Text

Proposed Changes to Tech Specs to Implement Guidance of Generic Ltr 88-16,relocating Listed Tech Specs to Core Operating Limits Rept
	ML20082P554
Person / Time
Site:	Seabrook
Issue date:	09/04/1991
From:	; PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	;
Shared Package
ML20082P538	List: ML20082P535; ML20082P542; ML20082P554;
References
	GL-88-16, NUDOCS 9109100378
	Download: ML20082P554 (92)
	v • d • e

, . - ___ . - _ _ _ . _ _ _ __

1

!NDEX 1.0= OEFINITIONS l QE?

Se5 SECTION PAGE 1.1 ACTI0N................'........................................ 1-1 1.2 ACTULTION LOGIC TEST.......................................... 1-1 1.3 ANALLs CHANNEL ODERATIONAL TEST.............,................. 1-1 1.4 AXIAL FLUX DIFFERENCE......................................... 1-1 1.5 CHANNEL CALIBRATION........................................... 1-1 1.6 CHANNEL CHECK................................................. 1-1

1. 7 CONTAINMENT INTEGRITY......................................... 1-2 1.B CONTROLLED LEAKAGE............................................ 1-2 1.9 CORE ALTERATION............................................... 1-2

'1.10v DOSE EQUIVALENT I-131........................................ 1-2 1.lfa E - AVERAGE DISINTEGRATION ENERGY............................ 1 ,2' 3 1.120 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.isN FREQUENCY N0TATION........................................... 1-3 1.14v GASE0US RA0 WASTE TREATMENT SYSTEM............................ 1-3 1.1S* IDENTIFIED LEAKAGE........................................... 1-3 1.16n MASTER RELAY TEST............................................ 1-3 1.172r MEMB E R ( 5) 0 F TH E PUB L I C . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-J' t' 1.18310FFSITE DOSE CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.lsAe0PERABLE - OPERABILITY....................................... 1-4 1.?du 0PERATIONAL MODE - M00E...................................... 1-4 1.2TS2 PHYSICS TESTS................................................ 1-4 1.2f13 PRESSURE BOUNDARY LEAKAGE.................................... 1-4

1. 2524 P ROC E SS CONTRO L P R0G R'AM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- r 6F

{fh 1.24 2 PURGE - PURGING.............................................. 1-A 3I r/ 1,254 QUADRANT POWER TILT RATI0.................................... 1- 5 .

1. 2 527 RAT E D TH E RMA L P0W E R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.272fREACTOR TRIP SYSTEM RESPONSE TIME............................ 1-5 1.26ASREPORTABLE EVENT.... ........................................

l 1-5 1.255 CONTAINMENT ENCLOSURE BUILDING INTEGRITY.....................

8 1-5 1.30p SHUTDOWN MARGIN.........................................'. ... F 6 1.af3SSITE B0VNOARY............................................ ... 1-5 6 i

1. 32 33 S LAVE R E LAY T E ST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 L 1.15PSOLIDIFICATION.............................................., 1-6 1.341sSOURCE CHECK................................................. 1-6 l 1.-3FswSTAGGERED TEST BASIS.... .................................... 1-6 L 1. 3637 TH E RMA L P 0WE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 l

1. FTW TRIP ACTUATING DEVICE OPERATIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . 1-6

! 1. 36 % UNI D ENT I F I ED L E A KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-F 7 L 1.}Ye UNRESTRICTED AREA............................................ 1 ,6' 7 ,

l 1. 46WVENTILATION EXHAUST TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . 1-7 L

i 1.AITIVENTING...................................................... 1-7 l

TABLE 1.1 FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 OPERATIONAL M00ES....................................... 1-8

f. / c C oR6 CPER A DN G L/sutTS RCPO W . .

l-3

.cf" .

%n l- SEABROOK - UNIT 1 i g jj 9109100378 910904

'PDR ADOCK 05000443 P PDR

INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown............................... .... 3/4 1-7 Flow Paths - Operating.............................. .... 3/4 1-8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - Operating............................... 3/4 1-10 Borated Water Sources - Shutdown......................... 3/4 1-11 Borated Water Sources - Operating........................ 3/4 1-12 Isolation of Unborated Water Sources - Shutdown.......... 3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height................. ............... ........... 3/4 1-15 TABLE 3.1-1 ACCIDENT-ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00................... 3/4 1-17 Position Indication Systems - Operating.................. 3/4 1-18 Position Indication System - Shutdown.................... 3/4 1-19 Rod Orop Time............................................ 3/4 1-20 Shutdown Rod Insertion Limit............................. 3/4 1-21

.;. Control Rod Insertion Limits. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-22 g

I! CURE 3.1-1 R00 EANK INSERTION LI"ITS VERSUS TllER"AL POWER FOUR-LOOP OPERA!!OP........ . ..... ... . ............ . g /43 1 23 g

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 I!OURE 3.2-1 AXIAL FL"X OIFFERENCE LIMITS AS A TUN & TION Or 4/dc RATED TliER"AL POWER...... .......................... .... p 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)..................... 3/4 2-4 9

r!0"RE 3.2-2 "(Z) - NORMALIZE 0 F (2)q AS ^ FUNCTION OF CORE HEIC"T.

3/4 2-5 26 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR................. 3/4 2-8 -

3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-9 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-10 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTR 6 MENTATION....... ........... 3/4 3-2 SEABROOK - UNIT 1 iii g] 9l Q

5

- - - - --~ .- . .- .

INDEX m 6.0 ADMINISTRATIVE CONTROLS ec:

a; SEC* ION PAGE 6.4 REVIEW AND AUDIT............................................ 6-6 6.4.1 STATION OPERATION REVIEW COMMITTer (50RC)

Function................................................ . 6-6 Composition................ ............................., 6-6 Alternates................................................ 6-6 Meeting Frequency......................................... 66 Quorum................................ ................... 6-6 Re s p o ns i b i l i ti e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6

. Records............... ................................... 6-8 6.4.2 NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC) function. .......................................... .... 6-8 Composition.... ... ..... ............................., 6-8 Alternates................................................ 6-8 Consultants............................................... 6-8 Meeting frequency......................................... 6-9 Quorum.................................................... 6-9 ,

Review............................................ ....... 6-9

.., Audits,................................................... 6-9

,s.@ Records................................................... 6-11 w

6.5 REPORTABLE EVENT ACTI0N..................................... 6-11 6.6 SAFETY LIMIT VIOLATION..................................,... 6-11 6.7 PROCEDURES AND PROGRAMS....... ............................. 6-12 6.8 REPORTING REQUIREMENTS 6.8.1 -ROUTINE REPORTS............ .............................. 6-14 Startup Report............................................ 6-14 Annual Reports..... ...................................... 6-15 Annual Rad.iological Environmental Operating Report........ 6-15 Semiannual Radioactive Effluent Release Report............ 6-17 .

Monthly Operating Reports................................. 6-18

- Radial Pc0 king F00t0 F Lini t R0p0rt. . . . . . . . . . . . . . . . . . . . . . . . 6-18 Cow CwArma J omrs Rcentr 6.8.2 SPECIAL REP 0RTS........................................... 6-19 6.9 RECORD RETENTION............................................ 6-19

6.10 RADIATION PROTECTION PR0 GRAM............................... 6-20

( "

.p l ?

SEABROOK - UNIT 1 xiv g8{

J3

g. g ,

DEF8N8T20NS

~-

' CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditiens are either:

1) Capable of being closed by an OPERABLE containment-automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions,

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification

3. 6.1., 3 ,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

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

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant-pump seals.

CORE ALTERATION 1.9- CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

- i DOSE EOUIVALENT I-131

//

1.M DOSE EQUIVALENT I-131 shall- be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133 - I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in NRC-Regulatory Guide 1.109, Revision 1, " Calculation of Annual Dosts to Man from i Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I.",

E - AVERAGE DISINTEGRATION ENERGY

/L 1.M E shall_ be the_ average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample with half-lives greater than 10 minutes.

SEABROOK - UNIT 1 1-2 $

N_

W

.. , , . _ . - .. -.- n. . . - - . . . . - . - . . - _ . _ . . . . - . - . . .

1

...- t-l l

INSERT ~ A FOR PROPOSED TECHNICAL SPECIFICATION CHANGES CORE OPER ATING LIMITS REPORT 1,10 limits for the :

The CORE current OPERATING operating LIMITS reload cycle. REPORT The p.vi..les cycle specific corecore operatingl imits shall be

-operating !

determined for each reload cycle in accordance with Specification' 6.8.1.6. Plant operation within these operating limits is addressed in individual specifications.

9

DEFINIT 10NS ENGINEERED SAFETY FEATURES RESPONSE TIME 1.1 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the manitored 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 loadino delays where applicable.

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

GASEOUS RADWASTE TREATMENT SYSTEM

/5 1 }4' 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.

IDENTIFIED LEAKAGE

/b 1 34 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE} into closed systems, such as pumo 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 botn 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

/7 1.}6 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include.

a continuity check of each associated slave relay.

MEMBER (5)0FTHEPUBl3

/r 1.Jf MEMBER (S) 0F THE PUBLIC shall include all persons who tre not occupa-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.

SEABROOK - UNIT 1 1-3 i!

e DEF2N!T10NS ;

0, 0FFSITE DOSE CALCULATION MANUAL .

7 /9 '

1. M -The OFFSITE OOSE CALCULATION MANUAL (0DCM) shall contain in Part A the radiological effluent sampling and analysis program and radiological environ-mental _ monitoring program. Part 8 of the 00CM shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive i gaseous and liquid effluents, in the cc' wlation of gaseous and liquid- '

effluent monitoring Alarm / Trip SetpointL and in the conduct of the Environ- i mental Radiological Monitoring Program.

l OPERABLE - OPERABILITY I 2.o ,

1.16 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 i 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 - MODE H l

..1.36 An OPERATIONAL MODE (i.e.,-MODE) shall correspond to any one inclusivo :

combination of core reactivity condition, power level, and average reactor ,

coolant-temperature specified in Table 1.2. :

. PHYSICS TESTS +

. *.Y: ) 2-1,74 PHYSICS TESTS shall be those tests performed to measure the fundamental

nuclear characteristics of the reactor core and related instrumentation: 5 (1) described in Chapter 14.0 of the FSAR, (2) authorized under the f provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE :

1. PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component !

body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM i 2y i 1.ps The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, '

sampling, analyses, tests, and des,erminations to be made to ensure that l processing and packaging of solid radioactive wastes based on demonstrated I

-processing of-actual or simulated wet solid wastes will be accomplished in i such a way as to assure compliance with 10 CFR' Parts 20, 61, and 71 and f Federal and State Regulations, burial ground requirements, and other require-- ;

ments governing the disposal of radioactive waste.

]

PURGE - PURGING

5 1.J4 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.

SEABROOK - UNIT 1 1-4 9I [

cS2b -Q j

. 4 DEFINITf0NS

,7f,; QUADRANT POWER TILT RATIO

py g4 .

-1.)6 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average uf the epper 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 outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER

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

REACTOR TRIP SYSTEM RESPONSE TIME U

1. g The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REDORTABLE EVENT 7-1

1. E8 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

CONTAINMENT ENCLOSURE BUILDING INTEGRITY 1.hCONTAINMENTENCLOSUREBUILDINGINTEGRITYshallexistwhen

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 Sys. tem is OPERABLE, and

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

-SHUTDOWN MARGIN

1. SHUT 00WN 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 SL 1.Jf The SITE B0UNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SEABROOK - UNIT 1 1-5 9/

cpB I g

~_. _ ._ ._ __ - -

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

DEFINITIONS __ _

I SLAVE RELAY TEST

$gl.e 1.Jt3.) A SLAVE RELAY TEST shall be the energization of each 41 ave relay and verification of CPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. !

SOLIOIFICATION 3 M i 1.)1 SOLIDIFICATION shall be the conversion of wet wastes into'a form that ,

meets shipping and burial giound requirements.

f SOURCECHEE l 35 t 1.)4 A SOURCE CHECK shall be the qualitative assessement of channel response r when the " annel sens.r is exposed to a source of increased radioactivity. ,

STAGGERED TEST BASIS I 30 1 175 A STAGGERED TEST BASIS shall consist of: l

a. A test schedule for n systems, subsystems, trains, or other -

designated components ob'ained by dividing the specified test interval into n equal subintervals, and r

b. The testing of one system, subsystem, train, or other designated l

,y component at the beginning of each subinterval. ;

LHERMALPOWER

1. THERMAL POWER shall be the totai reactor core heat transfer rate to the

] reactor coolant.

Trip ACTUATING DEVICE OPERATIONAL TEST i

1. A TRIP ACTUATING DEVILE 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 j

. at the required Setpoint within the required accuracy. ;

i UNIDENTIFIED LEAKAGE

~

1. JIDENTIFIED LEAKA'GE shall be all leakage which is not IDENTIFIED LEAKAGE' or r

TROLLED LEAKAGE. ;

UNRESTRICTED AREA 6

1.J!r An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY 4 access to which is not controlled by the i1censee for purposes of protection !

of individuals from exposurl to radiation and radioactive materials, or any l area within the SITE BOUNDARY used for residential quarters or for industrial, !

y. commercial, institutional, and/or recreational purposes, v.

l~ SEABROOK - UNIT 1 1-6 k ./ i 4n .

n - ._ , . _ - . - , _ . - - -

DEF8NITIONS i

,m VENTILATION EXHAUST TREATMENT SYSTEM i bh'g [4 '

1. 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.

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 TREATHENT SYSTEM components.

. VENTING VA.

1.ff VENTING shal's be the controlled process of discharging air or gas from a i confinement to maintain temperature, pressure, humidity, ventration, or other operating condition, in such a manner that repit,r.

air or gas is not provided or required during VENTING. Vent, used in systei.. names, does not imply a VENTING process.

r t

f r

I r

I e

f l

j i

f I

.gN i SEABROOK - UNIT 1 1-7 !

OCI b@

l i

-_-...f...... . . . . . _ , _ -

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

3/4.8 REACTIVITV CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUT 00VN 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 1-3% aklki d' ^ ' 't " ./ i~ " "<*/

Jf C/*C A1* 4,' M* $ 'h n^1 M t wns

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

With the SHUTOOWN MARGIN less than the limiting value, immediately initiate and continue boratien at greater than or toual to 30 gem of a solution containing greater than or equal to 7000 com baron or equivalent until the required SHUTOOWN. MARGIN is restored. l SURVEILLANCE REOUIREMENTS 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 inocerable.

If the inoperaDie control rod is immovable or untrippable, the aDove required SHUTOOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovaele or untrippable

' control rod (s);

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

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withcrawai is within the limits of Specification 3.1,3.6;

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

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 t

i

_ *See Special Test Exceptions Specification 3.10.1.

vOCr)

SEABROOK - UNIT 1 3/4 1-1

REACTIVITY CONTROL SYSTEMS B0 RATION CONTROL

(.y I' '

SHUTDOWN MARGIN T LE$5 THAN OR E0 VAL TO 200'F LIMITING CONDITION FOR OPERATION _

3.1.1.2 The SHUTDOWN MARGIN shall be greater than or eoual to t-5-ckA:- <--

Additionally, the Reactor Coolant System boron concentration shall be greater than or equal to 2000 ppm boron when the reactor coolant loops are in a drained condition. ,

jj, p ,j j , .,, f j{

c c Ar CAW F'G L/^f e G APPLICABILITY: MODE 5. I -

ACTION: +/ e liu s t <p c io' ed' i., H'e CS A 6

With the SHU 10WN MARGIN less than ' '*' "'k or the Reactor Coolant System boron concen'.r, tion less than 2000 ppm boron, immediately initiate and continue boration at gc_ater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTOOWN MMGIN and boron concentration are restored.

5.3N( [N,'((10.

SURVEILLANCE REQUIREMENTS Wd,f/4 I"

dr ai"' /

.. \

'd 'd 4.1.1.2 The SHUTDOWN MARGIN rd borea concentr4 Mon shall be determined to be greater *han or equal to 1.3 ok/k. , f% /,6+ rree,//</ /n M c- J o/A >d f At-

)

podu % /o.+Sgtv bor e n ce n:e n t <*t; con s),a tt la Se1rr#cdle be Drester +Aas W

a. Wituin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months af ter 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 SHUT 00WN 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 boren concentration,

2) Control rod position, -

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

'O SEABROOK - UNIT 1 3/4 1-3 J

$l @

REAC??VITV CONTROL Sv5TCHS BORATION CONTROL g,' .

MODERATOR 1EMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION

~

.L 3.1.1. 3 The moderator temperature coefficient (MTC) shall bed -st b M ( ban 4 yeerfied o n t i s. car mAa rin +m o rs werpr (cetA) ;

,s.

Less-positive-thea 0 Sk/c/or fe" the-4U-r4454tther4war 4eginning .

trf cycie 1ife-(B0L) - 50t-aw+-TMENAk SCWEL4ondition;--and-

b. Les+-negat4ve-thaptv2-.-*-10d-4k/'/*r for the au--thithec44,

.end-of-C-y440 ' ' f 0 ( ECL ) , " ^ TSO-THE AMAl--40WE R c cnd 444+n APPLICABILITY: d 6 ep;*+m ei cyd.f pee f4tet4e. e .fo/e {8ct)

.1..:. / Hit' 1 and 2* only**.

- MODES 5xcif4eetion 3.1.1. 36. - MODES 1, 2, and 3 only**

Sad of cy ele h /< (coq /,~'t-ACTION:

Bf , , , , , , , , , , ,,, g,,,z.

a. With the MTC more positive than the limit .of-Spec 444eet4en 3.1.1. 3e, above, operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 ok/k/ f W</Lv- .

.u within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STAND 0Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . I'**C Yh' These withdrawal limits shall be in addition to the insertion f/["

limits of Specification 3.1.3.6; c ,g a

2. The control rods are maintained within the withdrawal limits established above unsil a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and

3. A Special Report is prepared and submitted to the Commission, pursuant to Specification 6.8.2, within 10 days, deteribing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limi+. iar the all rods withdrawn condition.

b. With the MTC more negative than the limitS.+pedlin d in fle3.1.1.

Speci fication Co t-A 3tn *

-ebevej- be in HOT SHUT 00WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i *With kgf greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

SEABROOK - UNIT 1 3/4 1-4 kI

i-REACTIVfTV CONTROL SYSTEMS

[ BORAT10N CONTROL H00ERATOR TEMPERATURE COEFFICIENT Sj$VE!LLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits durino each fuel cycle as follows:

speal..rdin /4 CM

a. The MTC shall be measured and compared to the BOL limit +f Specifi-4: tier 3.1.1.32., a kve, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and

b. The MTC shall be measured at any THERMAL POWER and compared to 3.3 19 ' ak/k/AF (all rods withdrawn, RATED THERMAL POWER condi-tion) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more ,

negative than 0.0 x 1% ' ak/k/Ef, the MTC sht11 be remeasured, and compared to the EOL MTC limit -c' 5pt:f f4cet ten 3.1.1. 35. , at least once per 14 EFPD during the remainder of the fuel cycle, s $fteihed in ;

e / c. C 04- A. l Y

t h t- 3CO ffm su r v e r'//e n e t.

lin, o' f^ Spe si$a ed io, f 4 s.

Coe /L j i

l l

I r

4 i

f i

D SEi ,00K - UNIT 1 3/4 1-5 ,

e ;

_ _ _ . * - - . , . _ _ . _ . , . , . . _ . _ _ _ _ _ _ _ _ . - . ~ . . ._ _ ,__, - _.

. +

REACTIVITY CONTROL SYSTEMS

-BORATION SYSTEMS 0 FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following tnree boron injection flow paths shall be OPERABLE:

a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS), and

b. Two flow paths from the refueling water storage tank via charging pumps to the RCS.

APPLICABILITY: MODES 1, 2, and 3*

ACTION:

With only one of the above required boron injection flow paths tr. the RCS OPERABLE, restore at least two boron injection flow paths to the rlCS 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 and borated to a SHUTOOWN MARGIN equivalent to at least 4,-A-64 at 200*F within the nent 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

&;m) +A t //w l sf e e ifa ,f in f4L M &^MG SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a, At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

b. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal; and

c. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

The provisions of Specifications 3.0.4 and 4.0.4 are. not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of f

one or more of tne RCS cold legs exceeding 375'F, whichever comes first.

SEABROOK - UNIT 1 3/4 1-8

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS

$.e ,fk CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.*

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps 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 and borated to a SHUTDOWN MARGIN equivalent to at least 4-M-agat 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charwing pumps to OPERIBH status within the next /

7 days or be in COLD SHUTOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . /

-H e /,w & g eilied 4 Ha ccAcenorma SURVEILLANCE REOUIREMENTS 1 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2480 psid is developed when tested pursuant A&

. to Specification 4.0.5.

l l

i "The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 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

7,, . of the RCS cold legs exceeding 375 F, whichever comes first.

SEABROOK - UNIT 1 3/4 1-10 /

REACTIVITY CONTROL SYSTEMS

-. BORATION SYSTEMS

$$C BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2:

a. A Boric Acid Storage System with:

1) A minimum containec borated water volume of 22,000 gallons,

2) A minimum boron concentration of 7000 ppm, and

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volu'ne of 477,000 gallons,

2) A minimum boron concentration of 2000 ppm,

3) A minimum solution temperature of 50'F, and Nhk

4) A maximum solution temperature of 98'F.

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

a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the system 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 borated to a SHUTDOWN MARGIN equivalent to at least 1.35 M /k at 200*F: restore the Boric Acid Storage System to OPERABLE stat 6s within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With the RWST inoperable, restore the tank to OPERABLE status b.

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next b hours and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

//t lr% if re d h ) in +/L citPE~ OPErU T~tNC, L/M o rs l'E*PcFT (cci.A) for de aben Mopes SEABROOK - UNIT 1 3/4 1-12 l

l .

REACTfVITV CONTROL SYSTEMS

,s, 3/4.1.3 MOVABLE CONTROL _ ASSEMBLIES GROUP HEIGHT LIMITING CONDITTON FOR OPERATION 3.1. 3.1 All full-length shutdown and control rods shall be OPERABLE and positioned withia 12 steps (indicated position) of their group step counter i' demand position.

APPLICABILITY: . MODES 1* end 2".

ACTION:

a. With one or more full-length rods inoperable because of being immov-able as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN recuire-ment of Specification 3.1.1.1 is satisfied within I hour and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one full-lemp.i rod trippable but inoperable due to causes I other than addressed by ACTION a., above, or misaligned from '

its group step counter demand height by more than ! 32 steps (indicated position), POWER OPERATION may continue provided that .

w'ithin 1 hour: 1

$?.? 1. The rod is restored to OPERABLE status within the a1ove

'?

alignment requirements, or

2. The rod is dechired inoperable and the remainder of the rods in the group with the inoperable rod are aligned to with'n ! 12 steps of the inoperable rod while maintaining the rod sequerde and S .cita~ren. insertion limits of E' --

The THERMAL F0WER level shall 03 , / 3,4 be restricted pursuant 'to Specification 3.1.3.6 during sebsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these

conditions; b) The SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; 4

-, *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

SEABROOK - UNIT 1 3/4 1-15 4l

. . , - . -- .3..- ., ,

REACTIVITY CONTROL SVSTEMS

,. y- MOVA8LE CONTROL ASSEMBLIES Wll '

GROUP HEIGHT LIMITING CONDITION FOR OPERATION j I

3.1. 3.1 ACTION b.3 (Continued) c) A power distribution map is obtained from the aovable incore detectors and Fq(Z) and F g are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d) The THERMAL POWER lesel is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER.

c. With more than one rod trippable but inoperable due to causes other ;

than addressed by ACTION a. above POWER OPERATION may continue '

provided that:

1. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the rods in the bank (s) with the inoperable rods are aligned to within 12 steps of the i.h inoperable rods while maintaining the rod sequence and insertion :

Srcibeco,, limits of S g m .1-1. The THERMAL POWER level shall be i 3 l. 3 . I, restricted pursuant to Specification 3.1.3.6 during subsequent operation, and

2. The inoperable rods are restored to OPERABLE status within 72 -

hours.

d. With more than one rod misaligned from its group step counter demand height by more than 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1~.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions i at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except during time intervals when the rod position i deviation monitor is inoperable; then verify the group positions at least once '

per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days. ,

c I

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

N!

F y = c ....-----,----..--3 y.,,-.m.f.+.--.----r-.--- .-,.-r. m e y--. .--.--r c ,-.,yv- _m-a, - c, .w,.3 -,<-.ew- --,-,,,-w,,-,- y --vwr,-,7 -,,.,..y---,,r------,v-g-ve-+

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

REACTIVITY CONTROL SYSTEMS MOVABLE CONTROL ASSEMBLIES

($rs!'.

SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn,< 45 5f >e ci/<4 / /k f/*

c cAs* cerM rm'G Lis tr$ McPoA rccc A)

APPLICABILITY: MODES 1* and 2* **.

ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1. 3.1.

SURVEILLANCE REOUIREMENTS 4.1.3.5 Each shutcown rod shall be determined to be fully withdrawnCI ~~

. , . . as spr e.;f,t d m tA e. c ot A :

>Nji' a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or 0 during an approach to reactor criticality, and i

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

l l

l t!

I l4 '

i *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

,' **'iith k,ff greater than or equal to 1.

-}. ' .

h (7/ /3h7 SEABROOK - UNIT 1 3/4 1-21 h) f.

r e

REACTIVITY CON 7ROL SYSTEMS MOVABLE CONTROL AS$EMBLIES

' B,,s

i CONTROL R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as 4 6 iigure 3.1 1, spt e;itt 4 in tAs CMC cfcnn r,wa wo irs 66ren r /ccaA).

APPLICABILITY: MODES la and 2" **.

ACTION: '/'"

With the control tinks inserted beyond the -above-insertion limits except for surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the son 'icure or L.I in serta ra, fe m ts spe<<io r L. i n +ga_ CijX)

c. Be in at least HOT STANDBY within ,5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

.))y; SURVi1LLANCE REQUIREMENTS 4.1.3.6 The position of each control bank.shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except during time intervals when the rod insertion limit monitor is inoperable; . hen verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

s "See Special Test Exceptions Spe::ifications 3.10.2 and 3.10.3.

- With k,77 greater than or equal to 1.

SEABROOK - UNIT 1 3/4 1-22 ] 3h

. . I h 5 L E TF' 9

~ ,

(0.30,228) (0.844,228)

I Z ! I IJV 1) kIX /

/

BANK B l l/ jf 23, Jf f

- I X l l l / l i l

! l/ \ l / i i

/ f')

I N

E Il , ;

yg /(0.0,164) iN i/ f

{l 3 j -

I I

A l/ f'l -(1.0,14 6)-

i . i

! I !

m .

I

,y l ) l) 7; ag II I / O /

II N

~

!' I /l I /t i

!b ! / f~i X /' i

"?'

l /: / M

$ 80 I / fi I I } '\ i !

~

/l f: I l l l J/l ! T l l s t/! /1 1 i !II / I i i i

$ d,g' l'I V a

o 40 cr j

)

l I

I If

'6 '

BANK D l l l l

lh I -

I

/ (0.31.0.0)

I , I l l l . l l I I I ! l 0 i -

- 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER

,, TOUR-LOOP OPERATION SEABROOK - UNIT 1 3/4 1-23

_ - - . = . . ._

3/4.2 POWER OfSTRIBUTION LlHITS 3/4.2.1 AX1AL FLUX O!FFERENCE LlHITING CON 0lTION FOR OPERATION .---

3.2.1 The indicated AXI AL FLUX DIFFERENCE (AFD) shall be n,aintained within the L

-fo11c h 9s sp uta9 target t o, bandt-(flux tic ce4* cre dif e nference as L w aUnits) rs s mabout A T~ recut the tar

et flux dif ferencg

-a. 1-61Ir-fee-core-aveeege accumul4ted-burnup-.of 4ew-than-on-equal--to.

-3000 ND/Miih

-b.

Ag43-for-. core-aver 494-4ccumulat+d-bumup-obgreatee-thMOGO-

-WWHIUt-end-

c. - 3L J a fu--a ch-wb m us a-cy h y,gcq The indicated AFD may deviate outside the above- required target bandl at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits c' Figurr 3.2 1 and tne cumu-lative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . y The indicated AFD may deviate outside the ebeve required target bana k greater than 15% but less than 50% of RATED THERMAL POWER provioed the cumulative penalty deviation time does not exceed I hour during 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."

6 ACTION:

a. With the indicated AFD outside of the 4 e4+ required target band and witn '

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 ebeve required targat 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- O gure 3.2 ' and with THERMAL POWER less than 90% but equal to or greater than 50%

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 .

5ee 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 2-1 within the Acceptable Operation Limits of Figurc \. 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 $I r g D 4- .7_ mm- -,w -. - r,, .. ..ge.# e -.,.p. ,

y,g

POWER DISTR 8BUT80N 1,8MITS ;

3/4.2.1 - XIAL FLUX DIFFERENCE !

1.IMITING CONDITION FOR OPERATION ;

i 3.2.1 _ ;

ACTION: (Continued) 8"

~ '

c. With the indicated AFD outside of the-ahv+ required target bandf fer 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 and with THERMAL POWER less than 50% but greater [

than 15% of RATED THERMAL POWER, the THEPyAL POWER shall not be '

increased equal to or greater than 50% of RATED THERMAL POWER.

SURVEILLANCE REOUIREMENTS ,

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 OPERAELE 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 excere 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 i two or more OPERABLE excere 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

-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 mincts of POWER OPERATION outside of the target band at THERMAL POWER levele between 15% and 50% of RATED THERMAL POWER.

' ~

-4.2.1.3 The target flux difference of each OPERABLE excere channel shall te 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-target 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 between the- ,

L most recently measured value and the predicted value at the end of the cycle l~ life. The provisions of Specification 4.0.4 are not applicable, i-l **(Continued) operation may be accumulated with the AFD outside of the e w n L1% required target band,during testing without penalty deviation.

SEABROOK - UNIT 1 3/4 2-2 l N Y. hs ,

L L

!~ . - .. _ , _ -._ i ._ _ .. _ _ _ ._ _ _ _ . , _ . _ . , _ . _ . . _ _ . _ , _ . _ . . _ . ~ _ , _ . _ _ _ . _ . -

\

1)g w rr \

m 1 Y 120 eI ~

I E

! l Isp!!

3eg LI- ! l* !;

U,NACCEPTABLE OPEfATION l A UN' CCEPTABLE OPER4T10i[

j\ i-n.90i l <!).90) I l f Fu i l 'k I i ~L l \l l l # l 8 ,o i l I Nl/ i i VI' i

1 i l l ! , ! !

a , , . . ,

e \',

h l l [ 4 k TfBLE [ 'T]Oh \

s,

e I I / V I \

i i I /l l I # X l I l\ l Pu i / / \ \

g,Q q l-31.50) [ Nl (31.5 0) l 40 a l i# ! ! I ii\ i i

N I I l l l\ l

/ 1 1 \'

,_ i . l ! \

l }

i Il l l l k I

I'

! ! I I I .

e , ,

-50 -40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE (alW.

FIG'JRE 3.2-1

/J. AXIAL FLUX DIFFERENCE LIMITS A5 A FUNCTION OF U- RATED THERMAL POWER SEABROOK - UNIT 1 3/4 2-3 1l

POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLJXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

9 FA (2) $ M Z) for P > 0.5 [9 P

F(Z)1('.5QK(Z)forP$0.5 n nre Where-THERMAL POWER , a[ g@

'6 J -

P = RX_TED THERMAL POWER f K(Z) = the + ee44en Obtained 'ro- c19ere 3.2-2 <e ,

( 3 giver > core height !csat4en, /4e o or ,,,e/ / t e d Fa(z.) a s a Mc% ef wc A ey/+ as APPLICABILITY: MODE 1. c eA4.

s f a c c/ ce d A N ..

A_CTION:

With F (Z) exceeding its limit:

9

a. Reduce THERMAL POWER at leest 1% for each 1% F (Z) exceeds the 9

limit within 15 minutes and similarly reduce the Power Range Ty Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Set-points have been reduced at least 1% for each 1% F (Z) exceeds the limit, and 9 b, Identify and correct the cause of the out of-limit condition prior to increasing THERMAL POWER acove the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased, provided F (Z) is demonstrated through incere mapping to be 9

within its limit.

p i

w __ F*=q +Ae Fq /, ,., , .r. y RArip 77 casw RM 1

code orcM nm, un irs

.;p e.; p, e d in H<

RCr'cR T' (C ou R, ) , q n al l

l

% e.

j

. ~ , -

~[

1 i

l l l l l l l l l l IX l l kl l l I; I ! l l l i

l i

i i I

[

I l l lk I l h6.0.1.cd l l l } l [l I l l !I l

i 1 i

I l ! >

' 4.e.su>l i l !

~

l i l l l l I

' \l l l i l ! I : l l l o

i ! i l i i j ih,l hP I i t i l l g .e _

\7 , ,

f I i i ! I 1 : i 0 ! I i i. ! I l ! ._ lii i i l l W i! Ii : I i l l Xii ! "2^Y5' .

) ,,' l l l l l[ ' l l I I ki i l i !

I l l [ l ! l l l I l :1 i ; i !

6, ! Mi i l l I I i I i !

I 0 .4 l

i

)

i j

l

}

l l #

l

'i l i I. l i

I Kl i l [

D -

i i x l l i, I i l I lI , ,i i

i i i ,

I ! l !I I I I l l l l .' i i I (

c.2 l ! I IIi I l-I l

I l-l 1 l l ! i I

l l l l l i I I I I l l l l l I ! I i i ! I i l I !

~i l l I 1 !I I I I I i i I l l l l l i I I i ! !

i i , i I !

e.e . . ;

2 4 le 2 6 8 12 CORE HEIGHT (FT) l I

i i

b FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT 9

SEABROOK.- UNIT 1 3/4 2-5

{1

4 POWER DISTRIBUTf0N LIMITS HEATFLUXHOTCHANNELFACTOR-Fg l

..y

t. . .

SURVEILLANCE REQUIREMENTS

- i 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F yy shall be evaluated to determine ifqF (Z) is within its limit by:

a. Using the movable incere detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATCO THERMAL POWER,

b. Increasing the measured F,y component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties,

c. Comparing the F con.puted (F ) obtainud in Specification 4.2.2.2b.,

above, to:

1) The F limits for RATED THERMAL POWER (FRTP) for tu ap p priate measured core planes given in Specification 4.2.2.2e. and f.,

below, and

2) The relationship: 'PFq l RTP F =F xy gy 3.p) xy

//17 is t h Where F,l is the limit for fractional THERMAL POWER operation cy'{. ' [ expressed as a function of F and P is the fraction of RATED

'g g effad THERMAL POWER at which F,y was measured.

i,, H,E LA /

o. Remeasuring F,y according to the following schedule:

1) When F is greater than the F limit for the apprupriate measured core plane but less than the F relationship, additional power distribution maps shall be taken dF compared to F, and F,y either:

a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter exceeding by 20% of RATED THERMAL -

C POWER or greater, the THERMAL POWER at which F*Y ,,,),3g determined, or b)

At least once per 31 Effective Full-Power Days (EFPD),

whichever occurs first.

SEABROOK - UNIT 1 3/4 2-6 3

POWER DISTRIBUTf0N LIMITS HEATFLUXHOTCHANNELFACTOR-Fg y '

SURVEILLANCE REQUIREMENTS 4.2.2.2d. (Continued) l 2) khen the F,C is less than or equal to the F,RTP limit for the appropriate measured core plane, additional power distribution maps shall be taken and F,C compared to F,RTP and F, at least once per 31 EFPD. .

e. The F,y limits for RATED THERMAL POWER (F,RTP) shall be provided for all core planes containing Bank "D" control rods and all unrodded core planes in +-hdf:1 P0:Mr.g F:ctor L!*44-hem 4- per Specifica-tion 6.8.1.6: W CW C!rnerwa umrs ?recnt-

-f. The F,y limits of Specification 4.2.2.2e., above, are not applicable

, 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,

, . s,9 i *

3) Grid plane regions at 17.8 t 2%, 32.1 2%, 46.4 2 2%, 60.6 t 2%,

and 74.9 2 2%, inclusive, and

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

g. With F x C exceedingF,f,theeffectsofF xy n Fq (Z) shall be evaluated to determine if F q (Z) is within its limits.

4.2.2.3 When qF (Z) is measured for other than F,y determinations, an overall measured qF (2) 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. ,

4 4

SEABROOK - UNIT 1 3/4 2-7

%9/ [R 1

POWER DISTRIBuff0N LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR l 3]::

' LIMITING CONDITION FOR OPERATION 3.2.3 Ffg shall be less than y (1.0 + d (1-P)). { PFu )

Where: 7 P= THERMAL POWER RATED THERMAL. POWER --F7 \

APPLICABILITY: MODE 1. &[': tat F[ /, A ,* ,, r ,e,in s n e, . / /L,r pre; ACTION: syn,1au oe ne. ese marwa L m, rr N

e m r (ce w o ,. /

With FtH exceeding its lim'.: # ' * ^"# ' #" ^# #f yeerl ad m tu" ccLR,I ### .,4 '" #[N 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 L1HITING CONDITION FOR OPERATION is satisfied.

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

above; THERMAL PCWER may then be increased, provided F g is demonstrated through incore mapping to be within its limit.

SURVEILLANCE RE0VIREMENTS

- .;f 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 Ffg shall be demonstrated to be within its limit prior to operation 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.

b. UsingthemeasuredvalueofFhwhichdoesnotincludeanallowance for measurement uncertainty.

m

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 2 3/4.1.1 BORAT10N CONTROL 4

t 3/4.1.1.1 and 3/4.1.1.2 SHUT 00WN MARGIN L

$ A suf ficient SHUTOOWN MARGIN ensures that: (1) the reactor can be made h subcritical from all operating conditions, (2) the reactivity transients asso-t ciated with postulated accident conditions are controllable within acceDtable -

aa limits, and (3) the reactor will be maintained sufficiently subcritical to V

'?, t preclude inadvertent criticality in the shutdown condition. d 44 i 9 SHUTCOWN MARGIN revirements vary throughout core life as a function of

}{

fuel depletion, RCS boren concentration, and RC5 T,yg. The most restrictive ;

gj condition occurs at EOL, with T avg at no-load operating temperature, and is q Yt associated with a postulated steam line break accident and resulting uncon- N rM trolled RCS cooldown. In the analysis of this accident, a minimum SHUTOOWN t

,, 4 MARGINpF12 3*, Sh/k is required to control the reactivity transient. %

~

3& Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting '>

condition and is consistent with FSAR safety analysis assumptions. With T h less than 200'F, the reactivity transients resulting from a poststated steam ;

line break cooldown are minimal. A 1.2% ak/4 SHUTDOWN MARGIN W.d a boron concentration of greater than 2000 ppm are requireo to M,mit sulficient time

~

for the operator to terminate an inadvertent boren dilution event with T e less than 200'F. ""9 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of %

"~O t plant conditions; accordingly, verification of MTC values at conditions othar 4 d than those explicitly stated vill require extrapolation to those conditions in L Q order to permit an accurate comparison. j v >

The most negative MTC, value equivalent to the most positive moderator 4 density coefficient (MDC), was obtained by incrementally correcting the MDC ,

3>

i used in the FSAR analyses to nominal operating conditions. These corrections

& involved subtracting the incremental change in the MDC associated with a core i s condition of all rods inserted (most positive MDC) to an all rods withdrawn )

condition and, a conversion for the rate of change of moderator density with $.

)** _

temperature at RATED THERMAL POWER conditions. This value of the MDC was then ,

transformed into the limitina+MTC value' A.2 : 10 4 ok/k/". The.MTC value o T of -;.; ;-10

ak/k/ [ represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value,of H. 2-x 10 ' ak/k/ r.

soo ppm sur w \tme (i'*d]

SEABROOK - UNIT 1 B 3/4 1-1 kf

/cN go

REACTIVITV CONTROL SYSTEMS BASES BORATION CONTROL 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (Continued)

The Surveillance Requirements for measurement of tne MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the g reduction in RCS boron concentratio,n associated with fuel burnup, t x Q 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY

-This specification ensures that the reactor will not be made critical p2 with the Reactor Coolant System average temDerature less than 551?F. This limitation is required to ensu.e: (1) the moderator temperature coefficient i is within its analyzed temperature range, (2) the trip instrumentation is within 4

its normal operatiag range, (3) the pressurizer is capable of being in an y OPERABLE status with a steam bubble, and (4) the reactor vessel is above its g minimum RT ND7 temperature.

E 3/4.1.2 BORATION SYSTEMS e

The Boron Injection System ensures that negative rea(tivity control is W available during each mode of facility operation. The components required to

& perform this function include: (1) borated water sources, (2) charging pumps, L (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency

, power supply from OPERABLE diesel generators.

, N

With the RCS in MODES 1, 2, or 3, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an Q assumed failure renders one of the flow paths inoperable. The boration

.: capability of either flow path is sufficient to provide a SHUTDOWN

% MARGIN,from expected operating conditions-of 1.3% ak/h after xenon decay 3 and cooldown to 200'F. -The maximum expected boration capability requirement

} occurs at EOL from full power equilibrium xenon conditions and requires

, 22,000 gallons of 7000 ppm borated water from the boric acid storage tanks or a a minimum contained volume of 477,000 gallons of 2000 ppm borated water from a the refuelir.: water storage tank (RWST).

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 provides assurance that a mass addition pressure transient can De relieved by operation of a single PORY or an RHR suction, relief valve.

As a result of this, only one boron injection system is available. . This is acceptable on the basis of the stable reactivity condition of the reactor, the emergency power supply requirement for the OPERABLE charging pump and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the singie injection system becomes inoperable.

7 SEABROOK - UNIT 1 B 3/4 1-2 LY9 .

L .

REACTfVITY CONTROL SYSTEMS BASES 9 3/4.1.2 BORAT10N SYSTEMS (Continued) i I The boron capability required below 200'F is sufficient to provide a

't

$_HUTDOWN MARGIN f

$ 14PF. This con # 4 2% 'h/h after xenon decay and cooldown from 200 F to dition requires a minimum contained volume of 6500 gallons of 7000 ppm berated water from the boric acid storage tanks or a minimum Y

{ contained volume of 24,500 galloas of 2000 ppm borated water from the RWST.

v The contained water volume limits include allowance for water not available y because of discharge line location and other physical characteristics.

e The limits on contained water volume and boron concentration of the RWST Nh also ensure a pH value of between 8.5 and 11.0 for the solution recirculated gg y y within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on s< mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures M[g that this system is available for reactivity control while in MODE 6.

~

4

The limitations on OPERABILITY of isolation provisions for the Boron L Thermal Regeneration System and the Reactor Water Makeup System in Modes 3, 4, 5, and 6 ensure that the boron dilution flow rates cannot exceed the value 7 assumed in the transient analysis.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 1 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210, and 228 steps with-crawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication, Since the Digital Rod Position Indication System does not indicate the actual shutdown rod l position between 18 steps and 210 steps, only points in the indicated ranges .

are picked for ve ification of agreement with demanded position.

l The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. ' Misalignment of a red requires measurement

, of peaking factors and a restriction in THERMAL POWER. These restrictions pro-

! vide assurance of fuel rod integrity during continued operation. In addition, j those safety analyses affected by a misaligned rod are reevaluated to confirm I

that the results remain valid during future operation.

SEABROOK - UNIT 1 B 3/4 1-3 f 9 88-

+

3/4.2 POBER OlSTR! BUTTON LlHITS Q1 y

BASES The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and !! (Incidents of Moderate Frecuency) events by: (1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance.

criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F0 (Z) Heat Flux Hot Channel Fletor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation I divided by the average fuel rod heat flux, allowing for menufacturing tolerances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Cnannel Factor, is defined as the ratio of Fh*

the integral of linear power along the rod with the highest integrated power to the average rod power; and

^

F Radial Peaking Factor, is defined as the ratio of peak power density

*/(Z) to average power density in the horizontal plane at core elevation Z.

M N # *" M '* *

3/4*2*1 AXIAt* FLUX DIFFERENCE , crevnn um s an,n-(cccM The limits on AXIAL FLVX OlFFERENCE (AFO) assure that the Fn (Z) upper bound envelope of ' ' times the normalized axial peaking factor is not exceeded dur-ing either normal operation or in the event of xenon redistribution folio.<ing power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positiened within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these. conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL. POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations. .

Although it is' intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL' POWER levels. This deviation will not affect the xenon redistribution suf ficiently to change the

[, envelope of peaking factors which may be reached on a subsequent return to SEABROOK - UNIT 1 B 3/4 2-1 6{

D

POSER DISTRIBUTION LIMITS

.. g ,

BASES b

3/4.2.2 and 3/4.2.3 HEAT F_ LUX HUT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continuto)

N F

g will be maintained within its limits provided Conditions a. through

d. above are maintained. The relaxation of C SH as a function of THERMAL PCWER allows changes in the radial power shape for <ll permissible rod insertion limits.

Fuel rod bowing reduces the value of DNBR. Credit is available to offset this reduction in the generic margin. The generic mar 0 ins, totaling 9.1*. DNBR completely offset any rod bow penalties. This margin includes the following:

a. Design limit DNBR of 1.30 vs. 1.28,

b. Grid spacing (K,) of 0.046 vs. 0.059,

c. Thermal diffusion coefficient of 0.038 vs. 0.059,

d. DNBR multiplier of 0.86 vs. 0.88, and

e. Pitch reduction.

The applicable values of rod bow penalties are referenced in the FSAR.

h When an gF metsurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping Systsm, and a 3% allowance is apprcpriate for manufacturing tolerance.

The Radial Peaking Factor, F,y(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit. The RTP g F

xy limit for RATED . m,g- THERMAL POWER (Fxy ) as provided in the R N Ys k "

& cter L. 7. RepcM-per Specification 6.8.1.6 was determined from expected pcwer control maneuvers over the full range of burnup conditions in the core.

WhenRC5Fhismeasured,noadditionalallowancesarenecessarypriorto comparison with the established limit of a measurement error of 4% for F N has been allowed for in determination of the design DNBR value. OH 3/4.2.4 QUADRANT POWER TILT RATIO The purpose of this specification is to detect gross chpnges in core power distribution between monthly incore flux maps. During normal operation the QUADRANT POWER .'llT RATIO is set equal to zero ence acceptability r:f core peaking factors has been established by review of incore maps. The limit of l 1.02 is established as an indication that the power iistribution has changed

,.. enough to warrant further investigation.

q

SEABROOK - UNIT 1 B 3/4 2-3 \' '

42f/ go a

ADMIN 151RATIVE CONTROLS

%y SEMIANNUALRADIOACTIVEEFFLUENTRELEASEREPORJ 6.8.1.4 (Continued) to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable trethods for calculating the dose contribution from licuid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

l 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 liquio 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 ODCM, pursuant to Specifications 6.12 a N 6.13, respectively, as well as any major change to Liquid, Gaseous, or Sol.d Rad aste Treatment Systems pursuant to Specification 6.14 It shall also incluJe a listing of new locations for dose calculations and/or environmental nonitoring identitied by the Land Use Census pursuant to Specification 3.12.2, The Semiannual Radicactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.9 or 3.3.3.10, respectively; and description of the

., events leading to liquid holdup tanks or gas storage tanks exceeding the

i limits of Specification 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORT 5 6.8.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory. Commission, Washington, D.C. 20555, Attn: Cocument Control Desk, with a copy to the NRC Regional Administrator, no later than the 15th of each month following the calendar month covered by the report.

c. i ; . % . u

[e std 6.8.1.o F. limits for RATED THERMAL F0WERP(F.TP) shall be provided to g ny the U.S. Nucle m latory Commission, Washington, D. C. 20555, Attn:

Document Control Des , '+h a copy to the NRC Regional Administrator, for all core plants containing "

"0" control rods and all unrocced core planes T

and the plot of predicted (F *P q pg, Axibl Core Height with the liTit en-velope at least 60 days prior to each.cyc itial criticality unless other-wise approved by the Commission by lecter. In 'on, in the event that the limit should thange requiring a new sutstantial er a. ded suomittal to the Radial Peaking Factor Luit keport, it will be submitted .. 's prior to the date the limit would become e*fective unless otherwise approve . %e Conmis" sion by it.tt.e r. Any infornation needed to succort F will be by requ .

from tot NRC anc need not te included in this report.

SEABROOK - UNI 1 1 6-18

l ,

INSERT B FOR PROPOSED TECHNICAL SPECIFICATION CllANGES ADh)LNINTR ATIVE COHILOI.S CORE OPERATING IIMITS REPORT 6.8.1.6.a. Core operating limits shall be established and documented in the CORE OPERATING Lih.'TS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

1. SKUTDOWN M ARGIN limit for M O D Es 1, 2, 3, a n d 4 for Specification 3.1.1.1,

2. SHUTDOWN M ARGIN limit for MODE 5 for Specification 3.1.1.2,

3. Moderator Temperature Coefficient BOL and EOL limit s, and 300 ppm surveillance limit for Specification 3.1.1.3, 4, Shutdowa Rod insertion limit for Specification 3.1.3.5,

5. Control Rod Bank Insertion limits for Specification 3.1.3.6,

6. AXI AL FLUX DIFFEP.ENCE limits and target band for Specifi:ation 3.2.1,

7. Heat Flux Hot Channel Fact or, F""'o, K(Z), F"",,, and the Power Factor Multiplier for F,, for Specification 3.2.2,

8. Nuclear Enthalpy Rise Hot Channel Factor, F,,,,, n"" and the Powei Factor Multiplier for F ..n. o for Sp cification 3.2.3.

The CORE OPERATING LIMITS REPORT shall be maintained available in the Control Room.

5.8.1,6.b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1. WCAP-9272 P-A, " Westinghouse Reload Safety Evaluation Methodology" July 1985 (E F.oprietary)

Methodology for Specifications:

3.1.1.1 - SHUTDOWN M ARGIN limit for MODES 1, 2, 3 and 4 3.1.1.2 - SHUTDOWN M ARGIN limit for MODE 5 3.1.1.3 - Moderator Temperature Coefficient 3.1.3.5 - shutdown Rod Bank Insertion Limit 3.1.3.6 - Control Rod Bank Insertion Limits 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor

2. WCAP 11596 P-A, 'Oualification of the Phoenix P/ANC Nuclear Design System for Pressurized Water Reactor Cores

June 1988 (W Proprietary)

Methodology for Specifications:

3.1.1.1 - SfiUTDOWN M ARGIN limit for MODES 1, 2, 3 and 4 3.1.1.2 - SHUTDOWN M ARGIN limit for MODE 5 3.1.1.3 - Moderator Temperature Coefficient

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

3,' WCAP 8385 P A, " Power Distribution Control and Load Following Procedures Topical Report", Septembe 1974 (E Proprietary)

Methodology for Specifications:

3.1.3,5 Shutdown Rod Bank Inscrtion Limit 3.1.3.6 Control Rod Bank Insertion Limits 3.2.1 AXI AL FLUX DIFFERENCE 4 WCAP 7811, " Power Distribution Control of Westinghouse Pressurized Water Reactors". December 1971 (E Proprietary)

-Methodology for Specifications:

3.1.3.5 Shutdown Rod Bank Insertion L. ..J 3.1.3.6 Control Rod Bank Insertion Limits

5. Letter, T.M. Anderson to K. Kneil (Chief of Core Performan.,e, Branch, NRC),

January 31, 1980,

Attachment:

Operation and Safety Analys s Aspects of an improved Load Follow Package Methodology for Specification:

3.2.1 AXIAL FLUX DIFFERENCE

6. NUREG 0800, Standard Review Plan, US Nuclear Regulatory Commission, Section 4.3, N:: clear Design.' July 1981, Branch Technical Position CPB 4.31, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981 Methodology for Specification:

3.2,1 AXI AL FLUX DIFFERENCE

7. WCAP 7308 L, " Evaluation of Nuclear Hot Channel Factor Uncertainties'

? s cember 1971 (E Proprietary)

Methodology for Specification:

3.2.2 Heat Flux Hot Channel Factor

8. WCAP-8622, ' Westinghouse ECCS Evaluation Model, October,1975 Version",

November 1975 (E Proprietary)'

Methodology for Specification:

3.2.2 - Heat Flux Hot Channel Factor 9 WCAP-9220, ' Westinghouse ECCS Evaluation Model, February 1978 Version',

February 1978 (E Proprietary)

Methodology for Specification:

3.2.2 Heat Flux Hot Channel Factor

10. WCAP 7912-P A, " Power Peaking Factors", January 1975 (E Proprietary)

Methodology for Specification:

3.2.3 - Nuclear Enthalpy Rise Hot Charinel Factor

11. _ YAEC-1363 A, ?CASMO 3G Vrliuation,' April 1988, YAEC-1659 A, ' SIMULATE 3 Validation and Verification,' September 1988.

- Methodology for Specificatiods:

3.1.1.1 SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1,1,2 SHUTDOWN MARGIN for MODE 5 3.1,1,3 Moderator Temperature Coefficient 3.13.5 - Shutdown Rod Bank Insertion Limit 3.1.3.6 - Control Rod Bank Insertion Limits 3.2.1 AX1AL FLUX DIFFERENCE L -

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

l :.: ,

l r

' -3.2.2 Heat Flux Hot Channel Factor .

3.2,3 Nuclear Enthalpy Rise Hot Channel Factor l

12. Seabrook Station Updated Final Safety Analysis Report, Section 15.4.6,

Chemical and Volume Control System Malfunction that results in a Decrease in the Boron Concentration in the Reactor Coolant System".

(. Methodology for Specifications:

3.1.1.1 - SHIJTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5 6.8,1.6.c. The core operating limits shall be determined'so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal bydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident at alysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including an supplements thereto, shall be provided upon issuance,y to thetold NRC cycle revisions or Document Control Desk with copies to the Regional Administrator and the Resident laspector,

III, Retyne of Proposed Chances See attached retype of proposed changes to Technical Specifications.: The attached retype reflects the currently issued version of Technical Specifications. Pending Technical Specification changes or Technical Specification changes issued subsequent to thir submittal- are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to iscuance. 1 Revision bars are provided in the right hand margin in all cases where there is a change in the text. No revision bars-are utilized when the page is changed solely to accommodate the shifting of text due to !

additions or deletions, j

l l.

l 7 l

/ .

ATTACHMENT TO t1 CENSE AMEN 0 MENT NO.

FACIL11Y OPERATING LICENSE NO. NPE-81 DOCKET NO. 50-443 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overlap pages are provided for continuity.

Insert Remove i i INDEX) iii iii INDEX) xiv xiv INDEX) 1-2 l-2 1-3 1-3 1-4 1-4 1-5 1-5 1-6 1-6 1-7 1-7 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-3 3/4 1-4 3/4 1-4 ,/4 1-5 3/4 1-5 3/4 1-8 3/4 1-8 3/4 1-10 3/4 1-10 3/4 1-12 3/4 1-12 3/4 1-15 3/4 1-15 3/4 1-16 3/4 1-16 3/4 1-21 3/4 1-21 3/4 1-22 3/4 1-22 3/4 1-23 3/4 1-23 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 B 3/4 1-1 8 3/4 1-1 8 3/4 1-2 8 3/4 1-2 8 3/4 1-3 8 3/4 1-3 B 3/4 2-1 8 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4 2-3 6-18 6-18 6-18A*

6-188*

6-18C* d

Denotes new page i

INDEX 1.0 DEFINITIONS -

SECTION EAEE 1.1 ACTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2 ACTUATION LOGIC TEST. . . . . . . . . . . . . . . . . . . . . 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST . . . . . . . . . . . . . . 1-1 1.4 AXIAL FLUX DIFFERENCE , . . . . . . . . . . . . . . . . . . . 1-1 1.5 CHANNEL CALIBRATION . . . . . . . . . . . . . . . . . . . . . 1-1 1.6 CHANNEL CHECK . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.7 CONTAINMENT INTEGRITY . . . . . . . . . . . . . . . . . . . . 1-2 1.8 - CONTROLLED LEAKAGE. . . . . . . . . . . . . . . . . . . . . . 1-2 1.9 CORE ALTERATION . . . . . . . . . . . . . . . . . . .... 1-2 1.10 CORE OPERATING LIMITS REPORI. . . . . . . . . . . . . . . . . 1-2 1.11 ' DOSE EQUIVALENT l-131 . . . . . . . . . . . . . . . . . . . . 1-2 1.12 E - AVERAGE DISINTEGRATION ENERGY , . . . . . . . . . . . . . 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME. . . . . . . . . . . 1-3 1.14 FREQUENCY NOTATION. . . . . . . . . . . . . . . . . . . . . . 1-3 1.15 GASEOUS RADWASTE TREATMENT SYSTEM , . . . . . . . . . . . . 1-3 1.16 IDENTIFIED LEAKAGE . . . . . . . . . . . . . . . . . . . . . 1-3 '

1.17 MASTER RELAY TEST . . . . . , . . . . . . . . . . . . . . . . 1-3 1.18 MEMBER (S) 0F THE PUBLIC , . . . . . . . . . . . . . . . . . 1-4 1.19 0FFSITE DOSE CALCULATION MANUAL , . . . . . . . . . . . . . 1-4 1.20 OPERABLE - OPERABILITY. . . . . . . . . . . . . . . . . . . . 1-4 1.21 OPERATIONAL MODE - MODE , . . . . . . . . . . . . . . . . . . 1-4 1.22 PHYSICS TESTS . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.23 PRESSURE BOUNDARY LEAKAGE . . . . . . . . . . . . . . . . . 1-4 1.24 PROCESS CONTROL PROGRAM . . . . . . . . . . . . . . . . . . . 1-5 1.25 PURGE - PURGING . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.'26 QUADRANT POWER TILT RATIO , . . . . . . . . . . . . . . . . 1-5 -

1.27' RATED THERMAL POWER . . . . . , . . . . . . . . . . . . . . . 1-5 1.28 REACTOR TRIP SYSTEM RESPONSE TIME . . . . . . . . . . . . . . 1-5 1,29 REPORTABLE EVENT. . . . . . . . . . . . . . . . . . . . . . . 1-5 1.30 CONTAINMENT ENCLOSURE BUILDING INTEGRITY. . . . . . . . . . . 1-5 1.31 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.32 SITE B0UNDARY , . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.33 SLAVE RELAY TEST. . . . . . . . . . . . . . . . . . . . . -. . 1-6 1.34 SOLIDIFICATION. . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.35 SOURCE CHECK. . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.36 STAGGERED TEST BASIS. . . . . . . . . . . . . . . . . . ... . 1-6 1.37 THERMAL POWER , . . . . . . . . . . . . . . . . . . . . . . 1-6 1.38 TRIP ACTUATING DEVICE OPERATIONAL TEST. . . . . . . . . . . . 1-6 1.39 UNIDENTIFIED LEAKAGE. . . . . . . . . . . . . . . . . . . . . 1-7 1.40 UNRESTRICTED AREA . . . . . . . . . . . . . . . . . . . . . . 1-7 1.41 VENTILATION EXHAUST TREATMENT SYSTEM. . . . . . . . . . . . . 1-7 1.42 VENTING . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7

TABLE 1.1 FREQUENCY h3TATION , . . . . . . . . . . . . . . . . . . 1-8 TABLE 1.2 OPERATIONAL MODES. . . . . . . . . . . . . . . . . . . . 1-8 SEABROOK - UNIT 1 i Amendment No.

l

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

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVElllIANCE'RE0VIREMENTS

'SECTION PAGE 3/4.1.2 B0 RATION SYSTEM Flow Paths -- Shutdown................................ 3/4 1 Flow Paths - 0perating............................... 3/4 1-8 Charging Pump - Shutdown.............................. 3/4 1-9 Charging Pumps - Operating............................ 3/4 1-10 Borated Water Sources - Shutdown....................... 3/4 1-11 Borated Water Sources - Operating...................... 3/4 1-12 Isolation of-Unborated Water Sources - Shutdown........ 3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height........................................... 3/4 1-15 TABLE 3.1-1 ACCIDENT' ANALYSES REQUIRING REEVALUATION IN'THE EVENT OF AN IN0PERABLE FULL-LENGTH R00................... 3/4 1-17 i Position Indication Systems - Operating................ 3/4 1-18 Position Indication System - Shutdown................... 3/4 1-19 Rod Drop Time.......................................... 3/4 1-20 Shutdown Rod Insertion Limit........................... 3/4 1-21 Control Rod Insertion Limits....................... ... 3/4 1-22.

l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE,.................................... 3/4 2-1 I 3/4.2.2 HEAT-FLUX HOT CHANNEL FACTOR - F 3/4 2-4 3/4.2. 3 . NUCLEAR ENTHALPY RISE HOT CHANNEb(Z)

FACT 0R..................- ...

3/4 2-8 .......

I 3/4.2.4 QUADRANT POWER TILT RATI0................................. 3/4 2-9 >

, 3/4.2.5 DNB PARAMETERS............................................ 3/4 2-10 ;

3/4.3 INSTRUMENTATION i

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................... 3/4 3-1 l

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 i i

SEABROOK - UNIT 1 iii -

Amendment No.

~

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

JEDM 6.0 ADMINISTRATIVE CONTROLS SECTION PAGE-6.4 REVIEW'AND AUDIT ............................................. 6-6 6.4.1 STATION OPERATION REVIEW COMMITTEE (SORC)-

Function..................................................... 6-6 Composition.................................................. 6-6 Alternates................................................... 6-6 Meeting frequency............................................ 6-6 Quorum....................................................... 6-6 Responsibilities............................................. 6-6 Records...................................................... 6-8 6.4.2-NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC)

Function..................................................... 6-8 Composition.................................................. 6-8

-Alternates................................................... 6-8 Consultants.................................................. 6-8 Meeting Frequency............................................ 6-9 Quorum....................................................... 6-9 Review....................................................... 6-9 Audits....................................................... 6-9 Records...................................................... 6-11

-6.5 REPORTABLE EVENT ACT10N........................................ 6-11 6.6 SAFETY LIMIT V10LATION......................................... 6-11 6.7 PROCEDURES AND PR0 GRAMS........................................ 6-12 ,

6.8 REPORTING REQUIREMENTS 6.8;l ROUTINE REP 0RTS.............................................. 6-14 Startup Report............................................... 6-14 ,

i Annual Reports............................................... 6-15 Annual Radiological Environmental Operating Report........... 6-15 l Semiannual Radioactive Effluent Release Report...............- 6-17 Monthly Operating Reports.................................... 6-15 .

CORE OPERATING LIMITS REP 0RT................................. 6-18 l 6.8'2 SPECIAL REP 0RTS..............................................

. 6-19 [

L o

6.9 RECORD RETENTION............... . ............................... 6-19 .

i 6.10 RADIATION PROTECTION PR0 GRAM.................................. 6-20 ;

f SEABROOK - UNIT 1 xiv Amendment No. j t.

i k

DFFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3,6.1.2, and

e. The sealing mechanism associated with each penetration (e.g.,

welds, bellows, or 0-rings) is OPERABLE.

(ONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT provides core operating limits for the current operating reload cycle. The cycle specific core operating limits shall be determined for each reload cycle in accordance with Soecification 6.8.1.6. Plant operation within these operating limits is addressed in individual specifications.

DOSE EOUIVALENT l-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, 1-133, 1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109, Revision 1,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1."

SEABROOK - UNIT 1 1-2 1 Amendment No. l

l DEFINITIONS E - AVERAGE DISINTEGRATION ENFRGY 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample with half-lives greater than 10 minutes.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitoring 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 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASE0US RADWASTE TREATMENT SYSTEM 1.15 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 f or delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

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 ar.d 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 1.17 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

SEABROOK - UNIT 1 1-3 Amendment No.

DEFINITIONS MEMBER (S) 0F THE PUBLIC-1.18 MEMBER (S) 0F THE PUBliC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its coatractors, 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 u'se portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.19 The OFFSITE 00SE CALCULATION MANUAL (ODCM) shall contain in Part A the radiological effluent sampling and analysis program and radiological environmental 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 Environmental Radiological Monitoring Program.

SPERABLE - OPERABILITY

-1.20 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 al,;o capable of performing their related support. function (s).

( OPERATIONAL MODE - MODE

, 1.21 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 i

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

-(l) 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 L 1.23 PRES $URE B0UNDARY LEAKAGE shall be leakage (except steam generator tube l leakage) through a nonisolable fault in a Reactor Coolant System component-L body, pipe wall, or vessel wall.

L l

SEABROOK - UNIT 1 1-4

-Amendment No.

DEFINITIONS PROCESS CON 1ROL PROGRAM 1.24- 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 requirements governing the disposal of radioactive waste.

PURGE - PURGING

-1.25 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.

- OVADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated

. outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitoring paraneter exceeds its Trip Setpoint at the channel sensor rntil loss of stationary gripper coil voltage.

REP 0RTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section l 50.73 of 10 CFR-Part 50.

CONTAINMENT ENCLOSURE BUILDING INTEGRITY 1.30 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 i-

c. The sealing mechanism associated with each penetration.(e.g.,

welds, bellows, or 0-rings) is OPERABLE.

SEABROOK - UNIT 1 1-5 Amendment No.

l t_______-___.. . . , . . ._ -

p.

t i

= DErlN[11QNS

~ SHUTD0WN MARGIN 1.31 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactoriis subcritical or would be subcritical from its present condition j l assuming all-full-length rod cluster. assemblies (shutdown and control) are l fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY ,

1.32 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST 1.33 A SLAVE RELAY TEST =shall be the energization of each slave relay and verification of OPERABILITY of each relay. _The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION 1.34 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets. shipping.and= burial ground requirements, i SOURCE CHECK -

1.35 A SOURCE CliECK shall be the qualitative assessment of channel response when the channel sensor ~is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS-1.36 EA STAGGERED TEST BASIS shall consist of:

a.- A test schedule for n systems, subsystems, trains, or other 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 1,37 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

4 TRIP ACTUATING DEVICE OPERATIONAL TEST 1.38. 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 1 RIP ACTUATING DEVICE OPERATIONAL TEST shall include

adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Satpoint within the required accuracy.

--SEABROOK-- UNIT 1 1-6 Amendment No.

DEFINITIDNS UNIDENTIFIED' LEAKAGE

-1,39 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.40- An CNRESTRIClED AREA shall be any area at or beyond the SITE B0UNDARY-access t; .'hich is not controlled by the licensee for purposes of protection of faciviouals from exposure to radiation and radioactive materials, or any area within the SITE B0UNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.41: A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l 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 particulates-from the gaseous exhaust stream prior to the release to the

-environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Saf ety Features Atmospheric Cleanup Systems are not considered to-be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING _

l.'42 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 or gas is not provided or required during VENTING. Vent, used in system names, does not

imply a-VENTING process.

SEABROOK - UNIT 1 1-7 Amendment No. ]

l

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORAT10N CONTROL SHUTDOWN MARGIN -Twa. 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 the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

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 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE00lREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall oe determined to be greater than or equal to the limiting value:

a. Within I hour 'after detection of an inoperable control rod (s) and at least once per 12 hou*s thereaf ter 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);

b. When in MODE 1 or MODE 2 with k.u greater than or equal to 1 at

-least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

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

Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, ay consideration of the factors of Specification l- 4.1.1.1.le below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

See Special Test Exceptions Specification 3.10.1.

SEABROOK - UNIT 1 3/4 1-1 Amendment No. e

-er c w w

lEACTIVITY CONTROL SYSTEMS 4 30 RATION CONTROL" j

! SHUTDOWN MARGIN -Tip LESS THAN OR E0 VAL TO 200*F

~

LIMITING CONDITION FOR OPERATION ,

3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR). Additionally, the Reactor Coolant System boron concentration shall be greater than or equal to

'2000 ppm boron when the reactor coolant loops are in a drained condition.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than the limit specified in the COLR or the Reactor Coolant System boron concentration less than 2000 ppm boron, l 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 boron concentration are restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2.The SHUTDOWN MARGIN shall e determined to be greater than or equal to the limit specified in the COLR and the Reactor Coolant System boron concentration -shall be determined to be greater than or equal to 2000 ppm boron when the reactor coolant loops are in a drained condition:

a. Within I hour 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 allewance for the withdrawn worth of the immovable or untrippable control rod (s); and 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-b.

factors:

, 1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation, 4

5) Xenon concentration, and 4

6) Samarium concentration.

i l

SEABROOK - UNIT 1 3/4 1-3 Amendment No.

l

REACTIVITY CONTROL SYSTEMS l RORATION CONTROL MODERATOR TEMPERATURE COEFFICIENT llMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the CORE OPERATING LIMITS REPORT (COLR):

APPLICABillTY: Beginning of cycle life (BOL) limit MODES I and 2* only**.

End of cycle life (EOL) limit - MODES 1, 2, and 3 only**.

ACTION:

a. With the MTC more positive then the BOL limit specified in the COLR, operation in MODES 1 and 2 may proceed provideu.

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

2. The controi rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limits for the all rods withdrawn condition; and

3. A Special Report is prepared and submitted to the Commission, pursuant to Specification 6.8.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition,

b. With the MTC more negative than the EOL limit specified in the COLR, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With k.a greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

SEABROOK - UNIT 1 3/4 1-4 I Amendment No.

4 't M T1VITY CONTROL' SYSTEMS

- HORATION CONTROL MODERATOR TEMPERATURE COEFFICIENT-SVRVE?llANCE RE0VIREMENTS J

4.1.1.3 The MTC shall be determined to be within its limits during each fuel :

cycle as follows:

! a. The MTC shall be measured and compared to the BOL limit specified in the COLR, prior to initial operation above 5% of RATED THERMAL-POWER, after each fuel loading; and I

h. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods l withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the i event this comparison indicates the MTC is more negative than the i 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the E0L MTC limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel l cycle.

l

)

7 SEABROOK - UNIT 1 3/4 1-5 Amendment No.

REACTIVITY CONTROL SYSTEMS-BORAT10N CONTROL FLOW PATHS - OPERATING LIMITING CONDITION'FOR OPERATION .

1.

3.1;2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant _ System (RCS), and

b. Two flow paths from the refueling water storage tank via charging ,

pumps to the RCS.

APPLICABILITY: MODES 1, 2, and 3*

ACTION:

With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to

- 0PERABLE 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 and borated to a SHUTDOWN MARGIN equivalent to at least the limit specified in the CORE

- 0PERATING LIMITS REPORT (LOLR) for the above' MODES at 200'F within the next 6 '

hours;- restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,

SURVEILLANCE REQUIREMENTS I

4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual,_ .

power-operated, or-automatic) in the flow path that is not locked, ?

-sealed, or otherwise secured in position, is-in its correct position;

b. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a' safety injection test signal; and

c. At least once per 18 months by verifying that the flow path' '

required by Specification 3.1.2.2a. delivers at least 30 gpm to ;

the RCS.

a

The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging. pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is '

restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one oor more of the RCS cold ' legs exceeding 375'F, whichever comes first.

SEABROOK - UNIT 1 3/4 1-8 -

Amendment No. ;

i

[

i

, -. -4

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

i REACTIVITY CONTROL SYSTEMS 1

BORATION_ CONTROL THARGING PUMPS-- OPERAT1'g.J

-LIMITING CONDITION FOR OPERATION 3.1.2.4 At-least two charging pumps shall be OPERABLE.

APPLICABillTY: MODES 1, 2, and 3.*

ACTION:

With only one charging pump OPERABLE, restore-at least-two charging pumps 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 and borated to a

' SHUTDOWN MARGIN equivalent to at least the limit specified in the CORE OPERATING LIMITS REPORT (COLR) for the above MODES at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE stctus within the next 7 days or be.in COLD SHUTOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

t SURVEILLANCE RE0VIREMENTS 4,1,2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each

. pump of greater than or equal to 2480 psig is developed when tested pursuant

-to Specification 4.0.5.

The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal . charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 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.

L :SEABROOK - UNIT 1 3/4 1-10

' Amendment No.

REACTIVITY CONTROL SYSTEMS B0 RATION SYSTEM 1 80 RATED WATER-SOURCES - OPERATING

~ LIMITING CONDITION FOR OPERATION 3.1,2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of 22,000 gallons,

2) A minimum boron concentration of 7000 ppm, and ,

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWS1) with: j A minimum contained borated water volume of 477,000 gallons,

~

1)

2) A minimum boron concentration of 2000 ppm,

, 3) A minimum solution temperature of 50*F, and i

4) A maximum solution temperature of 98'F.

APPLICABillTY: MODES 1, 2, 3-and 4, L ACTION:

I a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the-system 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 I

STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent-to at least the limit specified in the CORE OPEP.ATING LIMITS REPORT (COLR):for the above MODES at 200*F; restore the Boric. Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l- b. With the'RWST inoperable, restore the tank to OPERABLE status l

within I hour or be in at least HOT STANDBY with 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 l

l:

L SEABROOK - UNIT 1 3/4 1-12 Amendment No.

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBilES GROUP HEIGHT

,LLMITING CONDITION FOR OPERATION 3.1,3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.

MPLICABILITY: MODES 1* and 2*.

ACTION:

a. With one or more full-length rods inoperable because of being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within I hour and be in HOT STANDB' within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With one full-length rod t' .._ie but inoperable due to causes other than addressed by AC110N a., above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within I hour:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable red are aligned to within i 12 steps of the inoperable red while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1,1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed re;ults of these accidents remain valid for the duration of operation under these conditions; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

See Special Test Exceptions Specifications 3.10.2 and 3.10.3. ;

I SEABROOK - UNIT 1 3/4 1-15 Amendment No.

REACTIVITY CONTROL SYSTEMS

-MOVABLE CONTROL ASSEMBIIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 ACTION b.3 (Continued) c)-

A power incore distribution detectors and map Fo(Z)is andobtained from thetomovable Fb are verified be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d) The THERMAL POWER-level is reduced to less than or equal to 75% of RATED THERMAL POWER within- the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to.

85% of RATED THERMAL POWER.

-c. With~more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue provided that:

1. Within I hour, the remainder of the rods in the bank (s)_with the inoperable rods are aligned to wi_ thin i 12 steps of the_ inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and 2 '. The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d. -With more than one rod misaligned from its: group-step counter demand height by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE0VIREMENTS 4.1.3.1.1: The.' position of each full-length rod shall be determined to bc within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except during time intervals when the rod position L

I deviation monitor is inoperable; then verify the group' positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be

-determined to be OPERABLE by movement of.at least 10 steps in any one L direction at-least once per 31 days.

SEABROOK - UNIT 1 3/4 1-16 Amendment No.

REACTIVITY CONTROL SYSTEMS MOVABLE CONTROL ASSEMBLIES SHUTDOWN R00 INSERTION LIMIT LIMITING CONDITION-FOR OPERATION 3.1.3.5 ~ All shutdown rods shall be fully withdrawn as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABillTY: MODES 1* and 2* **.

ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVElllANCE RE0VIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn as specified in the COLR:

a. Within 15 minutes prior to withdrawal of any rods in Control Barik A, B, C, or D during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 'thereafter.

r.

l.

L 1-

See Special Test fxceptions Specifications 3.10.2 and 3.10.3.

- With k,u greater than or equal to 1.

SEABROOK - UNIT 1 3/4 1-21 Amendment No.

REACTIVI*, LQNiROL SYSTEMS-MOVABLE CONTROL ASSEMBLIES

= CONTROL R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1* and 2* **.

ACTION:

With the control banks inserted beyond the insertion limits specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the COLR, or

c. Be in at least HOT STANC8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE0VIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except during time intervals when the rod insertion limit monitor is-inoperable; then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

- With k.tr greater than or equal to 1.

SEABROOK - UNIT 1 3/4 1-22 Amendment No.

i PAGE-INTENTIONALLY BLANK SEABROOK - UNIT 1 3/4 1-23 Amendment No.

3/4,2 POWER DISTRIBUTION LIMITS 3/4,2.1 AX1AL FLVX DIFFERENCE I

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (flux difference units) about the target flux difference as specified in the CORE OPERATING LIMITS REPORT (COLR):

The indicated AFD may deviate outside the required target band specified in the COLR at greater than or equal to 50% but less than 90% of RATED THERKTL POWER provided the indicated AFD is within the Acceptable Operation Limits specified in the COLR and the cumulative penalty deviation time does not exceed I hour 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 required target band specified in the COLR at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

ACTION:

a. With the indicated AFD outside of the required target band specified in the COLR 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 required target band spacified in the COLR 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 specified in the COLR 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 specified in the COLR, A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ' operation may be accumulated with the AFD outside of the SEABROOK - UNIT 1 3/4 2-1 f Amendment No.

j

' POWER DIS 1RIBUT N LIMITS 3/4.2.1 AXIAL FLUX DIFFEREN(1 LIMITING CONDITION FOR OPERATION _

3.2.1 ACTION

(Continued)

c. With the indicated AFD outside of the required target band as spt.cified in the COLR for mc e than I hour of cumulative penalty deviation time during 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 RE0VIREMENTS 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 assur.ed 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 regaired target band shall be accumulated on a time basis of:

a. One-minute penalty deviation for each 1 minute of POWER OPERATION 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 mint.te 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 target 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 interpolatica between the most recently measured value and the predicted value at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.

- (Continued) required target band specified in the COLR during testing without penalty deviation.

SEABROOK - UNIT 1 3/4 2-2 Amendment No.

.. , t i

?

)

i i

i t

I i

h I

i t

t

)

l t

i PAGE INTENTIONALLY BLANK l i

t i

f i

r I

L v

i i

,f i

SEABROOK - UNIT )- 3/4 2-3 Amendment No.

_~ _ . _ . ._..=_.s._.. .

F0WER DISTRIBUTION LlHITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR F9 f Z)

LlHITING (!)NDITION FOR OPERATION ,

3.2.2 Fn(Z) shall oo limited by the following relationships:

Fn (Z) i F"g" K(Z) for P > 0.5 7

Fn(Z) i F"qV K(Z) for P 10.5

.5 Where: P= THERMAL POWER , and RAILD THERMAL POWER l 1

F"g" =

the F 9limit at RATED THERMAL POWER (RTP) specie'ied in the CORE OPERATING LIMITS REPORT (COLR), and K(Z) - the normalized F Z) as a function af core height as specifl(ed in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With Fq (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% nF (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% Fn(Z) exceeds the limit, and

b. Identify and correct the cause of the out-of-limit condition 3rior to increasing THERMAL POWER above the reduced limit required )y ACTION a., above; THERMAL POWER may then be increased, provided F (Z)-is demonstrated through incore mapping to be within its llmit.

SEABROOK - UNIT l- 3/4 2-4 Amendment No.

l

1 6

l l

PAGE INTENTIONALLY BLANK 1

l l

[ ]

I: .

1 l- ;

[

f' i

i i

I t'

i l

l SEABROOK - UNIT'l 3/4 2-5 Amend % nt No.-

l-i:- - . . ~ . , .,..._,__.,._.____.,__,__.___,--u - . . - - _ . _

POWER DISTRIBUTION,1jjijl$,

HEAT FLUX HOT CHANNEL FACTOR - fnf7) .

i LIMITING COND1110N FOR OPERATION ,

4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 f, shc11 be evaluated to determine if fq(Z) is within its limit by,:

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

b. Increasing the measured f,y component of the power distribution map by 3% to account for manufacturing tolerances and further

, increasing the value by 5% to account for measurement uncertainties.

c. Comparing the f,, computed (f fy) obtained in Specification ;

4.2.2.2b., above, to:

l

-l

1) limitsforRATEDTHERMALPOWER(f"[}forthe The approprf,,iate measured core planes given in l Specification '

4.2.2.2e, and f. , below, and '

2) The relationship: '

f ,', - f"ll [l+Pf,,(1 P)),

Where f[y.is the limit for fractional THERMAL POWER '

l is the Power i operation expressed as a function specified in of thef"J

COL , Pf,E and P is the :

ffraction actorofMultiplier RATED THERMAfor f,'L POWER at which T., was measured.

d. Remeasuring f,, according to the following schedule:

L 1) When f f is greater than the f"l,' limit for the appropriate !

L measured core plane but less than the f relationship, additional power distribution maps shalfybetakenandff, i comparedtof"ly"andf,y either:

L a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL POWER or greater 'the THERMAL POWER at which f ly was last determined, or :

b) - At least once per 31 Effective full-Power Days (EfPD),

whichever occurs first.

4 i

l l- SEABROOK UNIT 1 3/4 2-6

Amendment No.

l

f.0WER DISTRIBUT10N tIMITS ,

HEAT FLUX HOT CHANNEL FACTOR - in(Z)

SVRVElllANCE RE0VIREMENIS 4.2.2.2d. (Continued)

2) When the Ff, is less than or equal to the F7l11mit for the appropriete mepured core plane, additional power distribution maps shall be taken and F f, compared to F",",' and Ff,atleastonceper31EFPD.

e.

The F,, planes all core limits for RATEDBank containing THERMAL*D" contro POWER (F7ll rods and all unrodded core planes in the CORE OPERATING LIMITS REPORT per Specification 6.8.1.61

f. The F limits of Specification 4.2.2.20., abo /e, are not appliEable in the following core planes regions as measured i.1 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 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%, and 74.9 1 2%, inclusive, and

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

g.

With F f exceeding evaluated Ff,, ifthe to determine effects Fq (Z) of f'n' its fimits.on F (Z) shall be is withi 4.2.2.3 When fo(Z)-is measured for other than F, determinations, an overall measured Fo(Z) shall be obtained from a power dis,tribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

SEABROOK - UNIT 1 3/427 Amendment No.

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

POWER DISTRIBUTION LlHITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEt FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 F5 shall be less than F"TL [1.0 + Pfa (1-P)).

Where: P= THERMAL DOWER , and RAHD 1HERMAL POWER

=

F"75 theF5limitatRATEDTHERMALPOWER (RTP) s pecified in the CORE-OPERAT11G LIMITS REPORT (COLR), and Pfa - thePowerfactorMultiplierforF5 specified in the COLR.

APPLICABillTY: MODE 1.

ACTION:

WithF5exceedingitslimit:

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-THERHAL POWER above the limit required b ACTION a.,

above; THERMAL. POWER may then be increased, provided F i is demonstrated through incore mapping to be within its 1 mit.

SUPVEILLANCE REQUIREMENTS

~4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F5 shall be demonstrated to be within its limit prior to operation above 75% RATED 1HERMAL POWER after e3ch 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.

b. Using the measured value of F5 which does not include an allowance for measurement uncertainty.

h.

SEABROOK - UNIT 1 3/4 2 8 Amendment No.

j 3/4.1 REACTIVITV CON 1ROL SYSTEMS BASES .

I t

3/4.1.1 B0 RATION CONTROL _

l 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHU1DOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions (2) the reactivity transients !

associated with postulated accident conditions are controllable within !

acceptable limits, and (3) the reactor will be maintained sufficiently :

subtritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of The most restrictive i fuel depletion, condition RCS occurs at boron concentration, E0L, with T., at noand loadRCS T.,in.

operat g temperature, and is !

associated with a postulated stea,m line break accident and resulting :

uncontrolled RCS cooldown. In the analysis of this accident, a minimum l SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) is !

required to control the reactivity transient. Accordingly, the SHUTDOWN !

MARGIN requirement is based upon this limiting condition and is consistent !

with FSAR safety analysis assumptions. With T less than 200* F, the i reactivity transients resulting from a postulat'ed steam line break cooldown f

-are minimal. A SHUTDOWN MARGIN as specified in the COLR and a boron concentration of greater than 2000 ppm are required to permit sufficient time f for the operator to terminate an inadvertent boron dilution event with T,,, i less than 200* F. !

3/4.1.1.3 M@ERATORTEMPERATURECOEFFICIENT l, The limitations on moderator temperature coefficient (MTC) are provided -)

to ensure that the value of this coefficient remains within the limiting ;

condition assumed in the FSAR accident and transient analyses. [

The MTC values of this specification are applicable to a specific set of :

plant conditions; accordingly, verification of MTC values at conditions other ;

than those explicitly stated will require extrapolation to those conditions in i order to. permit an accurate comparison.

~

The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC ,

used in the-FSAR analyses to nominal operating conditions. These corrections :'

involved subtracting the incremental change in the MDC associated with a core

condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with ;

temperature at RATED THERMAL POWER conditions. 1his value of the MDC was then r transformed into the limiting end of cycle life (E0L) MTC value as specified ;

in the COLR. The 300 ppm surveillance limit MTC value as specified in the !

COLR represents a conservative value (with corrections for burnup and soluble ;

boron) at a core condition of 300 ppm equilibrium boron concentration and is !

obtained by making these corrections to the limiting MTC value as specified in !

the COLR.

i SEABROOK - UNIT 1 B 3/4 1-1 [

Amendment No. i l

L 1

REACTIVITY CONTROL SYSTEMS BASES BORAT10N CONTROL 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (Continued)

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due

. principally to the reduction in RCS boron concentration associated with fuel burnup.-

3/4.1.1.4 HINIMUM TEMPERATURE FOR CRITL(AllTY This specification ensures that the reactor will not be made critical

- with the Reactor Coolant System average temperature less than 551' F. This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range,-(3) the pressurizer is capable of being in an OPERABLE sta is with a steam bubble, and (4) the reactor vessel is above its minimum RTnf temperature.

- 3/4.1.2 BORAT10N SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components recuired to perform this functinn include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS in-MODES 1, 2, or 3, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT from expected operating i conditions after xenon decay and cooldown to 200' F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 22,000 gallons of 7000 ppm borated water from the boric acid storage tanks or a minimum contained volume of 477,000 gallons of 2000 ppm borated water from the refueling water storage tank (RWST). ,

The limitation for a maximum of one centrifugal charging pump to be OPERABLE'and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 provides

- assurance that a mass addition pressure transient can be-relieved by operation of a single PORV or an RHR suction relief valve. .

As a result of this, only one boron injection system is available. This

+ is acceptable on the basis of the stable reactivity condition of the reactor, the-emergency power supply requirement for the OPERABLE charging pump and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity i changes in the event the single injection system becomes inoperable.

SEABROOK - UNIT I B 3/4 1-2 Amendment No.

1

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

i REACTIVITY CONTROL SYSTEMS l i

BASES 2/.LL 2 BORAT10N SYSTEMS (Continued)

The boron capability required below 200' F is sufficient to provide a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT-after xenon decay and cooldown from 200* f to 140' f. This condition requires a minimum contained volume of 6500 gallons of 7000 ppm borated water from the boric acid storage tanks or a minimum contained volume of 24,500 gallons of 2000 ppm -

borated water frum the RWST.

2 The contained water volume limits include allowance for water not available because of discharge line location and other physical >

characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

d The OPERABILITY of one Boron injection System uring REFUELING ensures ,

that this system is available for reactivity contr' lile in MODE 6.

The limitations on OPERABillTY of isolation provisions for the Baron Thermal Regeneration System-and the Reactor Water Makeup System in Modes 3, 4, 5, and 6 ensure that the boron dilution flow rates cannot exceed the value assumed in the transient analysis.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHVIDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated .

accident analyses are limited. OPERABILiiY of the control rod position i indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210, and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position _ indication System does not indicate the actual shutdown rod position between 18 steps-and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded-position.

The ACTION state;nents which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

SEABROOK - UNIT 1 B 3/4 1-3 Amendment No, w , , - - -. . - , , - , _ - . - - - - - - - - _ . - - - _ - . _ . . - . - _ _ _

i . .

l 3/4.2 POWER DISTRIBU110N LIMITS j

BASES I

The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and 11 (Incidents of Moderate frequency) i events by: (1) maintaining the minimum DNBR in the core greater than or equal i to 1.30 during normal operation and in short-term transients, and (2) limiting j the fission gas release, fuel pellet temperature, and cladding mechanical

[ properties to within assumed design criteria. In addition, limiting the peak i linear power density during Condition i events provides assurance that the

! initial conditions assumed for the LOCA analyses are met and the ECCS

acceptance criteria limit of 2200* f is not exceeded. ,

I f

) The definitions of certain hot-channel cnd peaking factors as used in

these specifications are as follows:

l fq(Z) Heat Flux Hot Channel factor, is defined as the maximum local heat j flux on the surface of a fuel rod at core elevation Z divided by the i average fuel rod heat flux, allowing for manufacturing tolerances on i fuel pellets and rods; 1

F ",,n Nuclear Enthalpy Rise Hot Channel factor, is defined as the ratio of j the integral of linear power along the rod with the highest 1 integrated power to the average rod power; and F,,(Z ) Radial Peaking factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

~

j

3/4.2.1 AXI Al FLUX _J)1FFERENCE i The limits on AXIAL FLUX DIFFERENCC (AFD) assure that the Fn (Z) upper

, bound envelope of the fn limit specified in the CORE OPERATING LihlTS REPORT

(COLR) times the normalized axial peaking factor is not exceeded during either l- -normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED 1HERMAL POWER is the target flux dif ference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level- . The-periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

Although.it is intendcd that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AfD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution SEABROOK - UNIT 1 B 3/4 2-1 Amendment'No.

POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AX1AL FLUX DIFFERENCE (Continued) sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within established limits while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ' actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AfD Monitor Alarm. The computer determines-the 1-minute average of each of the OPERABLE excore detector

, outputs and provides an alarm message immediately if the AFD for two or more

OPERABLE excore channels are outside the target band and the THERMAL POWER is j greater than_90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

' 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY-RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA, the peak

fuel clad temperature will not exceed the 2200* F ECCS acceptance criteria i limit.

Each of these is measurable but will normally only be determined periodically as specif_ied in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control-rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position; i b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

SEABROOK - UNIT 1 B 3/4 2-2 Amendment No.

, ,,- ,.y-.._ %. .. .____,._.__.m y. , _ , _ , . . . . , ,m ,_,._,.m___-.ym .wy._. , ,.,m.,_.,v.,.my.y%c__-_.,,y ,, .-..---.w

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY R11E HOT DJANNEL FACTOR (Continued) i fin will be maintained within its limits provided Conditions a. through

d. above are m;.intained. The relaxation of Fin as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion l limits.

Fuel rod bowing reduces the value of DNBR. Credit is available to l offset this reduction in the generic margin. The generic margins, totaling 9.1% DNBR completely offset any rod bow penalties. This margin includes the following:

a. P sign limit DNBR of 1.30 vs. 1.28,

b. Grid spacing (Ks) of 0.046 vs. 0.059, i

c. Thermal diffusion coefficient of 0.038 vs. 0.059,

d. DNBR multiplier of 0.86 vs. 0.88, and

e. Pitch reduction.

p The applicable values of rod bow penalties are referenced in the FSAR.

When an Fa measurement is taken, an allowance for both experimental i

error and manufacturing tolerance must be made. An allowance of 5% is l appropriate for a full-core map taken with the Incore Detector Flux Mapping .

j System, and a 3% allowtnce is appropriate for manufacturing tolerance. '

i The Radial Peaking factor, F,y(Z), is measured periodically to provide :

assurance that the Hot Channel Facto Fq (Z), remains within its limit. The L F,, limit for RATED THERMAL POWER (F" ) as provided in the CORE OPERATING -

l LI'MITS REPORT per Specification 6.8. 6 was determined from expected power ,

control maneuvers over the full range of burnup conditions in the core.

_l When RCS- F$n is measured, no additional allowances are necessary prior ,

l to comparison with the established limit of a measurement error of 4% for Fin i I

has-been allowed for in determination of the design DNBR value. ;

3/4.2.4 OVADRANT POWER TILT RATIO l

The purpose of this specification is to detect gross changes in core power distribution between monthly incore . flux maps. During normal operation l the QUADRANT POWER TILT RATIO is set equal to zero once acceptability of core ;

L peaking factors has been established by review of incore maps. The limit of -t i 1.02 is established as an indication that the power distribution has changed l enough to warrant further investication. j i

.i l

SEABROOK - UNIT 1 B 3/4 2-3 l Amendment No.

l t

- - - - ,. - - - - . , . . - - . . . -- - .~_.________-__- - -____ -_._.__-____..____ _ -_--._- _ - _ -

@jjulSTRATIVE CONTROLS _ , . . ,

},13LANNVAL RADI0 ACTIVE Eff tVENT BELEASE REPQ31 6.8.1.4 (Continued) to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculuing the dose contribution from liquid and gaseous offluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRIC1ED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

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

The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.9 or 3.3.3.10, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks ext ding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORTS-6,8.1.5 Routine reports of operating statist 4cs and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Washington, 0.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional _ Administrator, no later than the 15th of each month following the ,

calendar month covered by the report.

~

CORE OPERATING LIMITS R EQEl 6.8.1.6.a Core operating limits shall be established and documented in the COPE OPERATING LIMITS REPORT prior to each reload cycle, or prior-to any remaining portion of a reload cycle, for the following:

1. SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3.1.1.1,

L 2. SHUTDOWN MARGIN limit for MODE 5 for Specification 3.1.1.2, i

L 3. Moderator Temperature Coefficient BOL and E0L limits, and 300 ppm

. surveillance limit for Specification 3.1.1.3, I

SEABROOK - UNIT 1 o-18 '

Amendment No.

I b -_- - , _ ,--- - - _ _ . ., _ _ _ ._. _.__ _ .... m.__ _ _.. _ __ _ . _ . _ . . _ . _ . . _ _ . . . . . _ _ . . _ _ _ _ _ . . -

ADMINISTRATIVE CONTROLS 6.8.1.6.a.(Continued)

4. Shutdown Rod Insertion limit for Specification 3.1.3.5, ;

5. Control Rod Bank Insertion limits for Specification 3.1.3.6, i

6. AX1AL FLUX DIFFERENCE limits and target bar.d for Specification 3.2.1,

7. and the Power Heat FactorFlux Hot Channel Multiplier Far.

for F,y for tor, f *i[, K(Z), F""3.,2.2, Specification ,

8. and the Power Nuclear Enthalpyfor Factor-Multiplier Rise F5aHot for Channel Specification f actor,.2.3. F""$u .

The CORE OPERATING LIMITS REPORT shall be maintained available in the Control Room.

6.8.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1. WCAP 9272 P A, " Westinghouse Reload Safety Evaluation Methodology" July 1985 (H Proprietary)

Methodology for Specifications:

3.1.1.1 - SHUTDOWN MARGIN limit for MODES 1, 2, 3 and 4 3.1.1.2 - SHUTDOWN MARGIN limit for MODE 5 1.1.1.3 - Moderator Temperature Coefficient 3.1.3.5 - Shutdown Rod Bank Insertion Limit 3.1.3.6 - Control Rod Bank Insertion Limits 3.2.3 - Nuclear Enthalpy Rise Hot Channel f actor

2. WCAP-ll596-P A, "Qualificatic,n of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores" June 1988 (g Proprietary)

Methodology for Specifications:

3.1.1.1 - SHUTDOWN MARGIN limit for MODES 1, 2, 3 and 4 3.1.1.2 - SHUTDOWN MARGIN limit for MODE 5 3.1.1.3 - Moderator Temperature Coefficient

3. WCAP-8385-P-A, " Power Distribution Control and Load Following Procedures Topical Report", September 1974 (H Proprietary)

Methodology for Specifications:

3.1.3.5 - Shutdown Rod Bank Insertion Limit 3.1.3.6 - Control Rod Bank Insertion Limits

! 3.2.1 - AX1AL FLUX DIFFERENCE

4. WCAP-/811 " Power Distribution Control of Westingbouse Pressurized Water Reactors", December 1971 (H Proprietary) i' l

SEABROOK - UNIT 1 6-18A Amendment No.

ADMINISTRATIVE CONTR01S ,.

6.8.1.6.b. (Continued)

Methodolugy for Specifications:

3.1.3.5 - Shutdown Rod Bank Insertion Limit j 3.1.3.6 Control Rod Bank Insertion Limits 1 1

5. Letter, T.H. Anderson to K. Knell (Chief of Core Performan~ e Cranch,NRC), January 31, 1980,

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package Methodology for Specification:

3.2.1 AX1AL FLUX DlffERENCT

6. NUREG 0800, Standard Review Plan, US Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981, Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial of fset Control (CAOC), Rev. 2, July 1981 Methodology for Specification:

3.2.1 AX1AL FLUX DIFFERENCE

7. WCAP-7300 l, " Evaluation of Nuclear Hot Channel Factor Uncertaintles" December 1971 (M Proprietary)

Methodology for Specification:

3.2.2 - Heat flux Hot Channel Factor

8. WCAP 8622, " Westinghouse ECCS Evaluation .1odel. October,1975 Version", November 1975 (H Proprietary)

Methodology for Specification:

3.2.2 - Heat Flux Hot Channel factor

9. WCAP 9220, " Westinghouse ECCS Evaluation Model, February 1978 Version", February 1978 (H Proprietary)

Methodology for Specification: .

3.2.2 Heat Flux Hot Channel Factor

10. WCAP-7912-P A, " Power Peaking Factors", January 1975 (H Proprietary)

Methodology for Specification:

3,2,3 - Nuclear Enthalpy Rise Hot Channel Factor

11. YAEC-1363 A, "CASM0 3G Validation," April 1988.

YAEC-1659-A, " SIMULATE-3 Validation and Verification," 3cptember-1988.

-SEABROOK - UNIT 1 6 18B-Amendment No.

ADMINISTRAT1yECONTROLS 1 6.8.1.6.b. (Continued)

Methodology for Specifications' 5 3.1.1.1 - SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 ;

3.1.1.2 -- SHUTDOWN MARGIN for MODE 5 '

3.1.1.3 - Moderator Temperav're Coefficient 3.1.3.5 -

Shutdown Rod Bank .nsertion Limit >

3.1.3.6 -

Control Rod Bank Insertion Limits 3.2.1 -

AX1AL FLUX DlffERENCE 3.2.2 -

Heat Flux Hot Channel factor !

3.2.3 - Nuclear Enthalpy Rise Hot Channel f actor

12. Seabrook Station Updated Final Safety Analysis Report Section :

15.4.6, " Chemical and Volume Control System Malfunction That i Results in a Decrease in the Boron Concentration in the Reactor Coolant System". !

Methodology for Specifications: i 3.1.1.1 - SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 ;

3.1.1.2 -

SHUTDOWN MARGIN for MODE 5 l

6.8.1.6.c. The core operating limits shall be dotermined so that all ;

a)plicable limits (e.g., fuel thermal mechanical limits, core ;

tiermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. I The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid- l cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and r the Resident inspector.

[

i i

t f

i l i i

l.

i I

L !

L 6-18C i SEABROOK - UNIT 1 Amendment No.

l l

IV. Safety Evaluation of Proposed Chanaes I The current Technical Specification method of controlling reactor physics parameters to ensure confonnance to 10 CFR 50.36 (which requires the lowest functional performance levels acceptable for continued safe operation) is to specify the values determined to be within the acceptable criteria using an NRC-approved methodology for calculating the parameter. The methodologies for determining these cycle-specific parameter limits specified in proposed Technical Specification Section 6.8.1.6 have been reviewed and approved by the NRC and are consistent i with the-applicable analyses presented in the Updated Final Safety .

Analysis Report (FSAR).

lhe removal of cycle specific parameter limits from the Technical Soecifications has no impact upon plant operation or safety. No ,

safety related equipment, safety function, or plant o)erations will be-altered as a result of this proposed change. Since tie applicable FSAR i limits will be maintained and the proposed Technical Specifications will !

require operation within the cycle-specific parameter limits determined by _specifically identified NRC-approved methodologies, this proposed ,

change-is administrative in nature. The required actions to be taken if limits are violated are not affected by the proposed Technical Specification changes nor are the required surveillance requirements.

l The proposed Technical Specification changes will control the cycle specific parameters within the acceptance criteria and assure conformance-to 10 CFR 50.36 by using the approved methodology instead of sper.ifyi'ig Technical Specification values. The COLR will document the

specific parameter limits resulting from cycle design calculations, including mid-cycle or other revisions to parameter values. Therefore, the proposad d ange is in conformance with the requirements of 10 CFR >

50.36..

Any changes to the COLR will be made in accordance with the provisions of 10 CFR 50.59. From cycle to cycle, the COLR will be revised such that the appropriate cycle specific parameter limits for ,

the applicable cycle will a) ply. Technical Specifications will not be '

~

changed, thus eliminating tie associated administrative burden.

8

. . . . . - . - . . - - - - - ~ - - - - - - - . - - - - - -

I 4

f V. Determination of Sionificant Hazards for Proposed Chanaes ;

Pursuant to 10 CFR 50.91, New Hampshire Yankee has determined that operation of the facility in accordance with the proposed License Amendment Request does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR 50.92. The following discussion describes how the proposed amendment satisfies each 4

of the three standards of 10 CFR 50.92(c).

1) The proposed change does not involve a significant increase in the probability or consequences of an accident previcasly evaluated. ,

The removal of cycle specific parameter limits from the Seabrook Station Technical Specifications has no influence or impact on the 1 probability or consequences of any accident previously evaluated.

The cycle specific parameter limits, although not in Technical Specifications, will be followed in the operation of Seabrook Station. The proposed Technical Specification changes do not ,

effect the actions which must be taken if the cycle-specific

< parameter limits are exceeded nor do they affect the surveillance i reouirements associated with these limits.

Future core reload designs will be supported by a 10 CFR 50.59 evaluation which will examine each accident analysis addressed in i the Updated Final Safety Analysis Report (FSAR) with respect to .

, changes in cycle-specific parameter limits to ensure that the

. reload design itbounded by previously accepted analyses. This examination - which will be performed per the requirements of 10 CfR 50.59, ensures that future core reload designs will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2) Tho' proposed change does not create the possibility of a new or

- a different kind of accident from any accident previously evaluated.

As stated above the removal of the cycle specific parameter limits has no-influence or impact, nor does it contribute in any way to the probability or consequences of an accident. No safety related equipment, safety function, or plant operations will be. altered as a result of this proposed change. The cycle specific parameter limits are calculated using the.NRC-approved methods and submitted -

to the NRC to allow the Staff to continue to trend the values of these limits. The proposed Technical Specifications will require operation within the cycle-specific parameter limits. The proposed Technical Specification changes do not affect the actions which must be taken if the cycle-specific parameter limits are exceeded nor do they affect the surveillance requirements associated with these limits.

Therefore,' the proposed amendment does not in any way create the 4 possibility of a new or dif ferent kind of accident from any accident previously evaltated.

9 i

s-

-v.. ~ 4. -, , y,-. ,_,m J. .m., . . . . . . ,, ,_.r,_. _ ., , . _ _ e._.,sm,y .,,.,m,. mmc.,.,kg.,,__,m, , , . , .

3) The proposed amendment does not result in a significant' reduction in the margin of safety.

The margin of safety is not affected by the removal of 4 cycle-specific parameter limits from the Technical Specifications.

The margin of safety provided by current Technical Specifications remaint unchanged.- Surveillance requirements exist to monitor the values of these cycle-specific parameter limits. The proposed Technical Specification changes require operation within the .

cycle-specific parameter limits which are determined using :

NRC-approved reload design methodologies.

The development of the cycle specific parameter limits for future reload designs will continue to utilize only those methods described in NRC-approved documentation and identified in aroposed Technical Specification Section 6.8.1.6. In addition, eac1 future i

reload design will involve a 10 CFR 50.59 safety evaluation to assure that operation of the unit within the-cycle-specific parameter limits will not involve a significant reduction in a margin of safety.

Therefore..the )roposed changes are administrative in nature and

, do not impact tie operation of the Seabrook Station in a manner that involves a reduction in the margin of safety.

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists. This guidance (51 FR 7750) includes examples of the type of amendments.that are considered not likely to involve significant hazards considerations. The change-proposed is similar to the examples of administrative changes identified in 51 FR 7750. ,

Additionally, the proposed change is-consistent with the NRC policy for improving technical specifications (52 Fk 3788) and the proposed change '

is consistent with 10 CFR 50.36.

In view of the preceding, New Hampshire Yankee has determined that the proposed License Amendment Request does not involve any significant-hazards considerations.

10 v - -+ -e - - --4., ,- ------m n-,,,,,y,, ,.. . , .-,..,.w, ,_,,m--m-e-,rew wa , - . , -- ,

Vll. Other Sjipnortina Documentation See attached: ,

- DRAFT CORE OPERATING LIMITS REPORT, SEADROOK STA110N UNIT 1, CYCLE 2 1 i

r i

k i

l l

i i

i e

11 1

- .-- =-. .. . __ . _ . _ _

v b

I f

P DRAFT CORE OPERATING LIMITS REPORT SEAllROOK STATION UNIT 1 i CYCLE 2 t

I e

s 9

I -

l f

}

I l

l

o. .

1 1.0 EDJE OPERATI,NMIMITS REPORT This Core Operating Limits Report for Seabrook Station Unit 1 Cycle 2 has been prepared in accordance with the requirements of Technical Specification 6.8.L6.

The Technical Specifications affected by this report are:

f 1 3.1.1.1 SHUTDOWN MARGIN limit for MODES 1, 2, 3, 4 l 2 3.1.1.2 SilUTDOWN MARGIN limit for MODE $ !

3 3.1.1.3 Moderator Temperature Coefficient j 4 3.1.3.5 Shutdown Rod lasettion Limit >

$ 3.1.3.6 Control Rod Insertion Limits t 6 3.2.1 AXIAL FLUX DIFFERENCE 7 3.2.2 Heat Flux ilot Channel Factor 8 3.2.3 Nuclear Enthalpy Rise flot Channel Factor

-i 2.0 pFERATING LIMITS f

The cycle specific parameter limits for the specifications listed in section 1.0 are presented .

In the following subsections.. These !!mits have been developed using the NRC approved .

methodologies specified in Technical Specification- 6.8.16. i k

. 2.1 _ S,j,1UTDOWN M ARGIN LIMIT FOR MODES 1. 2. 3. AND 4 (Specification 3.1.1.1; .

The SHUTDOWN MARGIN shall be greater than or equal to 1.3% delta k/k. l

' 2.2 SHUTDOWN MARGIN LIMIT FOR MODE 5 (Specification 3.1.1.2) l The SHUTDOWN-MARGIN shall be greater than or equal to 1.2% delta k/k. _i

' 2.3 . MODERATOR TEMPERATURE COEFFICIENT (Specification 3.1.1.3) 2.3.1 The Moderator. Temperature Coefficient (MTC) shall be less positive than 0 l delta -k/k/'F for Beginning- of Cycle Life-(BOL). All Rods Out (ARO), liot ~

Zero Thermd Power conditions,- j 2.3.2 MTC shall be less negative than 4.2 x 10 delta k/k/'F for End of Cycle Life f (EOL),- ARO, Rated Thermal Power conditions. .

2.3.3 - The 300 ppm ARO, Rated Thermal Power MTC shall be ! css negative than I 3.3 x 10~ delta k/k/'F (300 ppm Surveillance Limit) _

i

- 2.4 SHUTDOWN ROD INSERTION LIMIT (Specification 3.1.3.5) {

t.4.1 The shutdown rods shall be fully withdran, f i

25 CONTROL ROD INSERTION LIMITS (Specification 3.1.3.6) j

-2.5.1 .The control rod banks shall be limited in physh:al insertion as specified in !

Figure 1. ;

. I i; ' 2.6 -

AXIAL FLUX DIFFERENCE (Specification 3.2.1) -

'I

,' 5

-2.6.1 The AXIAL FLUX DIFFERENCE (AFD) Target Band is + 3%,12%. -!

l 2.6 .* The AFD .shall. be maintained within the Acceptable Operation Limits as specified in Figure 2. ';

l l

h L ;

y :

,.,J. . , _ . _ . - . . a m._ . _.-, _,-._..,,._..m.--.__..__ -

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

. _. _m_ _ _ _ .

2.7 llIl&T Fl.lf L((DT Cl[6fSF1 FA!' TOR (Specification 3.2.2) 2.7.1 Fo* - 2.32 2.7.2 K(Z) is specified in Figure 3.

2.7.3 P F , = 0.2 2.7.4 Tbc F., limits for Rated Therma' Powc ..it!.;c specific core planes shall be:

2.7.4.1 F., (RTP) less than or equal to 1384 for all planes contain!nu banks D 4 C control rods for cycle burnups from 0 3000 MWD!MTU; 2.7.4.2 F,, (RTP control ro)ds from cycle burnup 3000 MWD /MTU onward;less than or equal to 1.83 2.7.4.3 F ,ds; and(RTP) less than or equal to 1.784 for all planes containing bank D control ro 2.7.4.4 F,, (RTP) less than or equal to 1.622 for all unrodded planes.

2.7.4.5 See Figure 4 for a plot of FQ(Z)*P(REL) versus axial core height.

2.8 FUCI.E AR ENTil ALPY HISE llOT Cil ANNEl, FACTOR _ (Specification 3.2.3) 2.8.1 F,,n, n* = 1.6 2.8.2 P F.,,,, , = 0.2 y

k

'8 ,, ,_ _ _ .. ,, ,_ - . _ __

, ~, w

O Se sL <,c h S+a w, m ,'/ .I f,7 , /

C cre Cr <a n g L h +t IGf er t'

,.,_ C y ' I' L

V' (0.30,228) (0.844,228) 228

/ \

/

/ /

200 / BANK B ,/

g ! / I I / i 2

< '/ l

/ -

i

/ rI i

C i .

/ l l E

H 160

/(0.0.164) l /'

3 l i j l

l l

i l/ i I

i

! ! I !

m .i a : : ; i

-(1.0.146. )/-

b

~

i !

l I I II, ly

(/'ANK C !//

D (20 /

o I! If l i

I fl I / I fi j t' l / ! / \

4 l / I I I y i !

~

/ I

l 1 i v'l ! l l E fl l l l 1 / I i E 2fg,g',44) i I

/ BANK D Il _

g 4, _

l I t l I /i ' '

I e i If l

/

/ (0.31.0.0) 0 I l/- ! l ! l l 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER

, iIOU R C --3 rl-1--

R00 BANK INSERTION LIMITS VERSUS THERMAL POWER

_g. FOUR-LOOP OPERATION

$ e a l., ,,,k sta +s un , y- ) ,C z, (or t (),o a t er Yo h3 j in< i f.s def rei Cv c/c

2. ;

Oj %$i.Y 120 --

8I i

Eh i

I e

UNACCEPTABl.E bPERAT10N 1

' UNACCEPTABLE OPERATION l j: rn.9ci l ai.hei l I g i i / \

i 2 i / \ j E ** '

I 1/1 1 !\' '

g j .

/

i l

\ !

I CC w

[l dCCEPTdBLE OP' ERA, TION k

E" /I \

o-l

/

/ i l\

w I. >

\ L

t. H g*-r. <E l-31.50) l (31.5 0) l C

l l l l l l l e '" Ii I! l i i i

,s. i i l l lI i i I I I 20- i l l ,! .

l l l l' ,

0

-50 -40 30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE (oI)%

-f!CU"E 3.F1

.:5h AXIAL FLUX OIFFERENCE LIMITS AS A FUNCTION OF E L' RATED THERMAL POWER 1 - di n gy)6 ,4 s

s t a ti;, -

u.,, t j R.5 9 S e & < < ol<

[ o r e. Ope r a f n y l-- i % f ' Of

W li

.. d yc/c A-1.2 I !II ! I i

j l i I l

! I i l 1 I i !i I i I hs.e.i.e> l I i l ! ! I 1.0 l l l l !I I I ! I I i 4 'ed*d i ! ! l l ! l l ! i '

l I l ! I i ! , I I { l f j i i l i i i l i i il 1 5 ! l i i i ; i i l I i -

\

f i ! -l I , 1 i ; ; i i c3 ,

I I 1, I I l I ! l l i

W 0.6 -

i l l , I , i I ; i s.e.vs>

' ' ' 3

_J I ! l i ! I i

, l l } l i ! i i l l l l l ! l !

CC l l l l i 9

~

! l $ l t

I l !

l I l ! i I : I 1 1.

t t

i e.e l i I

Q l l ! l l l 1 i i i i s l II l l l l II lI '

i I lil l I ;

I !! !I i C.2 l ! -l , l l Ii i lI

! l l l l

. .I I II i ! I I i l I l i ! l ! I I I I ! .

e.e a 2 4 6 8 to 12 CORE HEIGHT (FT)

-i!OURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTIO'. OF CORE HEIGHT 9

. . _ . . . , . . . . - - . 9' ..

M

([

Q. , . - . _ - . _

.,.,..A.,,.

R) '/

Sxalc.,8 S1ation un it J

c e r e C.n.ratiy l >>,th M p, f '

Q y . Y 4= A mM ,,. .. (i W vna c=Tg--

2.4 t

. . . . _ . _ . . . ..... 2,3. .. . - . . - 4 . t. 4 . _ . _ . ._ . . .. _ __ . ..._ _ - . . . . . . _ . _ - . . . . . . . . .- . . .

+

21 - + +

2.1 -

+

+ ,+++++++++++++

2- + +++++++++9++++ ++

+

1.9 -

0 +

& 1.8-L e

0 1.7 - +

h.

+

u 1.6-E-

1.5 -

E i 1.4- 4 2

1.3 -

1.2 -+

1.1 -

l

- +

1 l'

O.i l

l l l l l !

l l l ~l 4 6 8 to 12 o 2

\

, - ud + k fq l

VI. Et2E2. sed Schedule for license AmendmenLi;_tyAnce and Effectivenest New llampshire Yankee requests NRC approval and issuance of this License Amendment Request by December 15, 1991. Within 30 days of NRC issuance of the License Amendment, NHY will submit the CORE OPERATING LIMITS REPORT for Seabrook Station Unit 1, Cycle 2 to the NRC and will concurrently implement the Technical Specification changes authorized by the License Amendment. New Hampshire Yankee therefore proposes that the License Amendment become immediately effective upon issuance with implementation required within 30 days of issuance.

The schedule proposed by New llampshire Yankee will eliminate the necessity to submit a license Amendment Request for potential changes to cycle specific parameter limits in Cycle 3 and future operating cycles.

12

ML20082P554

Contents

Text

Attachment:

3.2.1 ACTION

Attachment:

Navigation menu

ML20082P554

Text

Attachment:

3.2.1 ACTION

Attachment:

Navigation menu

Search