Text

Forwards Revised Pages to Final Draft Tech Specs. Certification That Revised Final Draft Tech Specs Consistent W/Fsar,Ser,Ssers & as-built Facility Requested by 861014
	ML20211D512
Person / Time
Site:	Seabrook
Issue date:	10/10/1986
From:	Novak T; Office of Nuclear Reactor Regulation
To:	Harrison R; PUBLIC SERVICE CO. OF NEW HAMPSHIRE
References
	NUDOCS 8610220212
	Download: ML20211D512 (98)
	v • d • e

DCT10gggg DISTRIBUTION-Docket No.: 50-443 ~ Docket Files NRC PDR Mr. Robert J. Harrison Local PDR President and Chief Executive Officer PD#5 R/F MRushbrook Public Service Company of New Hampshire TNovak VNerses Post Office Box 330 JPartlow Manchester, New Hampshire 03105 BGrimes EJordan

Dear Mr. Harrison:

ACRS (101

SUBJECT:

CHANGES TO THE SEABROOK FINAL DRAFT TECHNICAL SPECIFICATIONS In a letter dated June 18, 1986, the Seabrook Unit 1 Final Draft Technical Specifications (TS) were transmitted to you for certification, under oath and affirmation, that the Final Draft TS are consistent with the Final Safety '

Analysis Report (FSAR), the Safety Evaluation Report (SER), and the as-built facility.

In a Public Service of New Hamphsire (PSNH) letter dated June 20, 1986, it was noted that certification was not possible until certain inconsistencies were eliminated. After a number of discussions between our staffs to climinate these inconsistencies a number of page changes were made to the Final Draft TS.

These were transmitted to you hy my letter of June 26, 1986.

In PSNH letter 4 fated July 15, 1986, it was noted that there were still inconsistencies in the additional changes. With additional discussions between our staffs to eliminate these inconsistencies, a number of additional page chan_ges to the Final Draft TS were made.

Enclosed are these page changes to the Final Draf t Seabrook Station, Unit 1 TS and they are heing forwarded to you for incorporation into the Final Draft transmitted you on June 18, Iq86.

You are requested to certify, under oath and affirmation, that the revised Final Oraft TS are consistent with the FSAR, the SER and its supplements and the as-built facility. Your certification is reouested by October 14, 1086.

For further information or clarification, please contact the Licensing Proiect Manager, Victor Nerses at (301) 497-8535.

Sincerely, Or W m19f m atYt 0610220212[=j$043 A00CK PDR PDR Thomas M. Novak, Acting Director A

Division of PWR Licensinq-A Office of Nuclear Reactor Regulation

Enclosure:

/\s stated cc:'Seinextpace *See previous concurrences PDf4

VNerses:es DIR:PD#5

VNoonan ADS-A-TfWak 10/10/96 10/10/P6 10//O/86

00710 M86

/

Docket No.: 50-443 Mr. Robert J. Harrison President and Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105

Dear Mr. Harrison:

/ s

SUBJECT:

CHANGES TO THE SEABROOK FINAL. DRAFT TECHNICAL. PECIFICATIONS In a letter dated June 18, 1986, the Seabrook Unit Final Dra ft Technical Specifications (TS) were transmitted to you for ce ification, under oath and 't affirmation that the Final Draft TS are'consisten with the Final Safety ~

AnalysisReport(FSAR),theSafetyEvaluationRfort(SER),andtheas-built facility.

In a Public Service of New Hamphsire, (PSNH ' letter dated June 20, 1986,- it was ntil certain inconsistencies were notedthatcertificationwasnotpossibleg)betweenourstaffstoeli eliminated. After a number of.discussio p these inconsistencies a number of page phanges were made to the Final Draft TS.

These were transmitted to you by my 1 tter of Jrne 26, 1986.

In PSNH letter dated July 15, 1986, it was noted that there were still inconsistencies in the additional thanges. With additional discussions between our staffs to eliminate t6ese inconsistencies, a number of additional page changes to the Final Draf S were made.

Enclosed are these page changes to the Final Draft Seabrook Station, Unit l' '

TS and they are being forwafded to you for incorporation into the Final Draft transmitted you on June 1 1986.

You are requested to certify, under oath and affirmation, that the revised Final Draft TS are con /istent with the FSAR, the SER and it supple; rents and the as-built facilit[ Your certification is requested by October 14, 1986.

For further informa' tion or clarification, please contact the !.icensing Project Manager, Victor ljdrses at (301) 492-8535.

/

/ Thomas M. Novak, Acting Director PWR Project Directorate No. 5

/

Enclosu're: As stated Divison of PW l.icensing-A

/

cc:/Seenextpage f Distribution:

cc: See next page

/ [ f -Occket NRC PDR Files BGrimes EJordan J' 5L DIR:PDJf A :. l.ocal PDR ACRS (10)

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/0/to/86 jy/l0/86 (g /gI /86

, TNovak VNerses JPartlow

/ %, ' UNITED STATES

[ g NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C, 20655 5 .j 0, a

%...../ 00I 10 % 5 i

Docket No.: 50-443 Mr. Robert J. Harrison President and Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 j Manchester, New Hampshire 03105

Dear Mr. Harrison:

SUBJECT:

CHANGES TO THE SEABROOK FINAL DRAFT TECHNICAL SPECIFICATIONS In a letter dated June 18, 1986, the Seabrook Unit 1 Final Draft Technical Specifications (TS) were transmitted to you for certification, under oath and affinnation, that the Final Draft TS are consistent with the Final Safety Analysis Report (FSAR), the Safety Evaluation Report (SER), and the as-built facility.

In a Public Service of New Hamphsire (PSNH) letter dated June 20, 1986, it was noted that certification was not possible until certain inconsistencies were eliminated. After a number of discussions between our staffs to eliminate these inconsistencies a number of page changes were made to the Final Draft TS.

These were transmitted to you by my letter of June 26, 1986.

In PSNH letter dated July 15, 1986, it was noted that there were still inconsistencies in the additional changes. With additional discussions between our staffs to eliminate these inconsistencies, a number of additional page changes to the Final Draft TS were made.

Enclosed are these page changes to the Final Draft Seabrook Station, Unit 1 '

TS and they are being forwarded to you #cr incorporation into the Final Draft transmitted you on June 18, 1986.

You are requested to certify, under oath and affinnation, that the revised Final Draft TS are consistent with the FSAR, the SER and its supplements and the as-built facility. Your certification is requested by October 14, 1986.

For further information or clarification, please contact the Licensing Project Manager, Victor Nerses at (301) 492-8535.

Sincerely, Thomas M. Novak, Acting Director Division of PWR Licensing-A Office of Nuclear Reactor Regulation

Enclosure:

As stated

. );c: See next page i

'O Mr. Robert J. Harrison

. Public Service Company of New Hampshire Seabrook Nuclear Power Station CC*

Thomas Dignan, Esq. E. Tupper Kinder Esq.

John A. Ritscher, Esq. G. Dana Bisbee, Esq.

Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State Hosue Annex Concord, New Hampshire 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear, Regulatory Commission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Sumer Street Mr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and I.icensing Yankee Atomic Electric Company Robert A. Backus, Esq. 1671 Worchester Road O'Neil, Backus'and Spielman Framingham, Massachusetts 01701 116 t.owell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager United Engineers & Constructors William S. Jordan, III 30 South 17th Street Diane Curran Post Office Box 8223 Harmon, Weiss & Jordan Philadelphia, Pennsylvania 19101 20001 S Street, NW Suite 430 Washington, D.C. 20009 Hr. Philip Ahrens, Esq.

Assistant Attorney General State House, Station 6 Jo Ann Shotwell, Esq.

Augusta, Maine 0433; Office of the Assistant Attorney General Environmental Protection Division Mr. Warren Hall

One Ashburton Place Public Service Company of Boston, Massachusetts 02108 !

New Hampshira Post Office Bu 330 D. Pierre G. Cameron, Jr. , Esq.

Seabrook, New h mpshire 03874 General Counsel Public Service Company of New Hampshire Seacoast Anti-Pollution t.eague Post Office Box 330 ,

Ms. Jane Doughty Manchester, New Hampshire 03105 l 5 Market Street Portsmouth, New Hampshire 03801 Regional Administrator, Region I U.S. Nuclear Regulatory Comission Mr. Diana P. Randall 631 Park Avenue 70 Collins Street King of Prussia, Pennsylvania 19406 Seabrook, New Hampshire 03874 J

Richard Hampe Esq.

New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301

_ . _ ._ _ _ h. ___ _ _ __ _____I

'l Public Service Company of Seabrook Nuclear Power Station New Hampshire cc:

Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent, City Hall Chairman ,

126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03601 Town of Salisbury, MA 01950 Ms. l.etty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C. 20510 Ms. Roberta C. Pevear Mr. Owen B. Durgin Chairman Town of Hampton Falls, New Fampshire Durham Board of Selectmen Drinkwater Road Town of Durham Hampton Falls, New Harpshire 03844 Durham, New Hampshire 03824 Ms. Sandra Gavutis Charles Cross, Esq.

Town of Kensington, New Hampshire Shaines, Mardrigan and RDF 1 McEaschern East Kingston, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Portsmouth, New Hampshire 03801 Chairman, Board of Selectmen RFD 2 South Hampton, New Hampshire 03827 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention ,

Mr. Angie Machiros, Chairman Comittee Board of Selectmen c/o Rye Town Hall for the Town of Newbury 10 Central Road Newbury, Massachusetts 01950 Rye, New Hampshire 03870 Ms. Cashman, Chairman Jane Spector Board of Selectmen Federal Energy Regulatory Town of Amesbury Comission Town Hall 825 North Capital Street, NE Amesbury, Massachusetts 01913 Room 8105 Washington, D. C. 20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division City Hall Public Service of New Hampshire '

Newburyport, Massachusetts 01950 Company 7910 Woodmont Avenue Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter 10 Front Street Mr. William B. Derrickson Exeter, New Hampshire 03823 Senior Vice President Public Service Company of i New Hampshire '

Post Office Box 700, Route 1 Seabrook, New Hampshire 03874

l

..- ~

INDEX 1.0 DEFINITIONS SCT 101986 SECTION PAGE 1.1 ACTI0N........................................................ 1-1 1.2 ACTUATION LOGIC TEST..........................................- 1-1 1 1.3 ANALOG CHANNEL OPERATIONAL TEST............................... 1-1 1.4 AXIAL FLUX DIFFERENCE......................................... 1-1 1.5 C H ANN E L C A LI B RATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.6 CHANNEL CHECK................................................. 1-1 1.7 CONTAINMENT INTEGRITY.........................................~ 1-2 1 1.8 CONTROLLED LEAKAGE............................................ 1-2 1.9 CORE ALTERATION............................................... 1-2 1.10 00SE EQUIVALENT I-131........................................ 1-2 1.11 E - AVERAGE DISINTEGRATION ENERGY............................ 1-2 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.13 FREQUENCY N0TATION..........................................: 1-3

1.14 IDENTIFIED LEAKAGE........................................... 1-3

[ 1.15 7

MASTER REtAY 1EST............................................ 1-3 1.16 MEMBER (S) 0F THE PUBLIC...................................... 1-3 1.17 0FFSITE DOSE CALCULATION MANUAL.............................. 1-3 1.18 OPERABLE - OPERABILITY....................................... 1-4

{ 1.19 OPERATIONAL MOLE - M0DE...................................... 1-4 1.20 PHYSICS TESTS................................................ 1-4 1.21 PRESSURE BOUNDARY LEAKAGE.................................... 1-4 1.22 PROCESS CONTROL PR0 GRAM...................................... 1-4 1.23 PURGE - PURGING.............................................. 1-4 1.24 QUADRANT POWER TILT RATI0.................................... 1-5 1.25 RATED THERMAL P0WER.......................................... 1-5 1.26 REACTOR TRIP SYSTEM RESPONSE TIME............................ 1-5 1.27 REPORTABLE EVENT............................................. 1-5 1.28 CONTAINMENT ENCLOSURE BUILDING INTEGRITY..................... 1-5 1.29 SHUTDOWN MARGIN.............................................. 1-5 1.30 SITE B0VNDARY................................................ 1-5 1.31 SLAVE RELAY TEST............................................. 1-6 1.32 SOLIDIFICATION............................................... 1-6 1.33 SOURCE CHECK................................................. 1-6

}

1.34 STAGGERED TEST BASIS......................................... 1-6 1.35 THERMAL P0WER................................................ 1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 l 1.37 UNIDENTIFIED LEAKAGE......................................... 1-6 l 1.38 UNRESTRICTED AREA............................................ 1-6 1.39 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-7 1.40 VENTING...................................................... 1-7 1.41 1 F A P Y f" in'n w a t_

-2 i' T n f" A TLA F if*f" PUPTPM f r$ 1 P f masr L m im u m "y g ...........................

ay 1p/

GA% cut. MDwAME. TREM MEwT EsTETEMJ i TABLE 1.1 FREQUENCY N0TATION.................................... . 1-8 TABLE 1.2 OPERATIONAL M0 DES..................................... ,. 1-8 ,

{$ GA6Em> RADM TERW ymA t - L Sfs.Toi

_ % f.14 a J. a em .k]

SEABROOK - UNIT 1 i

" o DEFINITIONS DRAFT om . _

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall t,e that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION -

1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

UmERT % P %*T3 IDENTIFIED LEAKAGE 1.1/IDENTIFIEDLEAKAGEshallbe: '

s'

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.1hAMASTERRELAYTESTshallbetheenergizationofeachmasterrelayand verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (S) 0F THE PUBLIC a 1.lfMEMBER(S)0FTHEPUBLICshallincludeallpersonswhoarenotoccupa-tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recre-ational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.1 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain in Part A the radiological effluent sampling and analysis program and radiologichl environ-mental monitoring program. Part B of the ODCM shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

SEABROOK - UNIT 1 1-3

. . O DEFINITIONS vos a y g OPERABLE - OPERABILITY 9

1.15 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - HODE _

to 1.}s An OPERATIONAL MODE (i.e. , MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS .

1.2b PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation:

(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.2kPRESSUREBOUNDARYLEAKAGEshallbeleakage(exceptsteamgeneratortube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM 3

1.2/ The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State Regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste.

PURGE - PURGING 4

1.2/ PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

O SEABROOK - UNIT 1 1-4 j 1

o DEFINITIONS QUADRANT POWER TILT RATIO 6 00710 m 1.2/ QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated output ~s, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER _

c.

1.2$ RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from .

when the monitored parameter etceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT t

1.21 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

CONTAINMENT ENCLOSURE BUILDING INTEGRITY 9

1.2$ CONTAINMENT ENCLOSURE BUILDING INTEGRITY shall exist when:

a. Each door in each access opening'is closed except when the access opening is being used for normal transit entry and exit,

b. The Containment Enclosure Filtration System is OPERABLE, and

c. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

SHUTDOWN MARGIN 3o 1 25 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.3$, The SITE BOUNDARY shall be that line beyond which the land i's neither owned, nor leased, nor otherwise controlled by the licensee.

1 SEABROOK - UNIT 1 1-5

1 DEFINITIONS ;

SLAVE RELAY TEST 2.

1.37 ~A SLAVE RELAY TEST shall be the energization of each slave relay D and 1 0 1986 verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation d,evices.

~

SOLIDIFICATION l

S 1.3/ SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. -

SOURCE CHECK 1.3/ A SOURCE CHECK shall be the qualitative assessement of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS .

I 1.3fASTAGGEREDTESTBASISshallconsistof:

a. A test schedule for n systems, subsystems, trains, or other I designated components obtained by dividing the specified test '

interval into n equal subintervals, and

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 6

1.3$ THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.3% A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE 8

1.3/ UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA

~

1.36 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

SEABROOK - UNIT 1 1-6

1

. DEFINITIONS VENTILATION EXHAu3T TREATMENT SYSTEM MIIO1986 j Au l 1 75 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal ;

adsorbers and/or HEPA filters for the purpose of removing iodines.or particu-lates from the gaseous exhaust stream prior to the release to the environment. 1 Such a system is not considered to have any effect on noble gas effluents.

Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING t

1.4 VENTING shall be the controlled process of discharging air or gas from a

. confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air o'r gas is not provided or required during VENTING. Vent, used in system names, does not iniply a VENTING process. __

dD0 m ..J. 5.v.76.

I 3,M.. M. 1.nmm. m P M., 1_.7.en. D.. Y, "

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lA 1.A1 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

O SEABROOK - UNIT 1 1-7

l l

TABLE 3.3-1

! m g REACTOR TRIP SYSTEM INSTRUMENTATION 8

g MINIMUM

, TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

' U 1. Manual Reactor Trip 2 1 2 1, 2 1

- 2 3* , 4* , 5*

1 2 10

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 1, 2 2#

b. Low Setpoint 4 2 3 1###, 2 2#

3. Power Range, Neutron Flux 4 2 3 1, 2 2#

High Positive Rate

4. Power Range, Neutron Flux, 4 2 3 1, 2 2#

High Negative Rate

5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3

6. Source Range, Neutron Flux

a. Startup 2 1 2 2## 4

b. Shutdown 2 0 1 3,4,5 5

c. Shutdown 2 1 2 3* , 4* , 5* - 10

7. Overtemperature AT 4 2 3 1, 2 6#

8. Overpower AT 4 2 3 1, 2 6#

9. Pressur,izer Pressure--Low 4 2 3 1** 6# (1)

10. Presshfizer Pressure--High 4 2 3 1, 2 6# (1)

11. Pressurizer Water Level--High 3 2 2 1** 6/g

. o

, 4 u

1

TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION E

8

^

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Z 12. Reactor Coolant Flow--Low

~ a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 6[#

any oper- each oper-ating loop ating loop

b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 6 /#

below P-8) two oper- each oper- ,

ating loops ating loop

13. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 6# (1)

Level--Low-Low in any oper- each oper-R ating stm. ating stm.

gen. gen.

14. Undervoltage--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus 1** 6#

Pumps

15. Underfrequency--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus 1** 6#

Pumps

16. Turbine Trip

a. Low Fluid Oil Pressure 3 2 2 1*** 6/#

b. Turbine Stop Valve Closure 4 4 4 1*** 11#

17. Safety Injection Input from ESI, 2 1 2 1, 2 ,

9

18. Reactor Trip System Interlocks '

a. Intermediate Range Neutron Flux, P-6 g

2 1 2 2## 8 e

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued) 00T24mg ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL P0WER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to in::reasing THERMAL POWER above the P-6 Setpoint, and _

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum.

' Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers, suspend all operations involving positive reactivity changes and verify that valve RMW-V31 is closed and secured in position within the next hour.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per

""I) Specification 4.3.1.1.

ACTION 7 IWith the number of OPERABLE channels one less than the Total N, '

Number of Channels, STARTUP and/or POWER OPERATION may proceed untilperformanceofthenextrequiredANALOGCHANNELOPERATIONAL)

TEST provided tcondition within 6the inoperable channel is placed in the tripped >M hours.

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the interlock is in its ~ required state for the existing plant condition, cr apply Specification 3.0.3.

SEABROOK - UNIT 1 3/4 3-6

{

TABLE 4.3-1 (Continued)

, TABLE NOTATIONS 007 le ps

0nly if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.

- Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. .

~

- - Below P-10 (Low Setpoint Power Range Neutron Flux Interlock)-Setpoint.

(1) If not performed in previous 31 days. -

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE '

above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Initial platesu curves shall be measured for each detector. Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for antry into MODE 2 or 1.

I (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED g TEST BASIS.

(8)fWithpowergreaterthanorequaltotheInterlockSetpointtherequired)

ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive annun-(ciator window. ]Q (9) Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(10) Setpoint verification is not applicable. . .

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

SEABROOK - UNIT 1 3/4 3-12

TABLE 3.3-13 (Continued)

TABLE NOTATIONS At all times.

- GMEcus RADwASTE VtERTMEMT During "AOI0 ACTIVE CAC tlA';TE SYSTEM operation.

- - When the gland seal exhauster is in operation. -

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

ACTION STATEMENTS ACTION 32 -

With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at.least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 33-With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For RM-6504, RM 6503 may be used as an alternate.

ACTION 34 -

With the number of channels OPERABLE less than'the Minimum Channels OPERABLE requirement, operation of this RADIOACTIVE GAS WASTE SYSTEM may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 35 -

With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in the ODCM.

9 SEABROOK - UNIT 1 3/4 3-63

)

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE OCT 1 o 1986 SURVEILLANCE REQUIREMENTS 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit: ,

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, .

c. Prior to returning the valve to service following maintenance, '

repair, or replacement work on the valve, and

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

7 The provisions of Specification 4.0.4 are.not applicable,for entry into MODE 3 or 4.

e. Q eM i-! L A ASME h , A-tc._ E ,g h IWV-341 g

O e

SEABROOK - UNIT 1 3/4 4-23

~

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS .

4.6.1.2 (Continued)

d. Type B and C tests shall be conducted with gas at a pressure not less than P,, 49.6 psig, at intervals no greater than 24 months except for tests involving:

1) Air locks, and

2) Purge supply and exhaust isolation valves with resilient material seals. -

e. The combined bypass leakage rate shall be determined to be less than or equal to 0.E0 L, by applicable Type B and C tests at least once per 24 months.

f. Purge supply and exhaust isolation valves with resiJent material seals shall be tested and demonstrated OPERABLE by the requirements of Speci-fication 4.6.1.7./ or 4.6.1.7.(, as applicable; 2 3

g. Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3; and

h. The provisions of Specifications 4.0.2 are not applicable.

l l

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

l

. o CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OCTIe125 CONTAINMENT VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS .

4.6.1.7.1 Each 36-inch containr.ent purge supply and exhaust isolation valve shall be verified to be locked closed at least once per 31 days. _

4.6.1.7.2 At least once per 6 months on a STAGGERED TEST BASIS cc:5 'ecked-closed 36-inch containment purge supply and exhaust 4 chti:r. ;,&c with

- rer!'hnt : m 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 containstnt 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 shall be verified to be sealed closed or open in accordance with Specifi-cation 3.6.1.7.b at least once per 31 days.

pyt ; _ ) fik d __ _1,,,,,,1 pt g g AAh- ..- % _e.s k ,, .11 t '

k

8 t 3 Aa.II.o a d

_m. _ L },

j

?

i I

O SEABROOK - UNIT 1 3/4 6-13

TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY OCT 101986 SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY -

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

Determination % _

2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 10% of the allowable limit for ,

radiofodines~ .

b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10%

of the allowable limit for radioiodines.

A gross radioactivity analysis shall consist of the quantitative measurement i of the total specific activity of the secondary coolant except for radio- g t

nuclides with half-lives less than 10 minutes. Determination of the i contributors to the gross specific activity shall be based upon those energy j peaks identifiable with a 95% confidence level. >

O SEABROOK - UNIT 1 3/4 7-8

RADIOACTIVE EFFLUENTS GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE - SYSTEM OCT2sEE6 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:

GASE h RA%JASTE TOEATHm r

a. With the concentration of oxygen in the "ASTC OAS l;0 LOUR 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 GASEOUS RADWASTE TREAT-MENT SYSTEM shall be determined to be within the above limit by continuously monitoring the waste gases in the GASEOUS RADWASTE TREATMENT SYSTEM with the hydrogen or oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.10.

I O e 9

SEABROOK - UNIT 1 3/4 11-9

INSTRUMENTATION 00710 m BASES I

3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES l ACTUATION SYSTEM INSTRUMENTATION (Continued) - !

uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z + R S < TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. -Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy o eir measurement. TA or Total Allowance is the difference, in percent spa or Rack Error is the "as measured" deviation, in the percent span, f the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4,

in percent span, from the analysis assumptions. Use of Equation 3.3-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated'as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time.

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitiye to combina-tions indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety SEABROOK - UNIT 1 B 3/4 3-2

RADI0 ACTIVE EFFLUENTS BASES N GASE0US EFFLUENTS

' 3/4.11.2.3 -DOSE - 10 DINE-131, 10 DINE-133 TRITIUM, AND RADIOACTIVE 'ATERIAL M

IN PARTICULATE FORM (Continued) that the releases of radioactive materials in gaseous effluents at the SITE BOUNDARY will be kept as low as is reasonably achievable. The ODCM ,calcula-tional methods specified in the Surveillance Requirements implement the require-ments in Section III.A of Appendix I that conformance with the guides of Appen-dix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodol-ogy and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regula -

tory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for l Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine l Releases from Light-Water-Cooled Reactors," Revision 1. July 1977. These equa- !

tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were exam-ined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vege-tation with subsequent consumption by man, (3) deposition of radionuclides onto grassy areas where milk animals and meat producing animals graze followed by human consumption of that milk and naat, and (4) deposition of radionuclides on the ground followed by subsequent human exposure.

3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM cmn. eAmere Tymrmewr The OPERABILITY 'of'the MSTE C.*i ZLO'JP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when speci-fled, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable. This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits govern-ing the use of apprcpriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.'B pnd II.C of Appendix I to 10 CFR Part 50, for gaseous effluents.

SEABROOK - UNIT 1 B 3/411-4 8

{. .

_ _ . . . _ _ _ _ _ _ . . .. ._.__....____..._..m_..m.____.g.u 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 COMMUNICATIONS....... ................................... 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 Wa t e r L e v e 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 LEVE L - STO RAG E POO L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-1E 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

?/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 R EACTO R COO LANT L00PS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 Li qui d Hol dup Tan ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS 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 Gas Storage Tanks....................................... 3/4 11-10 3/4.11.3 SOLID RADI0 ACTIVE WASTES.............................l..~. 3/4 11-21 3/4.11.4 TOTAL 00SE..............................................".

3/4 11-1z 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 SEABROOK - UNIT 1 ix

__m___-- - - - ~ _ - _ - _ - -

INDEX BASES SECTION. PAGE 3/4.9.8 RESIOUAL HEAT REMOVAL AND COO LANT CIRCULATION. . . . . . . . . . . . . . B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM. . . . . . . . . . . ~. B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STO RAGE P00 L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM....... B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10. 4 REACTO R COO LANT L00PS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS........................................ B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS....................................... B 3/4 11-2 3/4.11.3 SOLID RADIOACTIVE WASTES............................c... B 3/4 11-5 3/4.11.4 TOTAL 00SE.............................................. B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0GRAMc..................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 5.0 DESIGN FEATURES 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1. 2 LOW POPULATION Z0NE......................................... 5-1 5.1. 3 MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS.................... 5-1 FIGURE 5.1-1 SITE AND EXCLUSION AREA B0VNOARY..................... 5-3 FIGURE 5.1-2 LOW POPULATION 20NE.................................. 5-5 FIGURE 5.1-3 LIQUID EFFLUENT DISCHARGE LOCATION................... 5-7 5.2 CONTAINMENT .'

5. 2.1 CONFIGURATION.'.............................................. 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-9 SEA 8 ROOK - UNIT 1 xii

--h ** * * - - -

._g'.pr> -M WMe h mF.*

l t

i 3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E SENSOR R TOTAL ERROR

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

,g E 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

3

-4 j e 2. Power Range, Neutron Flux q i j a. High Setpoint 7.5 '4.56 0 $109% of RTP*

.- $111.1% of RTP*

Aj b. Low Setpoint 8.3 4.56 0 125% of RTP*

.[ '

$27.1% of RTP*

S 3. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with 1 High Positive Rate- i time constant a time constant j

12 seconds 12 seconds "l

4. Power Range, Neutron Flux, 1.6 0.5 0 High Negative Rate $5% of RTP* with 16.3% of RTP* with m a time constant a time constant 7 4 >

__2 seconds >

_2 seconds

.j 5. Intermediate Range, 17.0 8.41 0 125% of RTP* 131.1% of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps $1.6 x 105 cps

} t l 7. Overtemperature AT 6.5 3.31 1.04** See Note 1 See Note 2 I j +0.478* !

8. Overpower AT 4.8 1.43 0.12 See Note 3 See Note 4 f

j 9. Pressurizer Pressure - Low 3.1 0.71 '1.69 11945 psig 11,935 psig 3 .. -

[

10. Pressurlier Pressure - High 3.1 0.71 1.69 $2385 psig i $2,395'psig

.- *RTP = RATED THERMAL POWER ij. **The sensor error for T,yg is 1.04 and the sensor error for Pressurizer Pressure is 0.47. "As measured"

'j 4 sensor errors may be used in lieu of either or both of these values, which then must be summed to deter-mine the overtemperature AT total channel value for S.

.f 1

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE -

The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation that would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and, therefore, THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the W-3 (R-Grid) correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is 'ndicative of the margin to DNB.

The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating

, conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, Fh, of 1.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance isincludedforanincreaseinFhatreducedpowerbasedontheexpression:

. Fh = 1.55 (1+ 0.2 (1-P)]

Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for i

the range of all control rods fully withdrawn to the maximum allowable control rod insertion, assuming the axial power imbalance is within the limits of the

f t(AI) fur.ction of the Overtemperaturn trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-i' temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

SEA 8 ROOK - UNIT 1 8 2-1

_ i_ _ l 1 __ . . . _ - . _ _ - - - -

-- - i -

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T, 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 3.8% Ak/k in MODES 1, 2, and 3 and 1.3% Ak/k in MODE 4.

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

ACTION:

With the SHUTDOWN 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 ppe boron or equivalent until the required -

SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limiting 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 SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); .

l b. When in MODE 1 or MODE 2 with k,ff greater than or equal to 1 at i .

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; l 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 i

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 Specification 3.1.3.6; and .

l 1

j *See Special Test Exceptions Specification 3.10.1.

l

'l SEA 8 ROOK - UNIT 1 3/4 1-1 j ..

li;F5d6 hv'5W"r="+7T h--

9 REACTIVITY CONTROL SYSTEMS BORATION CONTROL SHUTDOWN MARGIN - T,yg LESS THAN OR EQUAL TO 200*F

, LIMITING CONDITION FOR OPERATION -

1 ,~ ,

3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.2% Ak/k.

l Additionally, the Reactor Coolant System boron concentration shall be greater s

than or equal to 2000 ppe boron when the reactor coolant loops are in a drained condition.

APPLICABILITY: MODE 5.

ACTION:*

With tfie SHUTDOWN MARGIN less than 1.2% Ak/k or the Reactor Coolant System boron concentration less than 2000 ppm boron, 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 SHUTDOWN MARGIN and baron concentration are restored.

l SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN and boron concentration shall be determined to be greater than or equal to 1.2% Ak/k: .

^

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 SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

/

1) Reactor Coolant System boron concentration,

! 2) Control rod position, l

3) Reactor Coolant System average temperature, i

4) Feel burnup based on gross thermal energy generation, .

s' *

5) Xenon. concentration, and :

1

6) Samarium concentration. !

SEABROOK - UNIT'1 1/4 1-3

+4P7

s * , *

-._._n. - . _

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target bands (flux difference units) about the target flux difference:

a. i 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU; _

b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTU; and

c. + 3%, -12% for each subsequent cycle.

The indicated AFD may deviate outside the above required target band at. greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the in'di-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD may deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour dbring the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.*

ACTION:

~

a. With the indicated AFD outside of the above required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15, minutes either:

1. Restore the indicated AFD to within the target band limits, or

2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

b. With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:

1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and

2. The Power Range Neutron Flux * ** - High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions Specification 3.10.2.

- Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ' SEABROOK - UNIT 1 3/4 2-1

_._a -. . . -

-- - - -a POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE 1

LIMITING CONDITION FOR OPERATION 3.2.1 -

ACTION: (Continued)

c. With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time duri~ng the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

. a. One minute penalty deviation for each 1 minute of POWER OPERATION t

outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half-minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full-Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The ^arget flux difference shall be updated at least once per '

31 Effective Full-Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation oetween the most recently measured value and the predicted value at the end of the cycle life. The previsions of Specification 4.0.4 are not applicable.

i

- (Continued) operation may be accumulated with the AFD outside of the above l l required target band during testing without penalty deviation. I SEABROOK - UNIT 1 3/4 2-2 y- :-

_ :- _ - _ = _ _ -

? -

-= - - - . - -

L.l - _- A POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-Fg SURVEILLANCE REQUIREMENTS .

4.2.2.2d. (Continued) .

2) When the F x is less than or equal to the FxRTP limit for the appropriate measured core plane, additional power distribution

. maps shall be taken and F C compared to.F and Fxy at least l

xy once per 31 EFPD.

e. The Fxy limits for RATED THERMAL POWER (F ) shall be provided for all core planes containing Bank "D" control rods and all unrodded.

core planes in a Radial Peaking Factor Limit Report per Specifica

tion 6.8.1.6; The F xy limits of Specification 4.2.2.2e., above, are not applicable f.

in'the following core planes regions as~ measured in-percent of core height from the bottom of the fuel:

1) Lower core region from 0 to 15%, inclusive,

2) Upper core region from 85 to 100%, inclusive,

3) Grid plane regions at 17.8 t 2%, 32.112%, 46.4 i 2%, 60.612%,

and 74.9

2%, inclusive, and

4) Core plane region ~s within i 2% of core height i 2.88 inches about the bank demand position of the Bank "D" control rods.

0 l

. g. With F exceeding Fxy, the effects of Fxy n Fq(Z1 shall be evaluated to determine if F q (Z) is within its limits.

4.2.2.3 When Fq (Z) is measured for other than Fxy determinations, an overall measured qF (Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

l SEABROOK - UNIT 1 3/4 2-7 W W'

- = _ . . . .

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

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 Fhshallbelessthan 1.49 [1.0 + 0.2 (1-P)]. -

i Where: P= THERMAL POWER ~

RATED THERMAL POWER -

~

APPLICABILITY: MODE 1. ,

ACTION:

With Fh exceeding its limit:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required by ACTION a.,

above; THERMAL POWER may then be increased, provided Fh is demonstrated through incore mapping to be within its limit.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 Fhshallbedemonstratedtobewithinits'limitpriortooperation above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER.

J

b. UsingthemeasuredvalueofFhwhichdoesnotincludeanallowance for measurement uncertainty, i l l l

l l

i l

l SEABROOK - UNIT 1 3/4 2-8 l l

- - ~ Z_ Zi_ __ - - -. -- _ - - - ?---- --- - - -~ ~ -- I

l

. t

,t TABLE 3.3-3 L !

m 5!

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

!=

is

' MINIMUM -!

TOTAL NO. CHANNELS CHANNELS APPLICABLE j E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

.. . -4

')l w 1. Safety Injection (Reactor Trip, Feedwater Isolation, l Start Diesel Generators, a

Phase "A" Isolation, r

'l Containment Ventilation [,

Isolation, Emergency f Feedwater, Service Water I to Secondary Component Cooling Water Isolation, !

CBA Emergency Fan / Filter (

R

Actuation, and Latching

, Relay).

6 M a. Manual Initiation 2 1 2 1,2,3,4 17

b. Automatic Actuation 2 1 2 1,2,3,4 13

Logic and Actuation

]l Relays ,

c. Containment 3 2 2 1,2,3 14*

Pressure--Hi-1 *

d. Pressurizer 4 2 3 1, 2; 3# 18* l Pressure--Low l4

(

e. St'eam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# '.

14* I Pressure--Low any steam ' '

line 1

-] [

ll

, .?

.i !

I TABLE 3.3-4

\ v.

i

{ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS, 8

7 SENSOR '

TOTAL ERROR ,

E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE a .

J i w 1. Safety Injection (Reactor Trip, ;

Feedwater Isolation, Start Diesel -

Generators, Phase "A" Isolation, l

Containment Ventilation Isolation, and Emergency Feedwater, Service i Water.to Secondary Component Cooling Water Isolation, CBA g Emergency Fan / Filter Actuation, and Latching Relay).

3

$ a. Manual Initiation N.A. N.A. N.A. N.A. N.A. i

\

T 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 i

d. Pressurizer Pressure--Low 13.1 10.71 1.69 1 1850 psig 2 1840 psig ;

e. SteamLhnePressure--Low 13.1 10.71 1.63 > 585 psig > 568 psig*

2. Containment Spray j a. Manual Initiation N.A. N.A. N.A. N.A. -

N.A.

if b. Adtomatic Actuation Logic N.A. N.A. N.A. N.A. N.A. l and Actuation Relays i .

d c. Containment Pressure--Hi-3 3.0 0.71 1.67 5 18.0 psig i 18.7 psig '

- r I'

.i .

- . _ _ _ _ - _ , _ - . -m._-----

- _m TABLE 3.3-4 (Continued)

TABLE NOTATIONS

Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are T1 > 50 seconds and tg 1 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

- The time constant utilized in the rate-lag controller for Steam Line ' Pressure-Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant.is adjusted to this value.

P 6

SEABROOK - UNIT 1 3/4 3-29 ;

1

TABLE 4.3-2 u,

9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

,E SURVEILLANCE REQUIREMENTS o ,

I TRIP ANALOG ACTUATING MODES 4 E CHANNEL DEVICE MASTER SLAVE FOR WHICH U CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE

I w FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRE 0 t

1. Safety Injection (Reactor Trip, l; Feedwater Isolation, Start Diesel j; Generators, Phase "A" Isolation,

! Containment Ventilation Isolation, Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency

[

Fan / Filter Actuation, and Latching l y Relay).

y a. Manual Inftiatton N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 O b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 i Logic and Actuation f Relays [

c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 l Hi-1 [

d. Pressurizer Pressure S R M N.A. N.A. N.A. N.A. 1, 2, 3 f

]j Low i

e. Steam Line S R M N.A. N.A. . N.A. N.A. 1,2,3 i Pressure-Low I I 2. Containme'nt Spray ,

a. Manual Initiation N.A. N.A. N.A. R N.A. , N. A'. ,, N.A. 1,2,3,4 1 b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1). -Q 1,2,3,4 e

l Logic and Actuation i

Relays l!

' C. Containment Pressure- S R M N.A. s N.A. N.A. N.A. 1,2,3 l Hi-3 y 7

l L'

s I

TABLE 3.3-6 -

I r m RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS l 9 '

E MINIMUM 8 CHANNELS CHANNELS APPLICABLE ALARM / TRIP

f. FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE ' MODES- SETPOINT ACTION

,g 1. Containment

Q a. Containment - Post LOCA - 1 2 All 27

.s Area Monitor -< 10 R/h

b. RC5 Leakage Detection }

, 1) Particulate Radioactivity N.A. I 1,2,3,4 N.A. 26 1

2) Gaseous Radioactivity N.A. I 1,2,3,4 N.A. 26

2. Contairment Ventilation Isolation I a. On Line Purge Monitor 1 2

i 1, 2, 3, 4 23

b. Manipulator Crane Area Monitor 1 2 5, 6 ** 23 I t' 3. Main Steam Line 1/ steam line 1/ steam 1, 2, 3, 4 N.A. 27-1 {: line i

l

! 'i' w

4. Fuel Storage Pool Areas

" . [,

'l a. Fuel Storage Building i t

Exhaust Monitor N.A. 1 *** **** 25

5. Control Room Isolation [

a. Air Intake-Radiation Level

}. 1) East Air Intake 1/ intake 2/ intake All **** 24

] . 2) West Air Intake 1/ intake 2/ intake All **** 24 j~

j' 6. Primary Component Cooling Water I a. Loop A 1 'li All <2x 28 i 7 Background -

b. Loop 6 1 1 All <2x

, 28 .'

j Hac,kground l

TABLE NOTATIONS j

Two times background; purge rate will be verified to ensure compliance with Specification 3.11.2.1 requirements.

j ** Two times background or 15 mR/hr, whichever is greater. -

j *** With irradiated fuel in.the fuel storage pool areas.

- - - Two times background or 100 CPM, whicheverlis greater. .

i .,

) !

i

TABLE 4.3-3

! n I RADIATION MONITORING INSTRUMENTATION FOR PLANT E' l

OPERATIONS SURVEILLANCE REQUIREMENTS

= ;

8

^ DIGITAL t

' CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE 3

E FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED

1. Containment

]

a. Containment - Post LOCA -

l Area Monitor S R M All l b. RCS Leakage Detection r

1) Particulate Radio- S R M 1,2,3,4

! activity l

[ '2) Gaseous Radioactivity S R M 1,2,3,4 [

4

2. Containment Ventilation Isolation

a. On Line Purge Monitor S R M 1, 2, 3, 4 R b. Manipulator Crane Area S R M 5, 6 f

Monitor [

7

.w

$ 3. Main Steam Line S R M 1, 2, 3, 4 i

l

~

(

4. Fuel Storage Pool Areas t I

a. Radioactivity-High-Gaseous Radioactivity S R M *

5. Control Room Isolation

a. Air Intake Radiation Level f

1) East Air Intake S R H All i

2) West Air Intake S R H All j 6. Primary Component Cooling Water i!

j a. Loop.A. S R H All

b. Loop B S R M All t

f TABLE NOTATIONS .

With irradiated fuel in the fuel storage pool areas.

d

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

INSTRUMENTATION MONITORING INSTRUMENTATON MOVABLE INCORE DETECTORS i

LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

a.

At least 75% of the detector thimbles. _

b. A minimum of two detector thimbles per cora quadrant, and

c. Sufficient movable detectors, drive, and rea s ut equipment to map these thimbles.

APPLICABILITY: When the Movable Incore Detection System is used for: .

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or N

c.

Measurement of Fg , pq(Z) and F xy.

ACTION:

With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.D.4 are not applicable.

SURVEILLANCE REQUIREMENTS (Plant procedures are used to determine that the Movable Incore Detection System is OPERABLE.)

e SEABROOK - UNIT 1 3/4 3-40

_ __n___.________.c_,_.____.

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

INSTRUMENTATION j MONITORING INSTRUMENTATION

! SEISMIC INSTRUMENTATION l LIMITING CONDITION FOR OPERATION .

3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE. -

i APPLICA8ILITY: 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 i-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 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 submitted to the Commission pursuant to Specificatior. 6.8.2 within 14 days describing the magnitude, frequency spectrum, and resultant effect upon facility features important to safety.

~

I i

SEABROOK - UNIT 1 3/4 3-41

= ~ ~ ~ ,

f TABLE 3.3-9 (Continued) j y, REMOTE SHUTDOWN SYSTEM [

i j Q

=

3' l' READOUT 8 TRANSFER SWITCHES / CONTROL CIRCUITS LOCATION f

! ^ !

l

18. SG D Atmospheric Relief Valve MS-PV-3004 CP-108 8 i E 19. MS Isolation Valves MS-V-86/88/90/92 CP-108 A I y 20. MS Isolation Valves MS-V-86/88/90/92 CP-108 B g 21. Pressurizer Heaters, Group A !.

CP-108 A

1. 22. Pressurizer Heaters, Group B CP-108 B [

!; 23. Charging Pump CS-P-2A BUS S SWGR i

! 24. Charging Pump CS-P-28 [,

" BUS 6 SWGR ;.

25. Charging Pump Suction from RWST CS-LCV-1120 CP-108 A

! 26. Charging Pump Suction from RWST CS-LCV-112E j CP-108 8 t l 27. Pressurizer Relief Valve (PORV) RC-PCV-456A

'j j

28. Pressurizer Relief Valve (PORV) RC-PCV-456B CP-108 A CP-108 8

29. PORV Block Valve RC-V-122 CP-108 A

30. PORV Block Valve RC-V-124 CP-108 B y: R 31. High Pressure Injection SI-V-138 CP-108 A 1

32. High Pressure Injection SI-V-139 CP-108 B '

Y 33. VCT Discharge Isolation Valve CS-LCV-1128 CP-108 A

g; 34. VCT Discharge Isolation Valve CS-LCV-112C CP-108 8 i

j i

i i

A.

4 L t e 2

i >

t s

l l

1

'l TABLE 3.3-12 !

i j h RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l

i t

B o

i " }

e MINIMUM I E INSTRUMENT CHANNELS I 4 .

OPERABLE ACTION {,

r

, g 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release !

I a'. Liquid Radwaste Test Tank Discharge 1 29

b. Steam Generator Blowdown Flash Tank Drain 1* 30

c. Turbine Building Sumps Effluent Line 1 30

2. Flow Rate Measurement Devices N

a. Liquid Radwaste Test Tank Discharge 1 31 T ].

E b. Steae Generator Blowdown Flash Tank Drain 1* 31

c. Circulating Water Discharge 1** N.A.

3. Radicactivity Monitors Providing Alarm but Not Termination of Release y g

p,

a. Primary Component Cooling Water System (In lieu of service water monitors) 1 32 f,

4. Rate of Change Monitor I a. . Primary Component Cooling Water System Head Tank 1

33

]it (In lieu of service water monitors) i -

I

0nly applicable when steam generator blowdown is directed to the discharge transition structure. I t,

- Pump performance curves generated in place should be used to estimate " flow.

l k

P TABLE 3.3-13 -

h RAOI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION E i 8

^

I I

' MINIMUM CHANNELS INSTRUMENT I OPERABLE APPLICABILITY E ACTION y 1. ;

f RADI0 ACTIVE GAS WASTE SYSTEM EXPLOSIVE GAS g MONITORING SYSTEM !,.

I 3,

0xygeq Monitor (Process) 1 ** i 34 '

2. PLANT VENT-WIDE RANGE GAS MONITOR

a. Noble Gas Activity Monitor 1

i 33

[

b. Iodine Sampler 1

35 5:

8

c. Particulate Sampler 1

f.

+ 35 T d. Flow Rate Monitor 1

32

e. Sampler Flow Rate Monitor 1

32

3. GASEOUS WASTE PROCESSING SYSTEM (Providi.ng Alarm and Automatic Termination l of Release - RM 6504)

a. Noble Gas Activity Monitor (Process) 1

33

- t t *

(

.. I

! f i

d

h TABLE 3.3-13 (Continued) 3

. -l, 9 RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION E I t

E

' MINIMUM CHANNELS I INSTRUMENT OPERABLE

g. APPLICABILITY ACTION l' 4 4. # r TURBINE GLAND SEAL CONDENSER EXHAUST e

. a. Iodine Sampler 1 ***

35 I

b. Particulate Sampler 1 ***

35

}

c. Sampler Flow Rate Indicator 1 ***

32 I i

s I e Y i a e i

) .

4 E

J l I r;

r i <

1 E i -

l ..

t e

i i

i. .

l t i

I :

l l.

L--- _ _ .-. __

I-

i .

N ~ . - - - - - - I ~ ~~- -- A"~ ~

TABLE 3.3-13 (Continued)

TABLE NOTATIONS At all times.

i During RADI0 ACTIVE GAS WASTE SYSTEM-operation.

- - When the gland seal exhauster is in operation. -

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

ACTION STATEMENTS ACTION 32 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . .

ACTION 33- With the riumber of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per i

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within j 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For RM-6504, RM 6503 may be used as an alternate.

ACTION 34 - With the number of channels OPERA 8LE less than the Minimum Channels OPERABLE requirement, operation of this RADI0 ACTIVE GAS WASTE SYSTEM may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 35 - With the number of channels OPERA 8LE less than the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in the 00CM.

I I

SEA 8 ROOK - UNIT 1 3/4 3-63

- : . : :: - ' : L . L ======-

, , . - - ,-I TABLE 4.3-6 N

g RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i E i E CHANNEL MODES FOR WHICH g-

c. CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE -

g INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED H 1. RADI0 ACTIVE GAS WASTE SYSTEM EXPLOSIVE I GAS MONITORING SYSTEM f l

0xygen Monitor D N.A. Q(4) M **

l (Process) I j I

2. PLANT VENT-WIDE RANGE CAS MONITOR 6 I

a. Noble Gas Activity Monitor D M R(3) Q(2) *

!l l

t

$ b. Iodine Sampler W N.A. N.A. N.A.

f I i

[ c. Particulate Samp'er W N.A. N.A. N.A.

l * (-

l d. Flow Rate Monitor D N.A. R Q****

t e. Sampler Flow Rate Monitor * [

D N.A. R Q**** ;

J r

i J

[

j j **

e I

"! i ., E

(

i l -

t i t

f

('

ll .'

l l

l .

i TABLE 4.3-6 (Continued) !

l w

Q RA010 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS m ,

8

^ I

' CHANNEL MODES FOR WHICH l CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE E INSTRUMENT CHECK CHECK CALIBRATION TEST l

IS REQUIRED :

w 3. GASEOUS WASTE PROCESSING SYSTEM (Providing Alarm and Automatic L j Termination of Release) [

! g

a. Noble Gas Activity Monitor D N.A. R(5) Q(1)

l (Process) P

4. #

y TURBINE GLAND SEAL CONDENSER EXHAUST

a. Iodine Sampler W ***

, N.A. N.A. N.A s

b. Particulate Sampler W N.A. N.A. N.A. ***

. [

i q

$ c. Sampler Flow Rate Indicator D N.A. N.A. N.A. ***

l i

h

]1 1

t I.

! [

.. e i ,

t i I

i k (

e T l i .

l .

.. . - - w w .~.a. -- - ._. l .. ~ . _ . ,_. _ - . w .A ~ _.a 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 c 'c sist of a verification that the Radiation Data Management System (ROMS) indicated flow is consistent with the operational status of tha plant. -

f 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 tnat control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm Setpoir.t.

(3) The iniital CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) er using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These 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 oe 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 response is linear. Sources shall be used to verify the monitor response only for the intended energy range.

SEABROOK - UNIT 1 3/4 3-66

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

. - . ~ - , .

y _ ____. -.- _-A.2- u-.w.e - ~u h.--- ""-M REACTOR COOLANT SYSTEM 3/4.4.3 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. *

!. APPLICA8ILITY: MODES 1, 2, and 3.

~

ACTION:

4

a. With one or more PORV(s) inoperable, because of excessive seat

' leakage, within I hour either restore the ?3RV(s) to OPERABLE status

or close the associated block valve (s); c'.r-cwise, 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 43LD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . -

1 j b. With one PORV inoperable due to causes other tnan excessive seat

! leakage, within I hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANOBY 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 both PORV(s) inoperable due to causes other than excessive seat .

leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power ,

from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

, d. With one or more block valve (s) inoperable, within 1 hour:

! (1) restore the block valve (s) to OPERABLE status, or close the l'

block valve (s) and remove power from the block valve (s), or close the PORV and remove power from its associated solenoid valve; and (2) apply the ACTION b. or c. above, as appropriate, for the isolated '

, PORV(s). ,

e. The provisions of Specification 3.0.4 are not applicable, i

t 4

. G e

l t i SEA 8 ROOK - UNIT 1 3/4 4-11 i

!- x --7 _ ,__..c r.m=;= _ ,- w _ .:.-. --#-- .- - - ,---~~. - ..

. ,-..9._.-n... -. w.w ~ .- - w ~. - w .-- .,:. w - :- -" - w -v '

REACTOR COOLANT SYSTEM

! REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS

~

4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

~

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment drainage sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump .

seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days with the modulating valve fully open. The

! provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of a Reactor Coolant System water inventory balance -

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ~ after achieving steady-state operation

and at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter during steady-state operation, except that not more than 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months shall elapse between any two successive inventory balances; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i I

l "T,yg being changed by less than 5*F/ hour.

l SEA 8 ROOK - UNIT 1 3/4 4-22 4.~ - ~ . - - -. -- e

.._a........-..--...-...-~-..-a-2--w-~~-="---~'--'

-='"--'-:~""'".'"

~

TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS !

STEADY-STATE TRANSIENT ~

PARAMETER LIMIT LIMIT Dissolved Oxygen * < 0.10 ppm $ 1.00 ppe.

Chloride < 0.15 ppe $ 1.50 ppe Fluoride < 0.15 ppa 1 1.50 ppe -

Limit not applicable with T,yg less than or equal to 180*F.

P s

s O

SEA 8 ROOK - UNIT 1 3/4 4-26

_ cr : . x= n . , == . . .m -- . m -- . .~. -; .'. L.. - . - - - . ---- - - , - --

_ . .Y_ - A.~ w N N ~ . a.-c a - - - - c.~ Na--- ~~-M -

Controlling material: Base metal Copper content: Conservatively assumed to be 0.10 WT% (actual content = 0.06 WT%) -

RT initial: 40*F NDT RT NOT after 16 EFPY: 1/4T,110*F 3/4T,87'F Curve applicable for heatup rates up to 60*F/hr for the service period.up to 16 EFPY and contains margins of 10*F and 60 psig for possible instrument errors 2800

'2600 2400 LEAK 1l51._ j' r j LIMI' N s/ ) I

', 2200 f

I f f 7 7 ,.

I 2eee l l 1800 s$

~~

l I

}

} {y1600 {} j

l }

1400 /

g r w HEATUP 2 /

m-a CURVE N/, /

{2 / /

E 1888 / f

/

/ s CRITICALITY LIMIT 4

/ BASED ON INSERVICE --~~

N!

- ~ j HYDROSTATIC TEST ---

800 TEMPERATURE (255 F) ---

/g FOR THE SERVICE PERIOCL __

, 600 / UP T 3 % WY 400 200 3

I I 0

100 200 300 ase 400 500 RCS

TEMPERATURE (*F)

(10 F PER DIVISION) . .

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY SEABROOK - UNIT 1 3/4 4-31 2-.

=

. 2

, . _ _ _ . . _ . _ _...._.m-.m.m._.._._.___._m_m.._e._m..m

. _.m .. . m .m u. -. a _,u MATERIAL PROPERTY BASIS Controlling material: Base metal Copper content: Conservatively assumed to be 0.10 Wi% (actual content = 0.06 WTX)

RT NOT initial: 40'F RT NDT after 16 EFPY: 1/4T,110'F 3/4T,87'F .

Curve applicable for cooldown rates up to 100*F/hr for the service' period up to 16 EFPY and contains margins of 10*F and 60 psig for possible instrument errors 2800 -

2600 2400 COOLDOWN LIML15,---s

2200 .I

\ / -

2000 sl

}

- 1800 l 5

i

.B10 1600 l E t )

Q g 1400 I

f w 1200 /

teE '

)

m tee 0 i wg

~

800 A I(

600- N khmSFA 8

400- ae"F#

y

?

200 0

100 230 300 ase 400 RCS TEMPERATURE (*F) .

(10 V PER DIVISION)

FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY SEABROOK - UNIT 1 3/4 4-32

___ __ _____------_*am...-am. e ea _m e a e w e e en , e *ce h*em e.A-- --- - --

. .. N . ---- .

N-N -~ ~ ~ A- - -- ^ ^ - -

REACTOR COOLANT SYSTEM i

PRESSURE / TEMPERATURE LIMITS OVERPRESSJRE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION .

3.4.9.3 At least one of the following Overpressure. Protection Systems shall be OPERA 8LE:

a. Two residual heat removal (RHR) suction relief valves each with a setpoint of less than or equal to 450 psig, or -

b. Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or

c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 1.58 square inches.

APPLICA8ILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 329'F; MODE 5 and MODE 6 with the reactor vessel head on.

ACTION:

a. With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves to OPERA 8LE status within 7 days or depressurize and vent the RCS through at least a 1.58 square-inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 1.58-square-inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

. c. In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and

any corrective action necessary to prevent recurrence,

d. The provisions of Specification 3.0.4 are not applicable.

O SEABROOK - UNIT 1 3/4 4-34 mg wW PhW**"""* 4EI"' -

Wh**F***"***** - ^ '

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

REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE;

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and .

c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when

_ the RHR suction relief valves are being used for cold overpressure protection as follows:

a. For RHR suction relief valve RC-V89

1) By verifying at least once per 31 days that RHR RCS Suction Isolation Valve RC-V88 is open with power to the valve operator removed, and

2) By verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that RC-V87 is open.

. b. For RHR suction relief valve RC-V24

1) By verifying at least once per 31 days that RC-V22 is open with power to the valve operator removed, and

2) By verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that RC-V23 is open.

c. Testing pursuant to Specification 4.0.5.

_. __ 4.4.9.3.3. The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve that is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

SEABROOK - UNIT 1 3/4 4-35

- .- ... = . ._- -.-_..:--..-.-...---.

s .;

g i .

, [

.?

tl

~

I

}

! m '

k

' h as 2500 A

t t 1 O - -

__.y.- '

- - - - l A

K

~ VAIJD TOR THE FIRST 16 ETPY. SETPONT D

_ CONTAINS WARGIN Or 50 F FOR i t

C 4

/

TRANSIENT [FFECTS.

[t 3

~ -_ _.-

i - >

. 4 I

p 2000 } t l

g f {

f ic.s4 esicasm r

/

i i

i r i2.1 . , 3 1 ..a = = r,1> ,.o r i ' t' 5 ; t i

O 1500 / !

b a: t l M / 1 w > [

s a:

l'

& O I:

t' i

m 3 l

m ,

3 1000 ;

Y l .

8- '

/ i

.! ; / i r

I j .

500

- #'/ ;

i 4

5 1 - -.

h

- 200 '

j so 100 150

)

200 250 t j 300, 360

' RCS TEMPER ATURE ( F) i l

u, 1

-4 FIGURE 3.4-4 RCS COLD DVERPRESSURE PROTECTI,0N SETPOINTS i

1 5

?

,i

,...._m_-mum

_.-~__.m_.__..__. ..._ .___ . _ _. m _. ,a REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION -

1 3.4.11 At least one Reactor Coolant System vent path consisting of one vent valve and one block valve powered from emergency busses shall be OPERABLE and closed

at each of the following locations:

a. Reactor vessel head, and

b. Pressurizer steam space.

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

i ACTION:

a. With one of the above Reactor Coolant System vent paths inoperable.

STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve

' actuator of all the vent valves and block valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANOBY 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 both Reactor Coolant System vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuators of all the vent valves and block valves in the inoperable

' vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STAND 8Y 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 .

SURVEILLANCE REQUIREMENTS 4.4.11.1 Each Reactor Coolant System vent path block valve not required to be closed by ACTION a. or b., above, shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel from the control room.

4.4.11.2 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:

.I

a. Verifying all manual isolation valves in each vent path are locked in the open position, r

i "For an OPERABLE vent path using a power operated relief valve (PORV) as the vent path, the PORV block valve is not required to be closed.

SEABROOK - UNIT 1 3/4 4-38 h

Bh 65w

_ RC*-c.m m. ,_.-m< h____ . _ . _ , _ _ _ _

_ , )

I EMERGENCY CORE COOLING SYSTEMS ACCUMULATORS

, SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 5.1. 2 Each reactor coolant system accumulator isolation valve shall be shut i with power removed from the valve operator.

~

APPLICA8ILITY: MODES 4* and 5.

ACTION:

l

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

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

s SU_RVEILLANCE 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. l l

l l

l

~

"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 open as required by Specification 3.5.1.1.a.

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

l

, -- ""h

l EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,y GREATER THAN OR EQUAL TO 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsyst5ms-shall be 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.

A_PP,.ICABILITY:

P 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 STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hou"s.

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

During MODE 3, the discharge paths of both-Safety Injection pumps may be isolated by closing for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform surveillance testing as required by Specification 4.4.6.2.2.

- The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry-into MODE 3 for the centrifugal charging pump and the Safety Injection pumps declared inoperable pursuant to Specification 4.5.3.2 provided the centrifugal changing pump and the Safety Injection pumps are restored to OPERABLE status within at 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 legs exceeding 375*F, whichever comes first.

SEABROOK - UNIT 1 3/4 5-4

- - ~

N 9

____ -__m__~_.._-__.n_.m.____i __m_-__..._.

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350*F SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: ~

- a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that the ,following valves are in the indicated positions with power to the valve operators removed: -

Valve Number Valve Function Valve Position SI-V-3 Accumulator Isolation Open*

SI-V-17 Accumulator Isolation Open* .

SI-V-32 Accumulator Isolation Open*

SI-V-47 Accumulator Isolation Open*

SI-V-114 SI Pump to Cold-Leg Isolation Open RH-V-14 RHR Pump to Cold-Leg Isolation Open RH-V-26 RHR Pump to Cold-Leg Isolation Open RH-V-32 RHR to Hot-Leg Isolation Closed RH-V-70 RHR to Hot-Leg Isolation Closed SI-V-77 SI to Hot-Leg Isolation Closed SI-V-102 SI to Hot-Leg Isolation Closed ^

b. At least once per 31 days by: '

1) Verifying that the ECCS piping is full of water by . venting the ECCS pump casings and accessible discharge piping high points, and-

2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. By a visual inspection which verifies that no loose debris (rags,

. trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

l 1) For all accessible areas of the containment' prior to establish-t ing PRIMARY CONTAINMENT INTEGRITY,-and -

i -

2) Of the areas affected within containment at the completion of each containment entry when PRIMARY CONTAINMENT INTEGRITY is established.

l i

Pressurizer pressure above 1000 psig.

SEABROOK - UNIT 1 3/4 5-5

_ ,-_ _ _ cw : _ m u _- . .u.s m___ m._ _ __ _ .__ m .a . , _ ._ _ _ - ~ . - - - .s 7._._

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,y LESS THAN 350*F ECCS SUBSYSTEMS - T.VG EQUAL TO OR LESS THAN 200*F LIMITING CONDITION FOR OPERATION -

3.5.3.2 All Safety Injection pumps shall be inoperable.

APPLICABILITY: MODE S and MODE 6 with the reactor vessel head on. -

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS .

4.5.3.2 All Safety Infection pumps shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position at least once per 31 days. -

An inoperable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

SEABROOK - UNIT 1 3/4 5-10 l

.-.._-....---.---.--=:---------~~~:---- ~ ~ ~~~ '~- ~~ ~~ ~ " ~ ~ ~ ^

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

l l

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 consid'er-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 l valve open or not locked closed, close and lock close that valve 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 , 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 .

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) i to OPERA 8LE status or isolate the affected penetration (s) so that the measured leakage rate does not exceed the limits of Specifications 4.6.1.7.2 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 the purge supply if the 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 SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . - -

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.

SEABROOK - UNIT 1 3/4 6-12 v -q , ,-m ~_ ,n----'-

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CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS .

4.6.1.7.1 Each 36-inch containment purge supply and exhaust isolation valve shall be verified to be locked closed 'at least once per 31 days. -

4.6.1.7.2 At least once per 6 months on a STAGGERED T85T BASIS each locked closed 36-inch containment purge s~upply and exhaust isolation valve with resiliant seals 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 shall be verified to be sealed closed or open in accordance with Specifi-cation 3.6.1.7.b at least once per 31 days.

e e S

e SEABROOK - UNIT 1 , 3/4 6-13

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LIMITING CONDITION FOR OPERATION .

3.6.4.3 Two independent Containment Structure Recirculation Fan Systems shall be OPERA 8LE.

APPLICABILITY: MODES 1 and 2.

ACTION:

With one Containment Structure Recirculation Fan inoperable, restore the inoperable fan 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 .

SURVEILLANCE REQUIREMENTS 4.6.4.3 Each Containment Structure Recirculation Fan System shall be demon-strated OPERABLE:

a. At least once per 92 days on a STAGGERED TEST BASIS by starting each system from the control room and verifying that the system operates for at least 15 minutes, and

b. At least once per 18 months by verifying a system flow rate of at least 4000 cfm through the hydrogen mixing flow path.

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. 9 O

l SEABROOK - UNIT 1 3/4 6-20 ~

--^.9 M M *'

CONTAINMENT SYSTEMS CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM _

SURVEILLANCE REQUIREMENTS .

4.6.5.1 (Continued)

f. After each complete or partial replacement of a charcoal adsorber 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 N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 2100 cfm 2 10%.

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PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS -

4.7.1.2.2 Auxiliary feedwater flow paths to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days, or after maintenance on an auxiliary feedwater pump that could have-an effect upon pump performance, prior to entering MODE 2 by verifying normal flow to each steam generator from:

a. Each emergency feedwater pump, and

b. The startup feedwater pump via the main feedwater flow path and via the emergency feedwater header.

O 4 #

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SEABROOK - UNIT 1 3/4 7-5 -

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ATMOSPHERIC RELIEF VALVES LIMITING CONDITION FOR OPERATION -

3.7.1.6 At least four atmospheric relief valves and associated manual controls including the safety-related gas supply systems shall be OPERABLE.

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

ACTION:

With less than the required atmospheric relief valves OPERABLE, restore the required atmospheric relief valves to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 . .

SURVEILLANCE REQUIREMENTS

4. 7.1. 6 Each atmospheric relief valve and associated manual controls including the safety related gas supply systems shall be demonstrated OPERABLE:

3

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that the nitrogen accumulator tank is at a pressure greater than or equal to 500 psig.

b. Prior to startup following any refueling shutdown or cold shutdown of 30 days or longer, verify that all valves will open and close fully by operation of manual controls. -1

When steam generators are being used for decay heat removal.

SEABROOK - UNIT 1 3/4 7-10

-m- _ _ _ . . _

PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with: : I

a. A service water pumphouse water level at or above 5'-0", minus 37'-0" Mean Sea Level, USGS datum, and

b. A mechanical draft cooling tower comprised of one cooling tower cell with one OPERABLE fan and a second cell with two OPERABLE fans, and a contained basin water level of equal to or greater than 35.9* feet at a bulk average water temperature of less than or equal to 67.3*F, and

c. A portable tower makeup pump system stored to be OPERABLE for 30 days following a Safe Shutdown Earthquake.

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

ACTION:

a. With the service water pumphouse inoperable, restore-the service water pumphouse to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 folicwing 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . '

b. With the mechanical draft cooling tower inoperable, restore the cooling tower to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least HOT iTANDBY 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 the portable tower makeup pump system inoperable, continue operation and notify the NRC within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in accordance with the procedure of 10 CFR 50.72 of actions or contingencies to ensure an adequate supply of makeup water to the mechanical' draft cooling tower for a minimum of 30 days.

With the cooling tower in operation with valves aligned for tunnel heat treat-ment, the tower basin level shall be maintained at greater than or equal to 34.3 feet.

SEABROOK - UNIT 1 3/4 7-14

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PLANT SYSTEMS ULTIMATE HEAT SINK SURVEILLANCE REQUIREMENTS --

4.7.5 The ultimate heat sink shall be determined OPERABLE: ,'

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by:

1) Verifying the water level in the service water pumphouse to be at or above 5'-0", minus 37'-0" Mean Sea Level, and

2) Verifying the water in the mechanical draft cooling tower basin to be greater than or equal to a level of 35.9 feet.

b. At least once per week by verifying that the water in the mechanical draft cooling tower basin to be at a bulk average temperature of

67.3*F.

c. At least once per 31 days by:

1) Starting from the control room each UHS cooling tower fan that is required to be OPERABLE and operating each of those fans for at least 15 minutes, and

2) Verifying that the portable tower makeup pump system is stored in.its design operational readiness state.

d. At least once per 18 months by verifying automatic actuation of each cool.ing tower fan on a Tower Actuation test signal.

SEABROOK - UNIT 1 3/4 7-15 i

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PLANT SYSTEMS 3/4.7.6 CONTROL 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 Control 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 SHUTDOWN 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 inoperable 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 demcnstrated 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 months or after any significant modification to i the Control Room Area Ventilation Systems by verifying a system . flow rate of 25,700 cfm

10% through the air conditioner unit (3A and 3B) and a flow rate of at least 1200 cfm 2 10% makeup from each intake to the emergency filtration unit with a discharge of 2200 cfm i 10% from the filtration unit. , ,

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SEABROOK - UNIT 1 3/p7-16

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PLANT SYSTEMS CONTROL ROOM AREA VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS 4.7.6 (Continued) ,

c. At least once per 18 months by:

1) Verifying that on a high radiation signal from the control room makeup air intake, the subsystem automatically switches to the emergency recirculation mode of operation and the isolation dampers close within 5 seconds.

, 2) Verifying that on an S signal the emergency filtration fans start.

3) Verifying that the system maintains the control room area at a' positive pressure of greater than or equal to a pressurization 1/8-inch Water Gauge relative to adjacent areas during system operation at less than or equal to a pressurization flow of 1200 cfm i 101 SEABROOK - UNIT 1 -

3/4 7-17

ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS -

4.8.1.1.2 (Continued) ,

13) Verifying that the following diesel generator lockout features prevent diesel generator starting: .

a) Barring device engaged, or b) Differential lockout relay.

14) Simulating a Tower Actuation (TA) signal while the diesel generator is loaded with the permanently connected loads and auto-connected emergency (accident) loads, and verifying that the service water pump automatically trips, and that the cool-ing tower pump and fan (s) automatically start. After energiza-tion the steady state voltage and frequency of the emergency buses shall be maintained-at 4160 1 420 volts and 60 1 1.2 Hz;

~

and -

15) While diesel generator IA is loaded with the permanently connected loads and auto-connected emergency (accident) loads, manually connect the 1500 hp startup feedwater pump to 4160-volt bus E5. After energization the steady-state voltage and ;

frequency of the emergency bus shall be maintained at 4160 1 420 volts and 60

1.2 Hz.

g. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel <

generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 514 rpm in less than or equal to 10 seconds; and

h. At least once per 10 years by:

1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite

solution, or equivalent, and 2). Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure. -

SEABROOK - UNIT 1 3/4 8-8

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ELECTRICAL POWER SYSTEMS 3/4.8.2 0.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION -

3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE and energized:

a. Train A '

1) 125-volt Battery Banks 1A and IC,

2) One full-capacity battery charger on Bus #11A, and

3) One full-capacity battery charger on Bus #11C.

b. Train B

1) 125-volt Battery Banks 18 and 10,

2) One full-capacity battery charger on Bus #118, and

3) One full-capacity battery charger on Bus #11D.

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

ACTION:

a. With one of the required battery banks in one train inoperable, close the bus tie to connect the remaining operable battery bank to the D.C.

bus supplied by the inoperable battery bank within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; restore the inoperable battery bank to GPERABLE 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 and in COLD SHUTDOWN

l. within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one of the full-capacity chargers inoperable, restore the inoper-able charger to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 .

SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE: l

a. At least once per 7 days by verifying that:

i

1) The parameters in Table 4.8-2 meet the Category A limits, and

2) The total battery terminal voltage is greater than or equal to 128 volts on float charge. - -

b. At least once per 92 days and within 7 days after a batt'ery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:

No more than one battery at a time may be taken out of service for more than 30 days.

SEABROOK - UNIT 1 3/4 8-12

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,7 Er.ECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION -

3.8.3.1 (Continued)

1. Train A,125-volt D.C. Busses consisting of: -

1) 125-volt D.C. Bus #11A energized from Battery Bank 1A* or IC*, and

2) 125-volt D.C. Bus #11C energized from Battery Bank IC* or 1A*.

J. Train B,125-volt D.C. Busses consisting of:

1) 125-volt D.C. Bus #11B energized from Battery Bank 1B* or ID*, and

2) 125-volt D.C. Bus #110 energized from Battery Bank 1D* or 1B*.

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

ACTION: -

a. .With one of the required trains of A.C. emergency busses not fully energized, reenergize the train within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-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 . I

b. With ont A.C. Vital. panel either not energized from its associated inverter, or with the inverter not connected to its associated D.C.

bus: (1) reenergize the A.C. vital panel within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 ; and (2) reenergize the A.C. vital

. panel from its associated inverter connected to its associated D.C.

bus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 .

i c. With one D.C. bus not energized from its associated battery bank, reenergize the D.C. bus from its associated battery bank or close

, the bus tie to the alternate OPERABLE battery of the same train within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 .

l SURVEILLANCE REQUIREMENTS

~

l 4.8.3.1 The specified busses and panels shall be determined energized in the t required manner at least once per 7 days by verifying correct breaker alignment i and indicated voltage on the busses. l l

No more than one Battery Bank (IA, 1B, 1C, or 1D) at a time may be taken out of service for more than 30 days.

SEABROOK - UNIT 1 3/4 8-17

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ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION _

3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner,

a. One train of A.C. emergency busses consisting of the 4160; volt and the 480-volt A.C. emergency busses listed in 3.8.3.la. and b.

(excluding 480-volt Emergency Bus #E64);

b. Two of the four 120-volt A.C. Vital Panels 1A, 18, 1C, and 10

. energized from their associated inverters connected to their respective D.C. busses; ,

c. One of the two 120-volt A.C. Vital Panels IE or IF energized from l its associated inverter connected to ,the respective D.C. bus; and

d. Two 125-volt D.C. busses (in the same train) energized from their associated battery banks. .

APPLICABILITY MODES 5 and 6.

ACTION:

With any of the above required electrical busses and panels not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses and panels in the specified manner as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the RCS through at least a 1.58-square-inch vent.

SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified besses and panels shall be determined energized in the reqtired manner at least once per 7 days by verifying correct breaker alignment and fr.aicated voltage on the busses.

I I

SEABROOK - UNIT 1 3/4 8-18 .

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REFUELING OPERATIONS

, 3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine and auxiliary hoist shall be used for inovement of drive rods or fuel assemblies and shall be OPERABLE with: ~

a. The refueling machine used for movement of fuel assemblies having:

1) A minimum capacity of 4000 pounds, and

2) An overload cutoff limit less than or equal to 3900 pounds.

b. The auxiliary hoist used for latching and unlatching drive rods having:

1) ,A minimum capacity of 2100 pounds, and

~

2) A load indicator which shall be used to prevent lifting loads in excess of 1000 pounds.

APPLICABILITY: Ouring movement of drive rods or fuel assemblies within the reactor vessel.

ACTION:

With the requirements for refueling machine and/or hoist OPERABILITY not satis-fied, suspend use of any inoperable refueling machine and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor vessel.

SURVEILLANCE REQUIREMENTS 4.9.6.1 The refueling machine used for movement of fuel assemblies within i the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to i the start of such operations by performing a load test of at least 4000 pounds and demonstrating an automatic load cutoff when the refueling machine load exceeds 3900 pounds.

4.9.6.2 The auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a load test of at least 2100 pounds. .

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4.9.12b (Continued)

~

1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidanco in !

Regulatory Positions C.5.a, C.5.c, and C.S.d of Regulatory !

Guide 1.52, Revision 2, March 1978,* and the system flow rate 1 is 17,000 cfm 2 10%;

2) Verifying, within 31 days after removal, that a labo*atory analysis of a representative carbon sample obtained in accor-dance wtth Regulatory Position C.6.b of Regulatory Guide 1.52,-

Revision 2, March 1978,* meets the laboratory testing criteria ,

of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, j March 1978, by showing a methyl iodide penetration of less than i 1.0% when tested at a temperature of 20'C and at a relative hu- l midity of 95% in accordance with ASTM-D-3803; and

3) Verifying a system flow rate of 17,000 cfm i 12% 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 representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March I978,*

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,* by showing a methyl iodide penetration of less than 1.0% when tested at a tem-perature of 30*C and at a relative humidity of 95% in accordance with ASTM-0-3803.

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 inches Water Gauge while operating the system at a flow rate of 17,000 cfm i 10%,

.2) Verifying that the system maintainsithe spent fuel storage pool area at a negative pressure of greater than or equal to 1/4 inch Water Gauge relative to the outside atmosphere dtering system operation, -

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

a l SEA 8 ROOK - UNIT 1 3/4 9-14

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RADIOACTIVE EFFLUENTS GASEOUS EFFLUENTS GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION .

3.11.2.4 Yae VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS RADWASTE TREATMENT SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days dua to gaseous effluent releases, from each unit, to areas at and be-yond the SITE BOUNDARY'(see Figure 5.1-1) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 erad to air from beta radiation, or

c. 0.3 area to any organ of a MEMBER OF THE PUBLIC.

APPLICABILITY: At all times.

ACTION:

a. With radioactive gaseous waste being discharged without treatment.

and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that includes the following information:

1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

, 3. Summary description of action (s) taken to prevent a recurrence.

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

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the 00CM when Gaseous Radwaste Treatment Systems are not being fully utilized. , ,

4.11.2.4.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and' GASEOUS RADWASTE TREATMENT SYSTEM shall be considered OPERABLE by meeting Spect fica-tions 3.11.2.1 and 3.11.2.2 or 3.11.2.3.

4 SEABROOK - UNIT 1 3/4 11-8

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RADI0 ACTIVE 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 WASTE GAS HOLDUP 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 GASEOUS RADWASTE TREAT-MENT SYSTEM shall be determined to be within the above limit by continuously monitoring the waste gases in the GASE0US RADWASTE TREATMENT SYSTEM with the hydrogen or oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.10.

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i RADI0 ACTIVE EFFLUENTS GASEOUS EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION '

3.11.2.6 The quantity of radioactivity contained in the hydrogen. surge tank shall be limited to less than or equal to 198,000 Curies of noble gases (con-sidered as Xenon-133 equivalent). -

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in the hydrogen surge tank exceeding the above limit, immediately suspend all additions of

radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.8.1.4.

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

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage

~

tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

SEA 8 ROOK - UNIT 1 3/4 11-10

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RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTES a

LIMITING CONDITION FOR OPERATION 3.11.3 Radioactive wastes shall be SOLIDIFIED or dewatered in accofdance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during site. transit, and disposal site requirements when received at the disposal APPLICABILITY: At all times. .

~

ACTION:

a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures, and/or the Solid Waste System as necessary to prevent recurrence.

b. With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to prevent recurrence. _

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i SURVEILLANCE REQUIREMENTS 4.11.3 For cement SOLIDIFICATION of at least one representative test speci-men from at least every tenth batch of each type of wet radioactive wastes (e.g. , filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM,

' and a subsequent test veriffes SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM;

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION.

The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.12, to assure SOLIDIFICATION of subsequent batches of waste; and SEABROOK - UNIT 1 3/4 11-11 ;

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i POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS i The limits on the DNB-related parameters assure that each of is maintained within the normal steady-state envelope of operation,the assumedpara.neters in the transient and accident analyses. The Ilmits are consistent with the initial FSAR assumptions and have been analytica11v demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits <

of 594.3*F for T,yg and 2205 psig for pressurizer are not exceeded.

The measurement error of 2.1% for RCS total flow rate is based upon per-

, forming a precision heat balance and using the result to normaliz.t the RCS flow rate indicators. Potential fouling of the (gedwater venturi which afght not be detected could bias the result from the precision heat balance in a noncend servative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is applied. Any fouling which might bias the RCS ficw rate measurement greater than 0.1% can be detected by monitoring and trending vari-ous plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of these parameters through instrument

, readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient cperation.

'i The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outsioe the specified limit.

l SEABROOK - UNIT 1 B 3/4 2-4

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s REACTOR COOLANT SYSTEM BASES REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) ,

Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can'be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage. -

The total steam generator tube leakage limit of I gpm for all steam generators not isolated from the RCS ensures that the dosage contribution ;

from the tube leakage will be limited to a small fraction of 10 CFR Part 100 l

' dose guideline values in the event of either a steam generator tube rupture !

or steam line break. The 1 gpm limit is consistent with the assumptions used l in the analysis of these accidents. The 500 gpd leakage limit per stear '

generator ensures that steam generator tube integrity is maintained in the -

event of a main steam line rupture or under LOCA conditions.

The 10 gpm IDENTIFIED LEAXAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detaction Systems.

The CONTROLLED LEAXAGE limitation restricts cperation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.

This limitation ensures that in the event of a LOCA the rafety injection flow will not be less than assured in the safety analyses,.

The specified allowed leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected '

for a substantial length of time, verification of valve integrity is required.

Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves thould be tasted periodically to ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersysten LOCA. Leakage from the ECS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

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FIGUR2 8 3/4,4 1 FAST NEUTRON FLUENCE (E>1Mey) AS A FUNCTION OF FULL POWER SERVICE LIFE SEABROOK - UNIT 1 B 3/4 4-9 OMN T_N N Y T ' D N*' ,Y*T'*T"%"- 7 N O " .1 e# 3 h+* f,_ ' J N4g 4%_,'

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FAST NEUTRON FLUENCE (N/CM , E > 1 MeV) l FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT NDT FOR REACTOR VESSELS EXPOSED TO 550*F SEABROOK - UNIT 1 8 3/4 4-10

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Component Spec. No.

Code No. 'f%J $ Q f*FJ (ft-lb) (ft-Lb) }

s Closure Head Dome R1809-1 A533B,CL.1 0.15 0.011 -40 10 80.5 -

Closure Head Torus R1810-1 A533B,CL.1 0.08 0.012 -50 0 104 -

Closure Head. Flange R1802-1 A508,CL.2 -

0.013 10 10 105.5 -

Vessel Flange (

R1801-1 A508,CL.2 -

0.012 20 30 91 -

t Inlet Nozzle R1804-1 A508,CL.2 0.10 0.011 0 0 Inlet Nozzle 125 -

f R1804-2 A508,CL.2 0.09 0.010 -20 -20. 125 -

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l' Outlet Nozzle R1805-4 A508,CL.2 -

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J t Nozzle Shell R1807-1 A533B,CL.1 0.08 0.011 -30 30 66 -

t U Nozzle Shell R1807-2 A533B,CL.1 0.09 0.012 -40 30 66.5 -

c Nozzle Shell R1807-3 A533B,CL.1 0.06 0.010 -20 10 107 -

Inter. Shell R1806-1 A533B,CL.1 0.04 0.012 -30 40 82 139.5 h Inter. Shell R1806-2 A5338,CL.1 0.05 0.007 -30 0 102 143.5 Inter. Shell R1806-3 A5338,CL.1 0.07 0.007 -40 10 115 138 Lower Shell R1808-1 A533B,CL.1 0.05 0.005 -30 40 78 120.5 ,

g Lower:Shell R1808-2 A5338,CL.1 0.05 0.007 -20 10 77 127 l-Lower Shell R1808-3 A533B,CL.1 0.06 0.007 -20 40 78 130.5 Bottom Head Torus R1811-1~ A5338,CL.1 0.15 0.010 -50 0 94.5 -

Bottom Head Dome R1812-1 A533B,CL.1 0.09 0.009 -30 ~

0 97.5 -

Inter. & Lower Shell -

Long Weld Seams G1.72 Sub Arc Weld 0.07 ' O.008 -50 -50 200 -

Inter. & Lower Shell ,

Girth Weld Seam G1.72 Sub Arc Weld 0.07 0.008 -50 -50 200 -

NMWD - Normal to Major Working Direction

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PLANT SYSTEMS BASES .

3/4.7.1 TURBINE CYCLE (Continued)

~

3/4.7.1.2 AUXILIARY FEE 0 WATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that-the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of'a total loss-of-offsite power. _

The electric notor-driven emergency feedwater pump is capable of deliver-ing a total feedwater flow of 650 gpm at a pressure of 1185 psig to the en-trance of the steam generators. The steam-driven emergency feedwater pump is capable of delivering a total feedwater flow of 650 gpm at a pressure of.

. 1221 psig to the entrance of the steam generators. The startup feedwater pump serves as the third auxiliary feedwater pump and can be manually aligned to be powered from an emergency bus (Bus 5). The startup feedwater pump is capable

of taking suction on the dedicated emergency feedwater volume of water in the condensate storage tank and delivering a total feedwater flow of in excess of 650 gpm at a pressure of 1221 psia to the entrance of the steam generator via either the main feedwater header or with manual ~ alignment to the emergency feedwater flow path. This capacity is sufficient to ensure that adequate feed-water flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water vol-use ensures that sufficient water is available to cool the RCS to a temperature of 350'F. The OPERABILITY of the concrete enclosure ensures this availability of water following rupture of the condensate storage tank by a tornado generated missile. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line ruptLie.

This dose also includes the effects of a coincident 1 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a 4 team line '

rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses. 1 SEABROOK - UNIT 1 8 3/4 7-2

_ n__r_ _

PLANT SYSTEMS BASES 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensores that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RTNOT of 60*F and are sufficient to prevent brittle fracture l .

3/4.7.3 PRIMARY COMP 0NENT COOLING WATER SYSTEM The OPERABILITY of the Primary Component Cooling Water System ensures that sufficient cooling capacity is available for continued o related equipment during normal and accident conditions.peration of safety-The redundant cooling capacity of this system, assuming a single failure, is. consistent with the .-

assumptions used in the safety analyses.

3/4.7.4' SERVICE WATER SYSTEM The Service Water System consists of two independent loops, each of which can operate with either a service water pump train or a cooling tower pump train. The OPERABILITY of the Service Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equip-ment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is' consistent with the assumptions used in the safety analyses, which also assumes loss of either the cooling tower or ocean cooling.

~3/4.7.5 ULTIMATE HEAT SINK The limitations on service water pumphouse level, and the OPERABILITY requirements for the mechanical draft cooling tower and the portable tower makeup pump system, ensure that sufficient cooling capacity is available to either: (1) provide normal cooldown of the facility or (2) mitigate the effects of accident conditions within acceptable limits. This cooling capabil-ity is provided by the Atlantic Ocean except during loss of ocean tunnel water flow, when the cooling capability is provided by the mechanical draft cooling tower with tower makeup using portable pumps.

The limitations on' minimum water level and the requirements for mechanical draft cooling tower OPERABILITY are based on providing a 30-day cooling water supply to safety related equipment without exceeding its design basis tempera-ture and is consistent with the recommendations of Regulatory Guide 1.27,

" Ultimate Heat Sink for Nuclear Plants," March 1974.

3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM -

The OPERABILITY of the Control Room Area Ventilation System ensures that:

(1) the allowable temperature for co..tinuous-duty rating for the equipment and SEABROOK - UNIT 1 B 3/4 7-3 1

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RADIOACTIVE EFFLUENTS BASES GASEOUS EFFLUENTS 3/4.11.2.5 EXPLOSIVE GAS MIXTURE FOR THE WASTE GAS HOLDUP SYSTEM , '

This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the GASEOUS RADWASTE SYSTEM is main-tained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides a:surance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.6 GAS STORAGE TANK The tank included in this specification is that tank for which the quan-

-tity of radioactivity contained is not limited directly or indirectly by another Technical Specification. . Restricting the quantity of radioactivity contained in the gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a ME.'IBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem. This is consistent with Standard Review Plan Section 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981.

3/4.11.3 SOLID RADI0 ACTIVE WASTES This specification implements the requirements of 10 CFR 50.3Ea and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGPAM may include, but are not limited to, waste type, waste pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.

3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The

, specification requires the preparation and submittal of a Special Report when-l ever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 areas to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

i For sites containing up to four reactors, it is highly unlikely that the resul-tant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the doser design objec- -

tives of Appendix I, and if direct radiation doses from the raits (including outside storage tanks, etc.) are kept small. The Special Reprt will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190' limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF SEABROOK - UNIT 1 B 3/4 11-5

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RADI0 ACTIVE EFFLUENTS

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BASES 3/4.11.4 TOTAL DOSE (Continued)

THE PUBLIC from other uranium fuel cycle sources is negligible, with-the excep-tion that dose contributicas from other nuclear fuel. cycle facilities at the i same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requireme <ts of 40 CFR Part 190, the Special Report with a request for a variance (provided the release condi- ,

tions resulting in violation of 40 CFR Part 190 have not already been corrected), i in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is con-sidered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for

. dose Ifmitation of 10 CFR Part 20, as addressed in Specifications 3.11.1.1 and i 3.11.2.1. An ir.dividual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part oY the nuclear fuel cycle.

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5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shnwn in Figure 5.1-1. -

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LOW POPULATION ZONE 5.1. 2 The Low Population Zone shall be as shown in Figure 5.1-2.

MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIV LIQUID EFFLUENTS

.5.1.3 Information regarding radioactive gaseous and-liquid effluents, which

' will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS

-OF THE PUBLIC, shall be shown in Figures 5.1-1 and 5.1-3, respectively.

The definition of UNRESTRICTED AREA used in implementing these Technicc1 Spect-fications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Limiting Conditions for Operation to keep levels '

of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a.

5.2 CONTAINMENT CONFIGURATION i

5.2.1 The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 140 feet.

b. Nominsi inside height = 219 feet.

c.

Minimum thickness of concrete walls = 4 feet 6 inches.

d.

Minimum thickness of concrete dome = 3 feet 6 inches.

I i

e. Minimum thickness of concrete floor pad = 10 feet.

f. Nominal thickness of steel liner = 1/4, 3/8, and 1/2 inch for ,

the floor, wall, and dome, respectively. ,, .

I

g. Het free volume = 2.704 X 108 cubic feet.

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FIGURE 6.2-2 i . , t STATION ORGANIZATION i

                                                                                                                                               .                                         i i                                                                                                                                                                                    t, i                                                                                                                                                                                       f-I                                                                                                                                                                                      I

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ADMINISTRATIVE CONTROLS l MEETING FREQUENCY 6.4.2.5 The NSARC'shall' meet at least once per calendar quarter during the initial year of unit operation following fuel loading and thereafter at least once per 6 months t 6 weeks. QUORUM ~ l 6.4.2.6 The quorum of the NSARC necessary for the performance of the NSARC i " review and audit functions of these Technical Specifications shall consist of ! the Chairman or Vice-Chairman and at least four NSARC members including alter- ' nates.'No of operation morethe unit. than a minority of the quorum shall have line responsibility for ' The Vice Chairman, or his designated alternate, can participate as an NSARC member when the Chairman is in attendance. , REVIEW i

                                                                                                                                                 .        i 6.4.2.7 The NSARC shall be responsib1'e for the review of:

a. The safety evaluations for: (1) changes to procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an unraviewed safety question; _

b. Proposed changes to procedures, equipment, or systems that involve an unreviewed safety question as defined in 10 CFR 50.59;

c. Proposed tests or experiments that involve an unreviewed safety ques-l tion as defined in 10 CFR 50.59; i l

i

d. Proposed changes to Technical Specifications or this Operating l License; I

e. Violations of Codes, regulations, orders, Technical Specifications, '

license requirements, or of internal procedures or instructions having nuclear safety significance;

f. Significant operating abnormalities or deviations from normal and expected performance of station equipment that affect nuclear safety;

g. All REPORTA8LE EVENTS;

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and

1. Reports and meeting minutes of the 50RC. * * '

AUDITS 6.4.2.8 Audits of station activities shall be performed under the cognizance of the NSARC. The audits shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the specified interval SEABROOK - UNIT 1 6-9 g

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3 "- " L ADMINISTRATIVE CONTROLS 4 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 ca.lculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977. . t 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 i 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 i effluent monitoring instrumentation was not corrected within the time specified l 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 t

6. 8.1. 5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Resource Management, i

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 15tti of each month following the calendar month covered by the report. i RADIAL PEAKING FACTOR LIMIT REPORT

6. 8.1. 6 The F xy limits for RATED THERMAL POWER (FRTP) shall be provided to x

the NRC Regional Administrator with a copy to Director of Nuclear Reactor Regulation, Attention: Chief, Reactor Systems Branch, OPL-A, U.S. Nuclear Regulatory Commission, Washington, D. C. 20555, for all core planes containing Bank "D" control rods and all unrodded core planes and the plot of predicted (F PRel) vs Axial Core Height with the limit envelope at least E0,* days prior to each cycle initial e-iticality unless otherwise approved by the Commission by letter. In additi~on, in the event that the limit should 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 Commission by letter. Any informa-tion needed to support F RTP will be by request from the NRC and need not be x included in this report. SEA 8R00K - UNIT 1 6-18 y__ ,- - "y -ry- = e *er ? wa- + + y ew-e==e- _ e.**;; - = = - - -

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ML20211D512

Contents

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Dear Mr. Harrison:

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Dear Mr. Harrison:

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ML20211D512

Text

Dear Mr. Harrison:

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Enclosure:

Dear Mr. Harrison:

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Dear Mr. Harrison:

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