Text

Forwards Changes to Facility Tech Specs,Per 880504 NRC Meeting W/Licensee.Corrected Tech Spec Page,Originally W/ Typo,Also Included.Stated Addl Changes Will Be Provided Separately
	ML20150E163
Person / Time
Site:	Seabrook
Issue date:	07/08/1988
From:	George Thomas; PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	GL-88-06, GL-88-6, NYN-88092, NUDOCS 8807150020
	Download: ML20150E163 (56)
	v • d • e

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George S. Thomas Vice Preddent Nucleot Product 6on Pub 5c Service of New HompeNro New Hampshire Yankee Division NYN- 88092 July 8, 1988 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk References a) Facility Operating License NPF-56, Docket No. 50-443 b) USNRC Meeting Notice dated April 18, 1988, ' Forthcoming Heeting With Public Service Company of New Hampshire to Discuss Seabrook Technical Specifications," V. Nerses to R. Wessman

Subject:

Changes to Seabroole Station Technical Specifications Gentlemen: l On May 4, 1988, representatives of New Hampshire Yankee (NHY) met with the NRC Staff to discuss Technical Specifications for Seabrook Station.

Cspies of Technical Specification pages reflecting the changes discussed at tuis meeting are provided as described below.

Enclosure 1 provides copies of changes which were initially discussed at an April 9, 1987, meeting with the Staff and reviewed at the May 4, 1988, meeting. Enclosure 2 contains copies of mark-ups of current 8eabrook Station Technical Specifications reflecting changes which were discussed with the Staff at the May 4, 1988 meeting. Additionally, further review of the Seabrook Station Technical Specifications has revealed a typographical error in a cross reference. A corrected copy of this page is provided ss Enclosure 3.

l In addition to the above, changes related to 1) replacement of Veritrak/Tobar transmitters and 2) a combination of Generic Letter 88-06 and the present organization structure will be provided as separate submittals.

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8807150020 88070s PDR P ADOCK 05000443 PDC P.O. Box 300. Seabrook, NH 03874 . Telephone (603) 474 9574

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Units / Stotcs Nuclear Rsgulctory -Commission - July 8, 1988-Attention: Document Control Desk Page 2 If you have any questions regarding this matter, please contact Mr.-Richard Belanger at (603) 474-9574, extension 4048.

Very truly yours, 4

Ceor S. Thomas

'Mr. Victor Herses, Project Manager Project Directorate I-3 Division of Reactor Projects United States Nuclear Regulatory Commission Washington, DC 20555 Mr. William T. Russell Regional Administrator

-United States Nuclear Regulatory Commission

' Region I 425 Allendale Road-King of Prussia, PA 19406 Mr. Antone C. Cerne NRC Senior Resident Inspector Seabrook Station Seabrook, NH 03874 -

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ENCLOSURE 1 TO NYN- 88092 r

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INDEX 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS................................................. 2-1 l

2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REA'. TOR COOLANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 2.0 BASES 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 1EACTOR COOLANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-2 2.2 LIMITING SAFETY SYSTEN SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS. ............. B 2-3 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY,.... .......................................... .'/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL l Shutdown Margin - T,yt, Greater Than 200*F............. . 3/4 1-1 Shutdown Margir. - T,yg Less lhan or Equal to 205"F.,..... 3/4 1-3 Moderator Temperature Coefficient............ s .......... 3/4 1-4 Minimum Temperature for Criticality.................. ... 3/4 1-6

, l SEABR0OK - UNIT 1 11 1

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

SECTION PAGE !

TABLE 3.3-2 (This table number is not used)

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-14 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYJTEM INSTRUMENTATION.......................................... 3/4 3-16 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS........................... 3/4 3-24 TABLE 3.3-5 (Tbis table number is not used)

TABLE 4.3-2 ENGINrERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMEN"ATION SURVEILLANCE REQUIREMENTS................ 3/4 3-31 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-26 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION l

FO R P LANT O P E RAT IO NS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-37 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-39 Movable Incore Detectors................................. 3/4 3-40 Seismic Instrumentation.................................. 3/4 3-41 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-42 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-43 Meteorological Instrumentation........................... 3/4 3-44 TABLE 5.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-45 Remote Shutdown System................................... 3/4 3-46 TABLE 3.3-9 REMOTE SHUTOOWN SYSTEM................................ 3/4 3-47 Accident Monitoring Instrumentation...................... 3/4 3-49

! TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-50 l TABLE 3.3-11 (This table number is not used)...................... 3/4 3-53 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-55 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION 3/4 3-56 SEABROOK - UNIT 1 iv i

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g INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-11 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM................... 3/4 7-12 3/4.7.4 SERVICE WATER SYSTEM..................................... 3/4 7-13 3/4.7.5 ULTIMATE HEAT SINK....................................... 3/4 7-14 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM..................... 3/4 7-16 3/4.7.7 SNUBBERS................................................. 3/4 7-18 3/4.7.8 SEALED SOURCE CONTAMINATION.............................. 3/4 7-19 3/4.7.9 (This specification number is not used).................. 3/4 7-21 3/4.7.10 AREA TEMPERATURE MONITORING.............................. 3/4 7-22 iABLE 3.7-3 AREA TEMPERATURE M0NITORING........................... 3/4 7-23 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating................................................ 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................ 3/4 8-10 Shutdown.................................... ............ 3/4 8-11 3/4.8.2 0.C. SOURCES .

Operating................................................ 3/4 8-12 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-14 Shutdown................................................. 3/4 8-15 3/4.8.3 ONSITE POWER DISTP'dVTION

! 0perating................................................ 3/4 8-16 Shutaown.............................................. .. 3/4 8-18 Trip Ci rcui t for Inverter I-2A. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-19 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. Circuits Inside Primary Containment................. 3/4 8-20 Containment Penetration Conductor Overcurrent Protective Devices and Protective Devices for Class 1E Power Sources Connected to Non-Class i 1E Circuits............................................ 3/4 R-21 .

l Motor-Operated Valves Thermal Overload Protection. . . . . . . . 3/4 6-24 l

l 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................., 3/4 9-1 3/4.9.2 INSTRUMENTATION.......................................... 3/4 9-2 l 3/4.9.3 DECAY TIME............................................... 3/4 9-3 SEABROOK - UNIT 1 viii

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 C O MM U N I C A T I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-5 3/4.9.6 REFUELING MACHINE........................................ 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS.................. 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1......................................... 3/4 9-8 Low Water Level.......................................... 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM. ......... 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL............................. 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE P0OL .............................. 3/4 9-12 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM...... 3/4 9-13 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4.10.3 PHYSICS TESTS............................................ 3/4 10-3 3/4.10.4 RE ACTO R COO LANT L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 Dose..................................................... 3/4 11-2 Liquid Radwaste Treatment System......................... 3/4 11-3 l Liquid Holdup Tanks...................................... 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS l

Dose Rate................................................ 3/4 11-5 Dose - Noble Gases....................................... 3/4 11-6 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form............................. 3/4 11-7 Gaseous Radwaste Treatment System........................ 3/4 11-8 Explosive Gas Mixture - System........................... 3/4 11-9 3/4.11.3 SOLID RADI0 ACTIVE WASTES................. ........... 3/4 11-10 3/4.11.4 TOTAL 00SE............................... .............. 3/4 11-12 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l

3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 SEABROOK - UNIT 1 ix l

.o TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 SENSOR E TOTAL ERROR e

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE E 11. Pressurizer Water Level - High 8.0 2.18 1.82 a 192% of instrument $93.8% of instrument span span

- 12. Reactor Coolant Flow - Low 2.5 1.87 0.6 >90% of loop >89.4% of loop Besign flow

3esign flow *

13. Steam Generator Water 17.0 15.28 -1.76 >21.6% of narrow >20.5% of narrow tevel Low - Low range instrument range instrument l span span

14. Undervoltage - Reactor 15.0 1.39 0 >10,200 volts >9,822 volts Coolant Pumps

15. Underfrequency - Reactor 2.9 0 0 >55.5 Hz >55.3 Hz y Coolant Pumps u.

1 16. Turbine Trip i

! a. Low Fluid Oil Pressure N. A. N.A. N.A. >500 psig >450 psig

b. Turbine Stop Valve N.A. N.A. N.A. >1% open >1% open Closure

17. Safety Injection Input N.A. N.A. N.A. N.A. N.A.

from ESF

Loop design flow = 95,700 gpm

'o TABLE 4.3-1 (Continued)

$s REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8

SE TRIP

. ANALOG ACTUATING MODES FOR c CHANNEL DEVICE WHICH

'E CHANNEL CHANNEL OPERATIONfL OPERATIONAL ACTUATION SURVEILLANCE CHECK CALIBRATION TEST TEST LOGIC TESI IS REQUIRED

-] FUNCTIONAL UNIT Reactor Trip System Interlocks (Continued)

e. Power Range Neutron Flux, P-10 N.A. R(4) R N.A. N.A. 1, 2

f. Turbine Impulse Chamber Pressure, P-13 N.A. R R N.A. N.A. 1 tt 19. Reactor Trip Breaker N.A. N.A. N.A. N(7, 11) N.A. 1, 2, 3* ,

4*, S*

T L: 20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. M(7) 1, 2, 3*,

Logic 4*, S*

21. Reactor Trip Bypass Breaker N.A. N.A. N.A. M(7,14), N.A. 1, 2, 3*,

R(IS) 4*, S*

O TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

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' SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE w

- 1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation, and Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency Fan / Filter Actuation, and Latching Relay).

{ a. Manual Initiatic.. N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

c. Containment Pressure--Hi-1 4.2 0.71 1.67 $ 4.3 psig 5 5.3 psig

d. Pressurizer Pressure--Low 13.1 10.71 1.69 1 1875 psig 1 1865 psig l

e. Steam Line Pressure--Low 13.1 10.71 1.63 > 585 psig > 568 psig*

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Containment Pressure--Hi-3 3.0 0.71 1.67 $ 18.0 psig 5 18.7 psig

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

o O

SENSOR TOTAL ERROR

[ FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE z

El 7. Emergency Feedwater s

a. Manual Initiation (1) Motor driven pump N.A. N.A. N.A. N.A. N.A.

(2) Turbine driven pump N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Steam Generator Water 17.0 15.28 1.76 us Level--Low-low 2 21.6% of 1 20.5% of narrow l narrow range range instrument 2: Start Motor-Driven Pump instrument span.

us and Start Turbine-Driven span.

Pump O

d. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Start Motor-Driven Pump Allowable Values.

and Turbine-Driven Pump

e. Loss-of-Off site Power See Item 9. for Loss-of-Offsite Power Setpoints and Allowable Values.

Start Motor-Driven Pump and Turbine-Driven Pump .

8. Automatic Switchover to Containment Sump

a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

b. RWST Level--Low-Low 2.75 1.0 1.8 ~>122,525 gals. >121,609 gals.

Ccincident With Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

INSTRUMENTATION MONITORING INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALI-BRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4.

4.3.3.3.2 Each of the above required seismic monitoring instruments actuated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months except that Triaxial Peak Accelerographs 1-SM-XR-6702 and 1-SM-XR-6703 shall be restored within 7 days and a CHANNEL CALIBRATION performed within 30 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.8.2 within 14 days de-scribing the magnitude, frequency spectrum, and resultant effect upon facility features important to safety.

SEABROOK - UNIT 1 3/4 3-41

y y 1 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS-INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

'1. Triaxial Time-History Accelerographs*

a. 1-SM-XT-6700 Free Fi21d Control i Ig. 1**

Room East Air Intake, elevation 11' 6"

b. 1-SM-XT-6701 Containment Foundation, i lg 1**

elevation -26' 0"

c. 1-SM-XT-6710 Containment Operating 3g 1**

Floor, elevation 25' 0"

2. Triaxial Peak Accelerographs

a. 1-SM-XR-6702 Reactor Vessel Support, 0-20 Hz 1 Containment Building, elevation -13' 4"

b. 1-SM-XR-6703 Reactor Coolant System 0-20 Hz 1 Piping, Containment Building, elevation -7' 8"

c. 1-SM-XR-6704 PCCW Piping, Primary 0-20 Hz 1 Auxiliary Building, elevation 47' 0"

3. Triaxial Seismic Switch 1-SM-XS-6709 Containment Foundation #, 0.025g to 0.25g 1**

elevation -27' 0"

4. Triaxial Response-Spectrum Recorders

a. 1-SM-XR-6705 Containment Foundation, 1-30 Hz 1**

elevation -26' 0"

b. 1-SM-XR-6706 SG 118 Support, 1-30 Hz 1 Containment Building, elevation -20' 0"

c. 1-SM-XR-6707 Primary Auxiliary 1-30 Hz 1 Building, elevation 25' 0"

d. 1-SM-XR-6708 Service Water Pump House, 1-30 Hz 1 elevation 4' 0"

Trigger mechanism in accelerograph unit activates recorders in control room when it senses a ground motion of 0.01g.

- With reactor control room indication

Switch setpoint is 0.,13g for horizontal and vertical axis.

SEABROOK - UNIT 1 3/4 3-42

TABLE 4.3-4

)

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG i CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST

1. Triaxial Time-History Accelerographs*

a. 1-SM-ki-6700 Free Field Control M R SA Room East Air Intake, elevation 11' 6"

b. 1-SM-XT-6701 Containment Foundation, M R N.A.

elevation -26' 0"

c. 1-SM-XT-6710 Containment Operating M R N. A.

Floor, elevation 25' 0"

2. Triaxial Peak Accelerographs

a. 1-SM-XR-6702 Reactor Vessel Support, N.A. R N.A.

Containment Building, elevation -13' 4"

b. 1-SM-XR-6703 Reactor Coolant System N.A. R N.A.

Piping, Containment Building, elevation -7' 8"

c. 1-SM-XR-6704 PCCW Piping, Primary N.A. R N.A.

Auxiliary Building, elevation 47' 0"

3. Triaxial Seismic Switch 1-SM-XS-6709 Containment Foundation,** M R N.A.

elevation -26' 0"

4. Triaxial Response-Spectrum Recorders

a. 1-SM-XR-6705 Containment Foundation,** M# R N.A.

elevation -26' 0"

b. 1-SM-XR-6706 SG 11B Support, Con- N.A. R N.A.

tainment Building, elevation -20' 0"

c. 1-SH-XR-6707 Primary Auxiliary N.A. R N.A.

Building, elevation 25' 0"

d. 1-SM-XR-6708 Service Water Pump House, elevation 4' 0" N.A. R N.A.

'Each accelerograph has a triaxial trigger to activate the recorder.

- With reactor control room indications.

CHANNEL CHECK to consist of turning the test / reset switch and verify all lamps illuminate on 1-SM-XR-6705.

SEABROOK - UNIT 1 3/4 3-43

.c. ,

TABLE 4.3-6 (Continued)

TABLE NOTATIONS

At all times.

During RADI0 ACTIVE WASTE GAS SYSTEM operation.

When the gland seal exhauster is in operation.

- - - The CHANNEL OPERATIONAL TEST for the flow rate monitor shall consist of a verification that the Radiation Data Management System (ROMS) indicated flow is consistent with the operational status of the plant.

Noble Gas Monitor for this release point is based on the main condenser air evacuation monitor.

(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm / Trip Setpoint.

(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control [

room alarm annunciation occurs if the instrument indicates measured levels above the Alarm Setpoint.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of l the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. Theso standards shall per-mit calibrating the system over its intended range of energy and measure-ment range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

a. One volume percent oxygen, balance nitrogen, and

b. Four volume percent oxygen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall be performed using sources of various activities covering the measurement range of the monitor to verify that the l response is linear. Sources shall be used to verify the monitor response only for the intended energy range.

i l

SEABROOK - UNIT 1 3/4 3-66 l-

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REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AhD COOLANT CIRCULATION HOT STANOBY SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 The required steam generators shall be determined OPERABLE by verify-ing secondary side water level to be greater than or equal to 21.6% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l 4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 l

l l

l SEABROOK - UNIT 1 3/4 4-3 l

REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION HOT SHUTDOWN SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary-side water level to be greater than or equal to 21.6% at l 1 east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SEABROOK - UNIT 1 3/4 4-5

REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION COLD SHUTOOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION

3. 4 1. ?.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:

a. One additional RHR loop shall be OPERABLE **, or

b. The secondary-side water level of at least two steam generators shall be greater than 21.6%.

l APPLICABILITY: MODE 5 with reactor coolant loops filled ***.

ACTION:

a. With one of the RHR loops inoperable and with less than the required steam generator water level, inmediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.1.2 At least ona RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration and (2) core outlet temperature is maintained at least 10'F below saturation temperature.

- 0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing l provided the other RHR loop is OPERABLE and in operation.

- - A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold-leg temperatures, i SEABROOK - UNIT 1 3/4 4-6

, ;,- 0

~

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpa UNIDENTIFIED LEAKAGE,

c. 1 gpm total reactor-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,

c. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 psig i 20 psig, and

f. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.* l APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDB) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Test ptessures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted for the actual test pressure up to I

2235 psig assuming the leakage to be directly proportional to pressure dif-ferential to the one-half power.

SEABROOK - UNIT 1 3/4 4-21 l

TABLE 4.4-3 h REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM E

8 TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY

[ AND ANALYSIS REQUIRED 25

-d

1. Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 1,2,3,4 w

Determination

2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days. 1 LENT I-131 Concentration

3. Radiochemical for 5 Determination

1 per 6 months ** 1

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the 1#, 2#, 3#, 4#, 5#

Including I-131, I-133, and I-135 specific activity exceeds 1 pCi/ gram DOSE EQUIVALENT I-131 or 100/E pCi/ gram of gross l radioactivity, and w

]; b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 1, 2, 3 2 following a THERMAL POWER change

', exceeding 15% of the RATED THERMAL i

POWER within a 1-hour period.

A radiochemical analysis for 5 shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in the reactor coolant. The specific activities for these individual radionuclides shall be i used in the determination of E for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.

- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 h9drs or longer.

Until the specific activity of the Reactor Coolant System is restored within its limits.

EMERGENCY CORE COOLING SYSTEMS ACCUMULATORS HOT STANDBY, STARTUP, AND POWER OPERATION SURVEILLANCE REQUIREMENTS 4.5.1.1.1 (Continued)

c. At least once per 31 days when the RCS pressure is above 1000 psig by l verifying that power to the isolation valve operator is disconnected.

d. At least once per 18 months by verifying that each accumulator isola-tion valve opens automatically under each of the following conditions:

1) When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint, and

2) Upon receipt of a Safety Injection test signal.

P 4.5.1.1.2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE:

a. At least once per 31 days by the performance of an ANALOG CHANNEL OPERATIONAL TEST, and

b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.

L SEABROOK - UNIT 1 3/4 5-2

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,y GREATER THAN OR EQUAL TO 350'F SURVEILLANCE REQUIREMENTS 4.5.2 (Continued)

d. At least once per 18 months by:

1) Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System to ensure that: l a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 365 psig, the interlocks prevent the valves from being opened, and b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 660 psig, the interlocks will cause the valves to automatically close.

2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. At least once per 18 months, during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on (Safety Injection actuation and Automatic Switchover to Containment Sump) test signals, and

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.

f. By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1) Centrifugal charging pump, 1 2480 psid;

2) Safety Injection pump, 1 1445 psid; and l

l 3) f!HR pump, 1 176 psid.

i SEABROOK - UNIT 1 3/4 5-6 l

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fled in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI /

N45.4-1972:

a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 i 10 month intervals during shutdown at a pressure not less than P,, 49.6 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection;

b. If any periodic Type A test fails to meet 0.75 L,, the test schedule l for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L,,

a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumed;

c. The accuracy of each Type A test shall be verified by a supplemental test which:

1) Confirms the accuracy of the test by verifying that the supple-mental test result, L c, is in accordance with the following equation:

lLc ' Cl am + l o)l 1 0.25 L, where L is the measured Type A test leakage and L, is the am superimposed leak;

2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and

3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment during the supplemental test is between 0.75 La and 1.25 L,.

SEABROOK - UNIT 1 3/4 6-3

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:

a. Each 36-inch containment shutdown purge supply and exhaust isolation valve shall be closed and locked closed, and

b. The 8-inch containment purge supply and exhaust isolation valve (s) shall be sealed closed except when open for purge system operation for pressure control; for ALARA, respirable, and air quality consider-ations to facilitate personnel entry; and for surveillance tests that require the valve (s) to be open.

APPLICABILITY: MODES 1*, 2*, 3, and 4.

ACTION:

a. With a 36-inch containment purge supply or exhaust isolation valve open or not locked closed, close and lock closed that valve l or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one or more of the 8-inch containment purge supply or exhaust isolation valves open for reasons other than given in Specifica-tion 3.6.1.7.b above, close the open 8-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , othe mise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one or more containment purge supply or exhaust isolation valves having a measured leakage rate in excess of the limits of Specifications 4.6.1.7.2 or 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status or isolate the affected penetration (s) so that the me n ured leakage rate does not exceed the limits of Specifications 4., : .7.2 or 4.6.1.7.3 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and close tne purge supply if th affected penetration is the exhaust penetration, otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l l

The 8-inch containment purge supply and exhaust isolation valves may not be opened while in MODE 1 or MODE 2 until installations of the narrow-range con-tainment pressure instrument channels and alarms are completed.

l l SEABROOK - UNIT 1 3/4 6-12

CONTAINMENT SYSTEMS I 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Spray System capable of taking suction from the RWST* a ferring suction to the containment sump. [

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Spray System inoperable, restore the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is n b.

By verifying, that on recirculation flow each pump develops a differ 9ntial pressure of greater than or, equal to 262 psi when tested pursuant to Specification 4.0.5; c.

At least once per 18 months during shutdown, by:

1)

Verifying that each automatic valve in the flow path actuates to its correct signal, and position on a Containment Pressure-Hi-3 test 2)

Verifying that each spray pump starts automatically on c Containment Pressure-Hi-3 test signal.

d.

At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

"In MODE 4, when the Residual Heat Removal System is in operation, an OPERAPLE flow path is one that is capable of taking suction from the refueling water storage tank upon being manually realigned.

SEABROOK - UNIT 1 3/4 6-14

~

?

l CONTAINMENT SYSTEMS COMBUSTIBLE GAS CONTPOL ELECTRIC HYOR0 GEN RECOMBIN5RS LIMITING CONDITION FOR OPERATION ,

3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERA 3LE.

APPLICABILITY: MODES 1 and 2.

ACTION:

With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , iNCE REQUIREMENTS 4.0.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:

a. At least once per 6 months by verifying during a Hydrogen Recombiner System functional test that the minimum heater sheath temperature increases to greater than or equal to 850*F within 90 minutes.

l Upon reaching 850 F, increase the power setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 65 kW; and

b. At least once per 13 months by:

1) Perforning a CHANNEL CALIBRATION of all recombiner instr intation and control circuits,

2) Verifying throug'r a visual examination that there is no evidence of abnor mal conditions within the recombiner enclosure (i.e., loose wiriqg or structural connections, deposits of foreign materials, etc.), and

3) Verifying the integrity of cil heater electrical circuits by performing a resistance to ground test following the cbove l

required functional test. The resistance to ground for any j heater phase shall be greater thca or equal to 10,000 ohms.

SEA 8 ROOK - UNIT 1 3/4 6-19

3 CONTAINMENT SYSTEMS 3_L4J.5 CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent Containment Enclosure Emergency Air Cleanup Systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Enclosure Emergency Air Cleanup System inoperable, re-store the inoperable system to OP2RABLE statLs within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.5.1 Each Containment Enclosure Emergency Air Cleanup System shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes;

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charr.oal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the system by:

1) Verifying that the cleanup system satisfies the in place pene-tration leakage testing acceptance criteria of iass than 0.05%

and uses the test procedure guidance in Regulatory Posi-tions C.S.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978*, and the system flow rate it 2100 cfm l t 10%;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.t of Regulatory Guide 1.52, Revision 2, March 1978*, meets the laboratory testing criteria

ANSI NH0-1980 shall be used in place of ANSI H510-1975 referenced in Regulatory Guide 1.52, Rev. 2, Maech 1978.

SEACRC0K - UNIT 1 3/4 6-21

o CONTAINMENT SYSTEMS CONTAINMENT ENCLOSURE BUILOING CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM SURVEILLANCE REQUIREMENTS 4.6.5.lb.2 (Continued) of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978*, by showing a methyl iodide penetration of I less than 2.14% when tested at a temperature of 30 C and at a relative humidity of 95% in accordance with ASTH-03803; and

3) Verifying a systen, flow rate of 2100 cfm i 10% during system operation when tested in accordance with ANSI N510-1980,

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978*, l meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978*, by showing a l methyl iodide penetration of less than 2.14% when tested at a tem-perature of 30 C and at a relative humidity of 95% in accordance with ASTM-03803.

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 incnes Water Gauge while operating the system at a flow rate of 2100 cfm i 10%,

2) Verifying that the system starts on a Safety Injection test signal,

3) Verifying that the filter cross connect valves can be manually opened, and l

4) Verifying that each system produces a negative pressure of greater than or equal to 0.25 inch Water Gauge in the annulus within 4 minutes after a start signal. I

e. After each complete or partial replacement of a high efficiency particulata air (HEPA) filter bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05% in accordance with ANSI NS10-1980 for a dioctyl phthalate (D0P) test aerosol while operating the system at a flow rate of 2100 cfm i 10%; and

ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.

SEABROOK - UNIT 1 3/4 6-22

PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEE 0 WATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.la. (Continued)

2) Verifying that the steam turbine-driven pump develops o discharge pressure of greater than or equal to 1460 psig at a flow of greater than or equal to 270 gpm when the secondary steam supply pressure is greater than 500 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3;

3) Verifying that the startup feedwater pump develops a discharge i pressure of greater than or equal to 1375 psig at a flow of greater than or equal to 425 gpm; l

4) Verifying that each non-automatic valve in the flow pa;.: that is not locked, sedled, or otherwise secured in position is in its correct position;

5) Verifying that each automatic valve in the flow path is in the fully open position wherever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER; and

6) Verifying that valves FW-156 and FW-163 are OPERABLE for l alignment of the startup feedwater pump to the emergency feedwater header.

b. At least once per 18 months during shutdown by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater System Actuation test signal;

2) Verifying that each emergency feedwater pump starts

as oesigned I automatical'y upon receipt of an Emergency Feedwater Actuation System test signal;

3) Verifying that with all manual actions, including power source and valve alignment, she startup feedwater pump sttrts within the required elapsed time; and

! 4) Verifying that -?.ch emergency feedwater cont,ol valve closc on receipt of a high flow test signal, l

i

For the steam turbine-driven pum,), when the secondary steam supply pressure l

is greater than 500 psig.

SEABROOK - UNIT 1 3/4 7-4

t a

o PLANT SYSTEMS 3/4.7.6 CANTROL ROOM AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two Control Room Area Ventilation Systems shall be OPERABLE.

APPLICABILITY: All MODES.

ACTION:

MODES 1, 2, 3, and 4:

With one C;ntrol Room Area Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6:

a. With one Control Room Area Ventilation System inoperable, restore the inopurable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Area Ventilation System in the recirculation mode.

b. With both Control Room Area Ventilation Systems inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.7.6 Each Control Room Area Ventilation System shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the Control Room Area Ventilation System is maintaining the temperature of equipment and instrumentation in the control room area below its limiting equipment qualification temperature.

b. At least once per 18 onths or after any Lignificant modification to l the Control Room Area Ventilation Systems by verifying a system flow rate of 25,700 cfm i 10% through the air conditioner unit (3A and 3B) and a flow rate of 1200 cfm i 10% makeup from each intake to the emergency filtratioa unit with a discharge of 2000 cfm i 10% from i the filtration unit.

l l

l SEABROOK - UNIT 1 3/4 7-16 l

l

ELECTRICAL POWER SYSTEMS A.C SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.9.1.1.1 Each of the above required independent circuits between the offsite tra.smission network and the Onsite Class 1E Distribution System shall be:

a. Determined OPERABLE at least once per 7 days by verifying correct breaker t.lignments, indicated power availability, and

b. Demonst,ated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the normal circuit to the alternato circuit.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:

1) Verifying the fuel level in the day fuel tank;

2) Verifying the fuel level in the fuel storage tank;

3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank;

4) Verifying the lubricating oil inventory in storage;

5) Verifying the diesel starts from ambient condition and acceler-ates to at least 514 rpm in less than or equal to 10 seconds.*

The generator voltage and frequency shall be 4160 420 volts and 60 1.2 Hz within 10 seconds

after the start signal. The diesel generator shall be started for this test by using one of the following signals:

a) Hanual, or b) Simulated loss-of-offsite power by itself, or i

, *All diesel generator starts for the purpose of this surveillance test may be l preceded by an engine prelube period. Further, all surveillance tests and all other engine starts for the purpose of this surveillance testing, with the l exception of once per 184 days, may also be preceded by warmup procedures (e.g.,

gradual acceleration and/or gradual loading greater than 60 seconds) as

-recommended by the manufacturer so that the mechanical stress and wear on the l diesel engine is minimized.

1 1 ,

l

! SEABROOK - UNIT 1 3/4 8-3 l

ELECTRICAL POWER SYSTEMS A.C. SOURCES ;

OPERATIN_G SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) c) Simulated loss-of-offsite power in conjunction with an SI Actuation test tignal, or d) An SI Actuation test signal by itself.

6) Verifying the generator is synchronized, loaded to greater than or equal to 6083 kW in less than or equal to 120 seconds *,

and operates with a load greater than or equal to 6083 kW for at least 60 minutes; and

7) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses,

b. At least once per 31 days and after each operation of the diesel where the period of operr' ion was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and remov i accumulated water from the day fuel talk;

c. At least once per 31 oas by checking for and removing accumula:ed water from the fuel oil storage tanks;

d. By sampling new fuel oil in accordance with ASTM-04057-81 prio to addition to storage tanks and:

1) By verifying in accordance with the tests specified in ASTM-0975-81 prior to addition to the storage tanks that the sample has:

a) An API Gravity of within 0.3 degree at 60 F, or a specific gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60 F ci grater than or equal to 0.81 but less than or equal to 0.89, e an API gravity of greater than or equal to 28 degrees but less than or equal to 42 degrees;

All diesel generator starts for the purpose of this surveillarce test may be preceded by an engine prelube period. Further, all surveillance tests and all other engine starts for the purpose of this surveillanca testing, with the exception of once per 184 days, may also be preceded by warmup procedures (e.g.,

gradual acceleration and/or gradual loading greater than 60 seconds) as recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.

SEABROOK - UNIT 1 3/4 8-4

e ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) b) A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification; c) A flash point greater than or equal to 125 F; and d) A clear and bright appearance with proper color when tested in accordance with ASTM-D4176-82.

2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-0975-81 are met when tested in accordance with ASTM-D975-81 except that the analysis for sulfur may be performed in accordance with A3TM-01552-79 or ASTM-02622-82.

e. At least once every 31 days: l

1) By obtaining a sample of fuel oil in accordance with ASTM-02276-78,

, and verifying that total particulate contamination is less than l 10 mg/ liter when checked in accordance with ASTM-D2276-78, Method A, and

2) By visually inspecting the lagging in the area of the flanged joints on the silencer outlet of the diesel exhaust system for leakage (also after an extended operatior,of greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

f. At least once per 18 months, during shutdown, by:

l 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service; 1

l

2) Verifying the generator capability to reject a load of greater than or equal to 671 kW while maintaining voltage at 4160 420 volts and frequency at 60 4.0 Hz; i 3) Verifying the generator capability to reject a load of 6083 kW without tripping. The generator voltage shall not exceed 4784 volts during and following the load rejection;

4) Simulating a loss-of-offsite power by itself, and:

i SEABROOK - UNIT 1 3/4 8-5

I* 0 ELECTRICAL POWER SYSTEMS ELECTRICAL EQJIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES AND PROTECTIVE DEVICES FOR CLASS lE POWER SOURCES CONNECTED TO NON-CLASS lE CIRCUITS l

LIMITING CONDITION FOR OPERATION 3.8.4.2 Each containment penetration conductor overcurrent protective device and each protective device for Class 1E power sources connected to non-Class 1E l circuits shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, 4, 5,* and 6.*

ACTION:

a. With one or more of the containment penetration conductor overcurrent pro-tective device (s) inoperable:

1) Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated circuit breaker or racking out or removing the inoperable protective device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the circuit breaker to be tripped or the inoperable protective device to be racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their circuit breakers tripped, or their inoperable protective devices racked out, or removed; or

2) Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. Witn one or more of the Class 1E power source protective device (s) inoper- l able, restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the circuit breaker or racking out or removing the inoperable protective device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected com- l ponent inoperable, and verify the circuit breaker to be tripped or the inoperable protective device to be racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their circuit breakers tripped, or their inoperable protective devices racked out, or removed.

SURVEILLANCE REQUIREMENTS 4.8.4.2 Each containment penetration conductor overcurrent and Class 1E power l source protective device shall be cemonstrated OPERABLE:

a. At least once per 18 months:

1) By verifying that the medium voltage 13.8-kV and 4.16-kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least one of the circuit breakers, and performing the following:

0nly for Class 1E power source protective devices. l l SEABROOK - UNIT 1 3/4 8-21

o ELECTRICALPOWERSYST5MS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES AND PROTECTIVE DEVICES FOR CLASS 1E POWTT T6DTCES CONNECTED TO NON-CLASS 1E CIRCUITS l SURVEILLANCE REQUIREMENTS 4.8.4.2.a.1) (Continued) a) A CHANNEL CAL.IBRATION of the associated protective relays (because of the large currents involved, it is impractical to inject primary side signals to current transformers; therefore, the channel calibration will be performed by injecting a signal cn the secondary side of those trans-formers at their test plug),

b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and c) For each circuit breaker found inoperable during these func-tional tests, one additional circuit breaker of the inoper-able type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

2) By selecting and functionally tes?.ing a representative sample of at least 10% of each type of lower voltage circuit breakers and overload devices. Circuit breakers and overload devices selected for functional testing shall be selected on a rotating basis.

Testing of air circuit breakers shall consist of injecting a cur-rent with a value equal to 300% of the pickup of the long-time delay trip element and 150% of the pickup of the short-time delay trip element. The instantaneous element shall be tested by inject-ing a current equal to 120% of the pickup value of the element.

Testing of thermal magnetic molded-case circuit breakers shall consist of injecting a current with a value equal to 300% of the circuit breaker trip rating and -25% to +40% of the circuit breaker instantaneous trip range or setpoint.

Testing of combination starters (a magnetic only molded-case circuit breaker in series with a motor starter and integral overload device) shall consist of injecting a current with a value equal to -25% to +40% of the circuit breaker instantaneous trip setpoint, and 200% and 300% of the ther. mal overload device trip rating to the respective devices.

Circuit breakers and/or overload devices found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker and or overload de-vices found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers and or overload devices of the inoperable type shall also be func-tionally tested until no more failures are found or all circuit breakers and or overload devices of that type have been func-tionally tested.

SEABROOK - UNIT 1 3/4 8-22

. j ELECTRICAL POWER SYSTEMS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT-PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES AND PROTECTIVE DEVICES FOR CLASS 1E POWER SOURCES CONNECTED TO NON-CLASS lE CIRCUITS SURVEILLANCE REQUIREMENTS 4.8.4.2.a (Continued)

3) Corrective actions for any generic degradation of overcurrent protective devices, such as setpoint drift, manufacturing deficiencies, material defects, etc., shall be applicable to all (Class 1E and non-Class 1E) protective devices of identical design.

b. At least once per 60 months by subjecting each circuit breaker to an inspection ard preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

l I

SEABROOK - UNIT 1 3/4 8-23 1

f RADIOACTIVE EFFLUENTS GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE - SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the GASEOUS RADWASTE TREATMENT SYSTEM shall be limited to less than or equal to 2% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in the GASE0US RADWASTE TREATMENT SYSTEM greater than 2% by volume, reduce the oxygen concentration to the above limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless the hydrogen concentration is verified to be less than 4% by volume,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen or oxygen in the GASE0US RADWASTE TREAT-MENT SYSTEM shall be determined to be within the above limit by continuously monitoring the waste gases in the GASE0US RA0 WASTE TREATMENT SYSTEM with the hydrogen or oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10. l l

l t

SEABROOK - UNIT 1 3/4 11-9

(

PSNH PRESIDENT

& CHIEF EXECUTIVE OFFICER I

NH YANKEE PRESIDENT & CEO SENIOR VICE PRESIDENT i I VICE PRESIDENT VICE PRESIDENT NUCLEAR ENGINEERING AND PRODUCTION QUALITY PROGRAMS l

STATION MANAGER NUCLEAR QUALITY MANAGER I I DIRECTOR OF O! RECTOR OF C RPO E MGMT. CONTROL Ep FIGURE 6.2-1 j OFFSITE ORGANIZATION SEABROOK - UNIT 1 6-2 l

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! CONTROL ROOM E E OPE RATOR HEAETH ggggggggy 1 me i CRO 8"*- PHYSSCS DEPARTMENT MAINTE NANCE INST. de CONT DEPARTME NT DEPART MENT DEPARTMENT I

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! STATION ORGANIZATION

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ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine station operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of station design and operating experience information, including units of similar design, which may indicate areas for improving station safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving station safety to the Executive Assistant to the Senior Vice President.

COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of station activities to provide independent verification

that these activities are performed ccrrectly and that human errors are reduced as much as practical.

RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to the Executive Assistant to the Senior Vice President.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Centrol Room Commander in the areas of thermal hydraulics, reactor engi-neering, and plant analysis with regard to the safe operation of the station.

6.3 TRAINING 6.3.1 A retraining and replacement licensed training program for the station staff shall be maintained under the direction of the Training Center Manager and shall meet or exceed the requirements and recommendations of Section 5.5 I of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental l

requirements specified in Sections A and C of Enclosure 1 of the NRC letter l dated March 28, 1980 to all licensees, and shall include familiarization with I

relevant indurtry operational experience.

I t

Not responsible for sign-off function.

SEABROOK - UNIT 1 6-5 l

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.4d. (Continued)

1) Training of personnel, and

2) Procedures for monitoring,

e. Post-Accident Sampling A program that will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samplos under accident conditions. The program shall include the following:

1) Training of personnel,

2) Procedures for sampling and analysis, and

3) Provisions for maintenance of sampling and analysis equipment.

f. Accident Monitoring Instrumentation

A program which will ensure the capability to monitor plant variables and systems operating status during and following an accident. This program shall include tnose instruments provided to indicate system operating status and furnish information regarding the release of radioactive materials (Category 2 and 3 instrumentation as defined in Regulatory Guide 1.97, Revision 2) and provide the following:

1) Preventive maintenance and periodic surveillance of instrumentation,

2) Pre planned operating procedures and backup instrumentation to be used if one er more monitoring instruments become inoperable, and

3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

6.8 REPORTING REQUIREMENTS ROUTINE REPORTS 6.8.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT 6.8.1.1 A summary report of station startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the station.

Implementation of this specification shall take effect when plant goes above 5% power for the first time.

SEABROOK - UNIT 1 6-14

ADMINISTRATIVE CONTROLS The Startup Peport shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall l also be described. Any additional specific details required in license condi- l tions based on other commitments shall be included in this report. I Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, ,

and resumption or commencement of commercial operation), supplementary reports I shall be submitted at least every 3 months until all three events have been l completed.

ANNUAL REPORTS

6.8.1.2 Annual Reports covering the activities of the station as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

Reports required on an annual basis shall include:

a. A tabulation on an annual basis of t N number of station, utility, and other personnel (including cont Lctors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions ** (e.g. , reactor operations and surveillance, inservice inspection, routine maintenance, special

.31ntenance [ describe maintenance], waste processing, and refueling).

The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), o-film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least C0% of the total whole-body dose received from external sources snould be assigned to specific major work functions;

b. The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed p"ior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; l (4) Graph of the I-131 concentration (pCi/gm) and one other radio-iodine isotope concentration (pCi/gm) as a function of time for the

A single submittal may be made for a multiple unit station. The submittal should combine those sections that ire common to all units at the station.

- This tabulation supplements the reqeirements of $20.407 of 10 CFR Part 20.

SEABROOK - UNIT 1 6-15

e l

. l ADMINISTRATIVE CONTROLS 6.8.1.2 (Continued) duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

c. Documentation of all challenges to the pressurizer powar-operated relief valves (PORVs) and safety valves.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.8.1.3 Routine Annual Radiological Environmental Operating Reports covering the operation of the station during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality and shall include copies of the preoperational Radiological Environmental Program of the unit for at least 2 years prior to criticality.

The Annual Radiological Environmental Operating Repor.ts shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls, as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include the results of the Land Use Census required by Specification 3.12.2.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Oose Calculation Manual, as well as summarized and tabulated results of these analyses and l measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some indi-vidual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing ret,ults.

The missing data shall be submitted as soon as possible in a supplementary repe r (..

lhe reports shall also include the following: a summary description of l

the Radiological Environmental Monitoring Program; at least two legible maps **

covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Specification 3.12.3; reason for not conducting the Radiological Environmental Monitoring Program as required by specification 3.12.1, and discussion of all deviations from the sampling schedule; discussion of environmental sample measurements that exceed the reporting levels but are not the result of plant effluents, pursuant to ACTION b. of Specification 3.12.1; and discussion of all analyses in which the LLO required was not achievable.

i l

l *A single submittal may be made for a multiple unit station.

l

- 0ne map shall cover locations near the SITE B0VHDARY; the more distant locations shall be covered by one or more additional n.aps.

SEABROOK - UNIT 1 6-16

ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT

6.8.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the station during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the station as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g. , cement).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year **. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured),

or in the form of joir.t frequency distributions of wind speed, wind direction, and atmospheric stability.*** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE B0UNDARY (Figure 5.1-3) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year

A single submittal may be made for a multiple unit station. The submittal

! should combine those sections that are common to all units at the station; l however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

- The dose calculations may be reported in a supplement submitted 30 days later.

! ***In lieu of submission with the Semiannual Radioactive Effluent Release

, Report, the licensee has the option of retaining this summary of required

! meteorological data on site in a file that shall be provided to the NRC l upon request.

SEABROOK - UNIT 1 6-17

o i

L ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 (Continued) to show conformance with 40 CFR Part 190, "Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the 00CM, pursuant to Specifications 6.12 and 6.13, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.14. It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.

The Semiannual Radioactive Effluent Release Reports shall also include the following: .an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORTS 6,8.1.5 Routine reports of operating statistics and shutdown experience shal e be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the repe ..

RADIAL PEAKING FACTOR LIMIT REPORT 6.8.1.6 The F xy limits for RATED THERMAL POWERx(FRTP) shall be provided to the NRC Regional Administrator with a copy to Director of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D. C. 20555, for l all core planes containing Bank "0" contol rods and all unrodded core planes and the plot of predicted (F pre;) vs Axial Core Heignt with the limit en-velope at least 60 days prior to each cycle initial criticality unless other-wise approved by the Commission by letter. In addition, in the event that the limit shculd change requiring a new substantial or an amended submittal to the Radial Peaking Factor Limit Report, it will be submitted 60 days prior to the date the limit would become effective unless otherwise approved by the Commis-sion by letter. Any information needed to support F will be by request from the NRC and need not be included in this report.

JABR00K - UNIT 1 6-18

- . _ . ~, .-

EN.'LOSURE 2 TO NYN- 88092 l

l l

i i'

l l

l

6

. 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 80 RATION CONTROL SHUTOOWN M RGIN - T,y GREATER THAN 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN for four-loop operation shall be greater than or equal to 3.5 ik/h in "00ES 1, 2, :nd 3 :nd-1,3% Ak/k,F"^^E i.

APPLICABILITY: MODES 1, 2*, 3, and 4.

ACTION:

With the SHUTOOWN MARGIN less than the limiting value, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUT 00WH MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.' The SHUT 00WN MARGIN shall be determined to be greater than or equal to the l'miting value:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTOOWN MARGIN shall be verified acceptable with an ir. creased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b. When in MODE 1.or MODE 2 with k,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c. When in MODE 2 with k,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum inser- '

tion limit of Specific:. tion 3.1.3.6; and i

See Special Test Exceptions Specification 3.10.1.

SEABROOK - UNIT 1 3/4 1-1

5 .

t s! '

' ' IABLE 3.307 S!!5MIC 9CNITCRING [NSTRUMENTAi!CN MINIMUM MEASUREFENT INSTRUMENTS INSTRU9ENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Time-History Accelerographs a.1-5M-XT-6700 Free field East Cont. : 1g 1" l Roots Air Intake ]

b. 1-SM-XT-6701 Containment Foundation : Ig 1" I

c. 1 SM-XT 6710 Cont. Opr. Floor : Ig 1=

1

2. Triaxial Peak Accelerograchs a.1-5M-XR-6702 ff(('.$ 'I$!: :::k t1. -2hYz[ " ~ 1

c. 1-SM-XR-6703 Safety in b; .cr .:j9ct ion

.. Ficin Elevation (-)24'0"g .

0-20 Hz. 1

c. 1-SM-XR 6704 PCCV Piping 0-20 Hz. 1

3. Triaxial Seismic Switches a.1-5M-XS-6700 Free Field N.A. 1"

b. 1-SM-XS-6701 Containment Foundation N.A. 1"

c. 1-SM-XS-6709 Containment Foundation 0.025g to 0.25g 1"

d. 1-SM-XS-6710 Cont. Opr. Floor H.A. 1" 4 Triaxial Response-Spectrum Recorders a.1-5M-XR-6705 Containment Foundation 1-30 Hz. 1" p S

b. ,,SM,XR,6706 .pn.t,Qn;sent Foundation nep,tgz.I-TK-9C, 3

. ...~r. . .

Elevation (-)26'0"

c. 1 SM-XR-6707 Prim. Aux. Bldg. 1-30 Hz. 1 c.1-5M-XR-6708 Service Vater Pump House 1- M Hz. 1 "Vith reactor control room indication SEASRC0K - UNIT 1 3/4 3-42 k- i

~

1

.L- -

/ 6 IABLE 4.3 6 SE!SMIC uCNITCRING IN5*RU9ENTA**CN SURVEILLANCE RECUIRE95NIS ANALOG CHANNEL CHANNEL CHANNEL OPERATICNAL INSTRL9ENTS AND SENSOR LOCATIONS CNECX CALIBRATION TEST

1. Triaxial Time-History Ac:elerographs a.1-5M-XT-6700 Free Field East Cont. M" R SA Room Air Intake

b. 1-SM-XT-6701 Containment Foundation M" R N. A.

c.1-5M-XT-6710 Cont, Opr. F'cor M" R N.A.

2. Triaxial Peak Acceleregraphs

4. Accumulator 1-SM-XR-6702 .1;;;. r 7::: ' Tank SI-TK-9C,N.A. Elevation 5c:::-t (-36'0" R N.A.

Sa ion ,

c. '.-5M-XR-6703 E::f :::-

e ty Inj,ec*Tip'i

...-. n,g' N.A. R N.A.

Elevation (-)2* O

c. 1-SM-XR-6704 PCCV Pioing N.A. R N. A.

3. Triaxial Seismic Swit:nes

a. 1-SM-X5-6700 Free Field M R SA

o. 1-SM-X5-6701 Containment Foundation"" M R N.A.

c. 1-SM-X5-6709 Containment Foundation"" M R N.A.

d. 1-SM-X5-6710 Cont. Opr. Floor "" M R N.A.

4 Triaxial Resoonse-Spectrum Recorders

, a. 1-SM-XR-6705 Containment Foundation"" M# R N.A.

b. 1-SM-XR-6706 mQoncainment

... ...... . Joundation nexk. .ko SI-TK-9C,

. x N.A.

. -)26'0" Elevation

c. 1-SM-XR-6707 Pris. Aux. (ldg.B N.A. R N.A.

d. 1-SM-XR-6708 Service Vater Pumo House N.A. R N.A.

i

'Exceot seismic trigger l '"Vitn reactor control room incications.

l # CHANNEL CHECX to consist of turning tne test / reset switen anc verify all l

lamps illuminate on 1-SM-XR-6705.

l l

SEABROOK - UNIT 1 3/4 3-43 i

I

q

~

a TABLE 3.3-9

v. REMOTE SHUTDOWN SYSTEM m

da TOTAL NO. MINIMUM

$ OF CHANNELS SE INSTRUMENT LOCATION CHANNELS OPERABLE c: 1. Intermediate Range Neutron Flux CP-108 A and B 2 1 35 2. Source Range Neutron Flux CP-108 A and B 2 1

-4

3. Reactor Coolant Temperature -

Wide Range for Loops 1 and 4

a. I CP-108 A nd B 2 2 c .r

b. T g CP-108E{h_andj[) 2 2

4. Pressurizer Pressure CP-108 A and B 2 2 S. Pressurizer Level CP-108 A and B 2 2

6. Steam Generator Pressure CP-108 A and 8 1/stm. gen. 1/sta. gen.

7. Steam Generator Water Level CP-108 A and B 1/stm. gen. 1/sta. gen.'

u, 8. Steam Generator-Emergency Feedwater 2: Flow Rate CP-108 A and B 1/sta. gen. 1/sta. gen.

u, 9. Boric Acid Tank Level CP-108 A and B 1/ tank 1/ tank

$1 TRANSFER SWITCHES / CONTROL CIRCUITS LOCATION

1. Emergency Fee Nater Pump Steam Supply Valves MS-V-393/127 CP-108 A

2. Emergency Feedwater Pump Steam Supply Valves MS-V-394/128 CP-108 8

3. Emergency Feedwater Pump Steam Supply Valves MS-V-395 CP-108 A and S

4. Emergency Feedwater Pump FW-P-37B BUS 6 SWGR S. Emergency feedwater Recirculation Valve FW-V-346 CP-108 A

6. Emergency Feedwater Recirculation Valve FW-V-347 CP-108 8

7. SG A EFW Control Valve FW-FV-4214 A CP-108 A

8. SG A EFW Control Valve FW-FV-4214 B CP-108 8

9. SG B EFW Control Valve FW-FV-4224 A CP-108 A

10. SG B EFW Control Valve FW-FV-4224 8 CP-108 B

11. SG C EFW Control Valve FW-FV-4234 A CP-108 A

12. SG C EFW Control Valve FW-FV-4234 B CP-108 8 13~. SG 0 EFW Control Valve FW-FV-4244 A CP-108 A

14. SG D EFW Control Valve FW-FV-4244 B CP-108 8

15. SG A Atmospheric Relief Valve MS-PV-3001 CP-108 A

16. SG B Atmospheric Relief Valve MS-PV-3002 CP-108 8

17. SG C Atmospheric Relief Valve MS-PV-3003 CP-108 A

, l e .

EMERGENCY CORE COOLING SYSTEMS ACCUMULATORS SHUT 00WN LIMITING CONDITION FOR OPERATION 3.5.1.2 Each reactor coolant system accumulator isolation valve shall be shut with power removed from the vai perator, r

APPLICABILITY: MODES 4* and 5' **

ACTION:

a. With one or more accumulator isolation valve (s) open and/or power available to the valve operator (s), imediately close the accumulator isolation valves and/or remove power from the valve operator (s).

b. The provisions of Specification 3.0.4 are not applicable for entry into MODE 4 from MODE 3.

SURVEILLANCE REQUIREMENTS 4.5.1.2 Each accumulator isolation valve will be verified shut with power removed from the valve operator at least once per 31 days.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to entry into MODE 3 from MODE 4 and if pressurizer pressure is greater than 1000 psig, each accumulator isolation valve shall be n open as required by Specification 3.5.1.1.a.

uregreaterth'an1i5b$sD,

' ~

SEABROOK - UNIT 1 ~ 3/4-5-3~

m CONTAINMENT SYSTEMS ,

PRIMARY CONTAINMENT CONTAINMENT LEAKAGE SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50;u ing the mcthed; and provisi;ns of AZ I/

N45,i-1072:

a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 i 10 month intervals during shutdown at a pressure not less than P,, 49.6 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection;

b. If any periodic Type A fails to meet 0.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 La '

a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumed;

c. The accuracy of each Type A test shall be verified by a supplemental test which:

1) Confirms the accuracy of the test by verifying that the supple-mental test result, L c, is in accordance with the following equation:

'll c~ (l am

bo)l 5 0.25 L, where L am is the measured Type A test leakage andoL is the superimposed leak;

2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and

3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment during the supplemental test is between 0.75 La and 1.25 L,.

SEABROOK - UNIT 1 3/4 6-3

e CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each 36-inch containment purge supply and exhaust isch tion valve shallbeverifiedtobelockedclosedatleastonceper31dayD 4.6.1.7.2 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard isolation valves with resilient material seals in each sealed closed 33-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L, when pressurized to P,,

4.6.1.7.3 At least once per 92 days each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.01 L, when pressurized to P,.

4.6.1.7.4 Each 8-inch containment purge supply and exhaust isolation valve snall be verified to be s uled closed or open in accordance with Specifi-cation 3.6.1.7.b at least once per 31 days.

l i

~~ -. -

If no centainment entry has been made since checking the inside containment -

l 36-inch valves locked closed, these valves do not have to be checked until a containment entry has been made. -

[

! fe'3 ROOK - UNIT 1 3/4 6-13

4 4 a

~#

TABLE 3.7-3 e: AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT ('F)

Control Room 75 1.

Cable"Spreading R om 104 2

3.

Switchgear Room - Train A 104

4. Switchgear Room - Train B 104-Battery Rooms - Train A 97 5.

6. Battery Rooms - Train B 97

7. ECCS Equipment Vault - Train A 104

8. EcCS Equipment Vault - Trein B 104

9. - Certrifugal Charging-Pump Room - Train A 104

10. Centrifugal Charging Pump Room - Train B 104

11. ' ECCS Equipment Vault Stairwell - Train A 104

12. ECCS Equipment Vault Stairwell - Train B 104

13. PCCW Pump Area- . 104

. Cooling Tower Switchgear Room - Train A . 104

1. Cooting Tower Switchgear Room - Train B 104

16. Cooling Tower SW Pump Area 127

17. SW Pumptcuse Electrical Room - Train A 104 18, 5W Pumphouse Electrical Room - Train B 104

23. SW Pump Area 104 120

20. Die. 1 Generator Room - Train A

21. Diesel Generator Room - Train 9 120

22. EFW Pumphouse _

104

23. Electrical Penetration Area . Train A 100 Electrical Penetration Area - Train B 85 24

25. Fuel Storage Building Spent Fuel Pool Cooling 104 Pucs Area

26. Main Sten: and Feedw:ter Pipe Chase - East 130 4 A . Main. Steam and reedwat.er. Pipe, Chase - We:t,__._ _ _ . _ _.130 7 104 \' ~s

18. Hydrogen Analyzer Room

9. MSFW East Pipe Chase Electrical Room 104

' x - ,

SEABR00% - 3/4 7-23 J

_j

- ~~ , ,

A E+v4,' 2 4' ' -

di '

t ENCLOSURE 3 TO NYN-88092 4

I f

n

, , ,s- , --,- - - - , -

f ,' '

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,y GREATER THAN OR EQUAL TO 350*F LIMITING CONDITION FOR OPERATMN 4

3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be y

OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE Safety Injection pump,

c. One OPERABLE RHR heat exchanger,

d. One OPERABLE RHR pump, and

e. An OPERABLE' flow path

capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.

MPLICABILITY: MODES 1, 2, and 3**.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitteo to tne Commission pursuant to Specification 6.8.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the dsage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

0uring MODE 3, the discharge paths of both Safety Injection pumps may be isolated by closing for a pe=iod of up to 2 nours to perform surveillance testing as required by Speci'ication 4.4.6.2.2.

- The provisions of Specifications 3.0.4 ar.d 4.0.4 are not applicable for er.try into MODE 3 for the centrifugal charging pump aid Safety Injection pumps declared inoperable parsuant to Specification (4.5.3.2,provided the centrifugal charging pump and the Safety Injection pumps are Glored to OPERABLE <tatus within a';. least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RCS cold lege, exceeding 375'F, whichever comes first.

- ys. s. /. 2.

SEABROOK - UNIT 1 3/a 5-4

]

e

ML20150E163

Text

Subject:

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