editorial changes proposed will have no impact since their-

~

only purpose is to correct errors in the Technical Specifications as currently written.

The Commission has provided guidance concerning the

. application of the standards.forimaking a "no significant hazard considerations" determination by.providing certain-

, examples in the Federal Register (F.R.)~Vol. 48, No.-67 dated April 6,'1984,' page 14870. The proposed. changes.all match at-

'least one of these examples.

L 7

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. The addition proposed to page 143 "and the change proposed on

~

page 254b were identified by the NRC in Reference 1 as needed to be consistent with NUREG-0737_ Item II.K.3.3, " Reporting Safety and Relief Valve-Failures-and Challenges." The proposed changes are, therefore, considered to match Commission example.(i), "A purely administrative change to Technical. Specifications; for example, a change to achieve consistency throughout the Technical Specifications,

. correction of an error, or a change in nomenclature."

.The deletion of item 3 of Section 4.6.E on page 143 and the renumbering of the items in that section was necessary as a

' result of previous changes to achieve consistency throughout the Technical Specifications. The proposed changes are, therefore, considered to' match Commission example (1).

The proposed changes on pages 142a, 142b, 143a, 143b, and the

. minor changes on page 143 are also editorial and considered to match commission example (1).

The proposed change on page 177 would make the wording consistent with Standard Technical Specifications, and is,

, therefore, considered to match example (1).

The proposed changes on pages i through 44, page 81, and on pages 211, 212, and 213 the pro. posed change to the title of Table 4.7-2 and the deletion of note (a) and its superscript can be classified as not:likely to involve significant hazard considerations, since these' changes are editorial in nature, and,.~therefore, fit Commission example (1).

On page 211, the change to penetrations X7A, B, C and D is typographical and matches example (1).

The changes to page 211a are typographical and match example (1).

The proposed change to the content of Table 4.7-2 was identified by the NRC in Reference 1 as a change needed to be

. consistent with NUREG-0737, Item II.E.4.1, " Dedicated Hydrogen Penetrations." This change is not expected to result in any

~

major change to facility operations. Since it is similar, therefore, to Commission example (vii), "A change to make a license conform to' regulations where the license change results in.very minor changes to facility operations," it is determined to involve no significant hazard considerations.

The proposed change to Section 6.0 on page 247 can be classified as not likely tx) involve significant hazard considerations since it is editorial in nature. It,

~ therefore, fits Commission example (1).

i 8

< Thp propossd now Itcm 11 in Snction'6.2 on pago 248 wne

~

identified by the NRC in R3forence 1 as a chcngo nocdad to ba l

consistent with NUREG-0737 Item I.A.1.3.1, Limit overtimo.

l The proposed change is considered to match Commission examples

-(vii) and (ii). Therefore, the proposed change is determined

. to involve no significant hazard considerations.

The proposed changes to Sections 6.3, 6.4 and 6.5 can be '

classified as not likely to involve significant hazard considerations,'since they are editorial in nature, and, therefore, fit Commission example (1).

The changes proposed on pages 121 and 121a were identified by the NRC in Reference 1 as changes needed to be consistent with ,

NUREG-0737 Item II.K.3.13, RCIC Restart. The proposed changes )

are considered _to match the above example (vii) and, therefore, to involve no significant hazard considerations.

The proposed changes to page 70a and 79 identified by_the NRC in Reference 1 as changes needed to be consistent with NUREG-0737 Item II.K.3.22, RCIC Suction. The proposed changes are considered to match the above example (vii) and, therefore, to involve no significant hazard considerations.

The proposed changes to pages 54, 60, 77a, 77b and 86a can be classified as not likely to involve significant hazard considerations, since they are changes to make a license conform to regulations where the license change results in very minor changes to facility operations, and, therefore, comply with example (vii).

The proposed change to Figure 6.2-1 on page 260 is needed to be consistent with NUREG-0737 / Item I.A.1.3.2., " Shift Manning". The proposed change is considered to match commission example (vii).

IV. Implementation of the Changes Implementation of the changes, as proposed, will not impact the fire protection program at FitzPatrick, nor will the changes impact the environment.

V. Conclusion The incorporation of these changes:

a) will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; 9

b) will.not incrocco ths poccibility of en accidsnt or malfunction of a different type then any evaluated previously in the Safety Analycis R3 port; c) will not reduce the margin of safety as defined-in '/

the basis for any Technical Specifications d) .does not constitute an unreviewed safety question, and e) involves no Significant Hazard Considerations, as

~

defined in 10.CFR.50.92.

VI. References

.1. NRC Generic Letter No. 83-02, dated January 10, 1983.

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

3. NUREG-07'37, " Clarification of TMI Action Plan Requirements," November, 1980.
4. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR), Rev. 1, July 1983.

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