ML20059H638

Proposed Tech Specs,Changing Administrative Controls in Section 6.0 to Redefine Composition of Sorc,Create Station Qualified Reviewer Program,Delete Requirement to Perform Periodic Procedure Reviews & Revise TS 6.4.2.9

ML20059H638
Person / Time
Site:	Seabrook
Issue date:	01/14/1994
From:	NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:
Shared Package
ML20059H630	List: ML20059H627 ML20059H632 ML20059H638
References
NUDOCS 9401310062
Download: ML20059H638 (67)

Text

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i fNDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0blREMENTS _

SECTION

.P.A_G_E_

3/4.9.4 CONTAINMENT BUIi. DING PENETRAT )

3/4.9.5 COMMUNICA TIONS. . . . . . . . . . . . . . l0 N S . . . . 3/4

... . . .9-4 3/4.9.6 REFUELING MACHINE........................................ 3/4 9-5 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE

................... ..... 3/4 9-6 3/4.9.8 AREAS............ ..... 3/4 9-7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION 3 H i g h Wa t e r L e v e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Low Water 3/4 9-8 i

i 3/4.9.9 Level.......................................... 3/4 9-9 L CONTAINMENT PURGE AND EXHAUST ISOLATION SY 3/4.9.10 3/4 9-10 WATER LEVEL - REACTOR VESSEL.............. STEM...... 3/4 9 ....

3/4.9.11 WATER LEVEL - STORAGE POO L . . . . . . ............... . . . . . . . . . . . . .3/4

. . . .9-12 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING 3/4.9.13 3/4 9-13 SYSTEM......

FIGURE 3.9-1 SPENT FUEL ASSEMBLY ST0 RAGE..............................3/4 9-16 FUEL ASSEMBLY BURNUP VS INITIAL ENRICHMENT.........

FOR SPENT FUEL ASSEMBLY STORAGE.................. .

3/4.9.14 NEW FUEL ASSEM3LY ST0 RAGE.............. 3/4 9-17

................. 3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN..........................................

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-1 l 3/4.10. .) PHYSICS 3/4 10-2 3/4.10.4 TESTS............................................

REACTOR COOLANT LOOPS........ 3/4 10-3 m

........................... 3/4 10-4 3/4 10-5 3/4 10.LiOSITIOk INDICATION SYSTEM - SHUT 00WN. .

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--w . ./. ./ ,., ,c. ./. .x.- .n,3 /A ',10- 5 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration......... .

Dose.................

Liquid Radwaste Treatment System........

3/4 11-1 3/4 11-2 Liquid Holdup Tanks....... ................ ... 3/4 11-3

..................... 3/4 11-4 3/4.11.2 GASEOUS EFFL3ENTS Dose Rate................................................

Dose - Noble Gases......... ............................. 3/4 11-5 Dose - Iodine-131, Iodine-133, Tritium and Ra 3/4 11-6 Material in Particulate Form. . . . . . . . . ,. . . . . . dioactive .......... 3/4 11-7 Gaseous Radwaste Treatment System....................... 3/4 11-8 Exd osive Gas Mixture - System........................... 3/4 11-9 3/4.11.3 SOLIO RADI0 ACTIVE WASTES......... 3/4 11-10 3/4.11.4 TOTAL 00SE............................................... ........................ 3/4 11-12 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM. ..... ....... ....................... 3/4 12-1 SEABR30K - UNIT 1 ix Amendment No. 6 e

9401310062 940114 f PDR ADOCK 05000447 e p PDP b

g 6.0 ADM?NISTRATIVE CONTROLS SECTION fiqi 6,4 REVIEW AND AUD1I . . . . . . . . . . . . . . . . . . . . . . 6-6 6.A.1 STATION OPERATION REVIEW COMMITTEE (SORC)

Function . . . . . . . . . . . . . . . . . . . . . . . . 6-6 Composition ...................... 6-6 Alternates . . . . . . . . . . . . . . . . . . . . . . . 6-6 Meeting Frequency ................... 6-6 Quorum . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 Respons Dilities . . . . . . . . . . . . . . . . . . . . 6-6 Records ........................ 6-8 6.4./3 NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC)

Function . . . . . . . . . . . . . . . . . . . . . . . . 6-8 +

Composition ...................... 6-84 Alternates . . . . . . . . . . . . . . . . . . . . . . . 6-g 'f Consultants ...................... 6-g?

Meeting Frequency ................... 6-9 Quorum . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 Review . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 Audits . . . . . . . . . . . . . . . . . . . . . . . . . 6-9' to Records ........................ 6-11 6,5 REPORTABLE EVENT ACTION .................. 6-Il .

6.6 SAFETY LIMIT VIOLATION . . . . . . . . . . . . . . . . . . . 6-11 6.7 PROCEDURES AND PROGRAMS .................. 5-12 6.8 REPORTING RE_0VIREMENTS 6.8.1 ROUTINE REPORTS .................... 6-14 Startup Report . . . . . . . . . . . . . . . . . . . . . 6-14 Annual Report s . . . . . . . . . . . . . . . . . . . . . 6-15 Annual Radiological Environmental Operating Report . . . 5-15 Annual Radioactive Effluent Release Report . . . . . . . 6-17 Monthly Operating Reports ............... 5-18 CORE OPERATING LIMITS REPORT . . . . . . . . . . . . . . 6-18 6.8.2 SPECIAL REPORTS .................... 5-19 6.9 RECORD RETENTION , , . . . . . . . . . . . . . . . . . . . 6-19 6.10 RADIATION PROTECTION PROGRAM ............... 6-20 4.4.2 5'T?$7-tCM Qum.l#F-.D hshR Woc.m . . . . . .(- 8.  !

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v .n.waa en munwasarea s-a SEABROOK - UNIT 1 xiv Amendment No. 9,22 I i

DEFINITIONS PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL PROGRAM (PCP) shall contain the current for>ulas, j sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State Regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste.

PURGE - PURGING 1.26 PURGE or PURGING shall be any controlled process of discharging air or-gas l 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.

QUADRANT POWER TILT RATIO 1.27 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore (

detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to lq the reactor coolant of 3411 MWt. )

REACTOR TRIP SYSTEM RESPONSE TIME 1.29 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor l/l until loss of stationary gripper coil voltage.

REPORTABLE EVENT .

9 1.30 A REPORTABLE EVENT shall be any of those conditions specified in !j Section 50.73 of 10 CFR Part 50.

CONTAINMENT ENCLOSURE BUILDING INTEGRITY 1.31 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.

Qw.my de w nD The Containment Enclosure hat 4en system i~s~0PERABLE, and

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

SEABROOK - UNIT 1 1-5 Amendment No. J, 9

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LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide flow. core protection against DNB as a result of complete loss of forced coolant The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous,t breakers shall not exceed (1.2@'teconds.of two or more the For underfrequency, reactor delaycoolant is set pump bus c so that the time required T6r a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second. ,

On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pumo Bus trips are automatically blocked by P-7 (a power level of approximately 10%

of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, the Undervoltage and Underfrequency by P-7. Reactor Coolant Pump Bus trips are reinstated automatically Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 20% of RATED THERMAL POWER); and on increasing power, the Reactor trip from the Turbine trip is reinstated automatically by P-9.'

Safety Injection Inout from ESF f

If a Reactor trip has not already been generated by the Reactor Trip  ;

System instrumentation, the ESF automatic actuation logic channels will initiate i

a Reactor trip upon any signal which initiates a Safety Injection. The ESF i

instrumentation channels that initiate a Safety Injection signal are shown in I Table 3.3-3. ,

Reactor Trip System Interincks The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source Range trip). On dec-reasing power, Source Range Level trips are automatically reactivated and high voltage is restored.

SEABROOK - UNIT 1 B 2-7

TABLE 4. 3-1 (Continued)

TABLE NOTATIONS (Continued)

(12) Hember not used.

[ ) The TRIP ACTUATING DEVICE OPERATIONAL lEST shall independently verify I the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify tha OPERABILITY of the Bypass ~ Breaker trip circuit (s).

(14) local manual shunt trip prior to placing breaker in service. t (15) Automatic undervoltage trip.

(16) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS.

(17) These channels also provide inputs to ESFAS. Comply with the applicable MODES and surveillance frequencies of Specification 4.3.2.1 for any por-tion of the ....

channel required to be OPERABLE by Specification 3.3.2.

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i SEABROOK - UNIT 1 3/4 3-13 Amendment No.17

INSTRUMENTATION MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPEP ..t accident monitoring instrumentation chan-nels less than the Tota, Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The provisions of Speci-fication 3.0.4 are not applicable. -

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b. With the number of OPERABLE acciden nitoring instrumentation __.

channels except the containment high range fe SitTgtt s W monitor, less than the Minimum Channe OPERA 8LE requirements of~

Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDRY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The pro-visions of Specification 3.0.4 are not applicable.

c. With the number of OPERABLE channels for the containment Post-LOCA high range area monitor less than required by the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and either restore the inoperable channel (s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specifi-cation 6.8.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERA 8LE status. .

SURVEILLAIICE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE:

a. Every 31 days by performance of a CHANNEL CHECK, and

b. Every 18 months by performance of a CHAMEL CALIBRATION.

SEABROOK - UNIT 1 3/4 3-49

a REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when the Reactor Trip System breakers are open:"

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,

b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and

d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.

APPLICABILITY: MODE 3. b ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With only one reactor coolant loop in operation and t'he Reactor Trip System breakers in the closed position, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the Reactor Trip System breakers.

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

• All reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.

• • See Specidi Test Exceptien Specific &tien 3.10.G SEABROOK - UNIT 1 3/4 4-2

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CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE REQUIREMENTS t_imlati valve shall be demonstrated OPERABLE during !

sewo@4f3<2_ Ear.h-cantains pA'pl1)4p.ffpo@jAfr JNT at least once per 18 months by:

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" Isolation valve actuates to its isolation position,

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" Isolation valve actuates to its isolation position, and  !

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c. Verifying that on a Containment Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation position.

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4.6.3.3 The isolation time of each power-operated or automatic cont'ainment l

isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

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SEABROOK - UNIT 1 3/4 6-17 Amendment No. 14

p PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a.

One motor-driven emergency feedwater pump, and one startup feedwater pump capable of being powered from an emergency bus and capable of being aligned to the dedicated water vol we in the condensate storage tank,- and b.

One steam turbine-driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.i W ACTION:

a.

With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least within the HOT STANDBY following within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With two emergency feedwater pumps inoperable, restore at least one emergency feedwater pump to OPERA 8LE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and restore both emergency feedwater pumps to OPERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and-in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c.

With one emergency feedwater pump and the startup feedwater pump -

inoperable, restore both emergency feedwater pumps to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and all three pumps to OP. ERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. With three auxiliary feedwater pumps inoperable, immediately initiate i

corrective action to restore at least one auxiliary feedwater pump t to OPERABLE s.tatus as soon as possible.' .

SURVEILLANCE REQUIRD4ENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demenstrated CPERA8LE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1) Verifying that the motor-driven emergency feedwater pump develops a di_scharge pressure of greater than or equal to 1460 psig at a flow of greater than or equal to 270 gpa; *

• Not requiret. in M ; 3 sotil initi;I critic;lity i: ;;hi v 1 SEABROOK - UNIT 1 3/4 7-3 e

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, hECIALTESTEXCEPTIONS 3/4 6 REACTOR COOLANT LOOPS LIMITING CON TION FOR OPERATION l i

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, , 1 3.10.6 The limi performance of testin ns of Specification 3.4.1.2 may be su ended during the involving natural circulation cond tions for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in MODE 3 provi d (1) at least two reactor co ant innps as listed in Specification 3.4.1.2 are ERABLE, (2) nc operatio are permitted that would .

icause dilution of the reacto coolant system boro concentration,'(3) the . '

lreactortripbreakersareopen, nd (4) core ou et temperature is maintained

at least 10*F below saturation te erature.

I fAPPLICABILITY:During performance of ing involving natural circulation jconditions. i

} i IACTION. f i

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!Withlessthantheabovereq red reactor coolant i ormance of testing involv .g natural circulation coops OPERABLE.during f

itions comply withper-the provisions of the ACTIO, statements,of Specification 3. .1. 2.

SURVEILLANCE REQUIR ENTS s  ;

4.10.6 At I the above required reactor coolant loops shall b determined (OPERABLE wi n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to initiation of testing involving na ral circulat n conditions and at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during testing in lying

,natura circulation conditions by verifying indicated power availability nd lbyv ifying secondary side narrow range water level to be greater than or  !

te al to 14%.

9 SEABROOK - UNIT 1 3/4 10-6 i

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REACTIV!TY CONTROL SYSTEMS' 5 4

BASES 3/4.1.2 BORATION 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 from the RWST.

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.  :

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

The limitations on OPERABILITY of isolation provisions for the Boro 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 distri- )

bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and' y (3) the potential effects of rod misalignment on associated. accident analyses.  ;

. are limited. OPERASILITY of the control-rod position indicators is required to -

determine control rod positions'and thereby ensure compliance with the control- 5 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-:

drawn for the Shutdown Banks provides assurances that the Digital-Rod Position Indicator'is operating correctly over the full range of indication. LSince the; Digital Rod Position Indication:Systen'does not indicate-the actual. shutdown rod q position between 18 steps and 210 steps, only points'in-the indicated rangesu are picked for verification of-agreement with demanded position.-

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' rod requires measurements of peaking factors and a restriction'in THERMAL POWER. :These restrictions' pro-vide 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. 0 1

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POWER DlSTRIBUTZON LIMITS BASES 3/4.2.5 DNB PARAMETERS l The limits on the DNB-related parameters assure that each of the parameters )

is maintained within the normal steady-state envelope of operation assumed in l the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3*F for Tavg and 2205 psig for pressurize 2tre not exceeded.  !

9 4 The measurement error of 2.4% for RCS total flow rate is based upon per.  !

forming a precision heat balance and using the result to normalize the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not ,

i be detected could bias the result from the precision heat balance in a noncon-  !

servative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is applied. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending vari-ous plant performance parameters. If detected, action shall be taken before  !

performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

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

F SEABROOK - UNIT 1 8 3/4 2-4 Amendment No. 7. 12

. I  ;

TNSTRUMENTAT10N BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAT10,N,(Continued) uncertainties of uncertainties in the instrumentation calibrating to measure the process variable and t(e the instrumentation. InEquation[3.11r2.2-1 )

Z + R S < TA, the interactive effects of the errors in the rack and~the---,'

sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span; R or Rack Error _is the "as measured" deviation, in the percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.

for a sensor drift factor, an increased rack drift factor, and provides a'~sUse of Equ threshold value for REPORTABLE EVENTS. 32-1 The methodology to derive the Trip Setooints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

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

Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time, i The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined. limits are being exceeded.

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

SEABROOK - UNIT 1 8 3/4 3-2

REACTOR COOLANT SYSTEM BASES 3/4,4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.

The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels.in ex-cess of the Steady-State limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting con-tinued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.

The Surveillance Requirements provide adequate assurance that concentra-ticas rectiveinaction.

excess of the limits will be detected in sufficient time to take cor-3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appro-priately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Seabrook site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

'~The ACTION statement permitting POWER OPERATION to continue for limited

[ 1 mi9ro time p/ riod/with a reacgor coo) ant's pecifi a9 tivi greater tha,n cur}ie/ gram OSE E IVAL T I-13', but eth % th allu ble l' it shown on igure/3.4-1 acc ates possi e io ne Iki g phe menon hich ay oc r followin chang in T LP ER. per tio with specifi acti ty 1 els xceed ng 1 roCur}/ gram OSE E IV4 ENT -131 ut wi, in the limits shown n F1 re 3.4 tha 800ha)urspefyear/

(app xima 61y 1 of the mustteresricteftogono unit's yearly opefatin time since/theactivity/

leve s al wed b Figure 3.4-ljincreas4 tt)4 2-h r th oid $6se at/the SITE BOUNDARYf y a f ctor of up to/20 fol%wirKJ a pystulated steam gerydratovtube '

rupture of cumulative operatipg tiev over,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in'any 6-monttr/' The,reporti i i

consecutive period v' th gre'ater/than/1 miegoCurie/ gram DOSE EQUIVA ENT I-131 Will allow sufficient' time for Commission evaluation of the circumstances prior to reaching the 800-hour lfmit.-

-) '

I SEABROOK - UNIT 1 B 3/4 4-5  !

l l

. t )

3/4.10 SPECfAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be posi-tioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to: (1) measure control rod worth and (2) determine the reactor stability index and damping factor under xenon oscillation conditions.

3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than o' equal to 5% of RATED THERMAL POWER with the RCS T slightly lower avg than normally allcwed so that the fundamental nuclear characteristics of tne core and related instrumentation can be verified. In order.for various charac-teristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 and the RCS T may be below the minimum temperature avg of Specification 3.1.1.4 during the measurement.

3/4.10.4 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain STARTUP and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.5 POSITION: INDICATION SYSTEM - SHUTDOWN This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE.

[1(/. 6 R5ACT0tC00kANTLOO'PS ~

((Thi sp d ene ize for he ial est eMpt np formance s all ct-r c a p ps o 's sta tup est i nd1]onswhilelhe. eactor is uberi, ica .'_

lvi turp [rcylat'/

SEABROOK - UNIT I B 3/4 10-1

(

, OESfGN FEATURES '

DESIGN PRESSURE AND TEMPERATURE 7 maximum internal pressure cf 52.0 psig and a temp .

5. 3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 taining 264 fuel rods clad with Zircaloy-4.The core shall contain active fuel length of 144 inches. Each fuel red shall have a nominal enrichment of 3.15 weight percent U-235.The initial core loading shall have a maximum Reload fuel shall be similar in phy-5.0 weight percent U-235.sical design to the initial core loading and CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod' assemblies.

length control material. rod assemblies shall contain a nominal 142 The inches full- of a dium, and 5%The nominal cadmium. values of absorber material shall be 80% silver,15%

All control rods shall be clad with stainless steel tubing .

5. 4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a.

In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and c.

For a temperature of 650 F, except for the pressurizer which is 680 F.

VOLUME 3

4. 2 12,1,g The total water and steam volume of the Reactor Coolant System is ubic feet at a nominal T

,2 avg of 588.5 F.

5. 5 METEOROLOGICAL TOWER LOCATION ,

5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

SEABROOK - UNIT 1 5-9 Amendment No. 6

l 1

l

. i l

ADMINISTRATIVE CONTROLS l l

6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION j

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

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

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

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

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

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

6.3 TRAINING 6.3.1 A retraining and replacement licensed training program for the station staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements andle.cogendations of Section 5.5 of ANSI N18.1-1971(rd Appedi; A cf 10 CR Part SJand the supplemental require-ment _s_specifiedJ c ,o apd C/s(EhcgggrefjgqCXeJUCA)M grc)f ?g E Vto j 13_e Ae's d nd shall include famillar1zation with relevant in s ry operationa experience.

• Not responsible for sign-off function.

SEABROOK - UNIT 1 6-5

. 1 ADMIN!STRATIVE CONTROLS 6.4 REVIEV AND AUDIT 6.4.1 STATION OPERATION REVIEW COMMITTEE (SORC)

FUNCTION 6.4.1.1 The 50RC matters related shall function to nuclear safety. to advise the Station Manager on all COMPOSITION c a.,..o msW- #

w, l

. t .w. -. e n. o. c__..33 s e_ __ nn_e_ma_

nr +s .

Cha' n: Station Manager Member: 'stant Station Manager

/

Member: Opera . Manager Member: Technical t Ma Memoer: Maintenance De .

Supervisor Member: Instrum ion and Cont Member: Ra r Engineering Departmentepartment Supervisor Memb - visor ealth Physics Department Supervisor

. er: Technical Support Supervisor

"=l r-Cheri:try Dep:r* :nt S;:r;i::: -

ALTERNATES 6.4.1.3 All alternate members shall be ap

%Chairman

11 artici _to_ser.ve-ofLa_ temporary baspO; pointed in writing by the 50RC st :: v

e. % LL Yt h v am; tin- Y :v kcs2ws, 5 :n 50P.C ;;ti.itic: wever, w~ m atn;

ny;;ra enc thad~ x &

tw; alternate time.

ro ri, e,cz _ er r> s=-- N MEhINGFREQUENCY ~ ~ ~ " 8 6.4.1.4 The 50RC shall meet at least once per calendar month and as convened bythe50RCChairman,opj 's designated alternate 5 QUORUM (c e w ,

6.4.1.5 The quorum of the SORC necessary for the performance of the 50RC shall consist of the Chairman orghis designated including alternates.

alternarespon d) ridGf y b our member _sLN-s.c R- j L RESPONSIBILITIES -

6 cea;6 5, e % - % W 6 4 1.6 The 50RC shall be responsible for- ***w**o s .7 m n wronx. rw w ne.c> % *, m* %* *w *~ ~*c" m"

a. Review of: L,w <%wcw 4.v. z c c, v.- ream n.x_

and changes thereto, (2) all proposed programs req P'"/

cation 6.7 and changes thereto, and (3) any other proposed procedures j or changes nuclear safetythereto as determined by the Station Manager to affect

_ '/

b.

Review of all proposed tests and experiments that affect nuclear safety; SEABROOK - UNIT 1 6-6

. V INSERT 1 6.4.1.2 The SORC shall, as a minimum, be composed of the Chairman and nine individuals who collectively have experience and expertise in the following areas:

Nuclear Power Plant Administrative Controls Mechanical Maintenance Electrical Maintenance Instrumentation & Control Chemistry llealth Physics Operations Technical Support / Engineering Reactor Engineeri-?

The Station Manager shall serve as Chairman of the SORC and shall appoint the SORC members in writing.

1 b'.

ADMINISTRATIVE CONTROLS RESPONSIBILITIES 6.4.1.6 (Continued)

c. Review of all proposed changes to Appendix "A" Technical Specificatio.,s:

d.

Review of all proposed changes or modifications to station systems or equipment that affect nuclear safety; e.

Investigation of all violations of the Technical Specifications,  !

including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence, to the Executive Director - Nuclear Production and to the Nuclear Safety Audit Review Committee (NSARC);

f. Review of all REPORTABLE EVENT 5;.

g. Review of station operations to detect potential hazards to nuclear safety;

h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Station Manager or the NSARC;

i. Review of the Security Plan and implementing procedures and submittal of recommendedhbanges.to the__ NSARC;

j. (P--Dew .w T@  ;

Review of the Emergendy~ Plan and implementing procedures and suomittal-of recommendedxtlanges to the NSA 1R -

k.

Q6o_erw/PA__) -

Review of any accidental, unplanned, or uncontrolled radioactive release including the prt*aration of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Executive  ;

Director - Nuclear Production and to the NSARC; j

1. Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and the Radwaste Treatment System; and .I i

m.

Review of the Fire Protection Program and implementing instructions andsubmittalofrecommended{hanges___to_theNSABC 6.4.1.7 The 50RC shall: Qwtur__vuem.ncu N W

~

a. Recommend in writing to the Station Manager approval or disapproval of items considered under Specification 6.4.1.6a. through d;

b. Render determinations in writing with' regard to whether or not each i item considered under Specification 6.4.1.6a. through e. constitutes an unreviewed safety question; and C.

Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Executive Director - Nuclear Production and the NSARC of disagreement between SEABROOK - UNIT 1 6-7 i

.. ADMINISTRATIVE CONTROLS RESPONSIBILITIES 6.4.1.7 (Continued) the 50RC and the Station Manager however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS 6.4.1.8 The 50RC shall maintain written minutes of each 50RC meeting that, at a minimum, document the results of all 50RC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Executive Director-Nuclear Production and the NSARC. -

< ~

6.4.Z er>mou ounueED 6.4(([NUCLEARSAFETYAUDITREVIEWCOMMITTEE(NSARC) 9%m FUNCTION Wh7~ 8 6.4 1 The NSARC shall function to provide independent review and audit of desi nated activities.(if/tpe AteWo6')

Nuclear power plant operations,

b. Nuclear engineering,

c. Chemistry and radiochemistry,

d. Metallurgy, "

Instrumentation and control, e.

f. Radiological safety, i

g. Mechanical and electrical engineering, and j h Quality assurance practices.

L.

The NSARC shall report to and advise the Senior Vi President on those areas

[IofresponsibilityspecifiedinSpecifications6.4 7and6.428.

\ '

s. '

COMPOSITION 6.4 The NSARC shall be composed of at least five (5) individuals. The Chairman,-Vice Chairman and members, including designated alternates, shall l be appointed in writing by the Senior Vice President. Collectively, the i individuals appointed to the NSARC shoul kb c =petent tc c;nict review

.i4cntified by @ecification 6.4.2.1. Each member shall meet the qualifications ,

of ANSI 3.1-1978, Section 4.7. w i-= c s-.tn3 m u v r. u m w ALTERNATE'S 6.4 All alternate members shall be appointed in writing by the Senior  !

Vice PresidentCtMene4nAEwopsWhasf$; however, no more than a minority  ;

shall participate as voting members in NSARC activities at any one time.

CONSULTANTS 6.4. Consultants shall be utilized as determined by the NSARC to orovide exp rt advice to the NSARC. .

SEABROOK - UNIT 1 6-8 l

,  ?.

i j

INSERT 2 FUNCTION i

6.4.2.1 The Station Manager may establish a Station Qualified Reviewer Program '

whereby required reviews of designated programs and procedures required by Specification 6.4.1.6.a are performot by Station QualiGed Reviewers and approved by designated managers. These reviews are in lieu of reviews by the.SORC. However, procedures which require a 10 CFR 50.59 evaluation must be reviewed by the SORC.

.I RESPONSilllLITIES j s

1 6.4.2.2 The Station Quali0ed r.eviewer Program shall: i 1

a. Provide for the review of designated programs, procedures, and changes thereto by a Qualified Reviewer (s) other than the individual who prepared the procedure, program, or change.

l

b. Provide for cross-disciplinary review of programs, procedures, and l changes thereto when organizations other than the preparing organization are afTected by the procedure, program, or change.

c. Ensure cross-disciplinary reviews are performed by a Quali0ed Reviewer (s) in affected disciplines, or by other persons designated by M cognizant Managers or Directors as having specific experthe required i to assess a particular program, procedure, or change. Cross-disciplinary reviewers may function as a committee. I i

d. Provide for a screening of designated programs, procedures, and changes thereto to determine if an evaluation should be performed in ,

accordance with the provisions of 10 CFR 50.59 to verify that an l unreviewed safety question does not exist. This screening will be performed by personnel trained and qualified in performing 10 CFR 50.59 evaluations.

e. Provide for written recommendation by the Quali0ed Reviewer (s) to ~!

the responsible Manager for approval or disapproval of programs and procedures considered under Specification 6.4.1.6a and that the .

procedure or program was screened by a qualiGed individual and found not to require a 10 CFR 50.59 evaluation.

i 6.4.2.3 If the responsible manager determines that a new program, procedure, or change thereto requires a 10 CFR 50.59 evaluation, that Manager will ensure the required evaluation is perfonned to determine if the new procedure, program, or -

change involves an unreviewed safety question. The new procedure, program, or change will then be forwarded with the 10 CFR 50.59 cvaluation to SORC for review.

6.4.2.4 Personnel recommended to be Station Qualified Reviewers shall be designated in writing by the Station Manager for programs or procedures within the scope of the Station Qualified Reviewer Program.

- = _ . - - -

'f

, e-

-e 6.4.2.5 Changes to procedures prior to SORC review and approval shall be made in i

. accordance with Specification 6.7.5 with the exception that changes to procedures for ,

which reviews are assigned to Qualified Reviewers will be reviewed and approved as  :

described in Specification 6.4.2.2. l RECORD'

.. p 6.4.2.6 De review of programs and procedures performed under the Station Qualified Reviewer Program shall be documented in accordance with administrative procedures, TRAINING AND Olla 1IFICATION i 6.4.2.7 The training and qualification requirements of personnel designated as a  ;

Qualified Reviewer in accordance with the Station Qualified Reviewer Program shall be in accordance with administrative procedures.

i i

1 i

i i

'j le i

2

'1 :

ADMfN!STRATfVE CONTROLS MEETING FREQUENCY 3

6.4.I.5 The NSARC shall meet 'at) easy oge ptrgalp6dpf qufrKr derfnsVCnV

@it#ayyept 91 upit ppepatffnfollpGintf fu61 J'oacirvand'ttjeraaff.et/ Tit least once per 6 months 6 weeks.

OUORUM 6.4.19 2,4 The quorum of the NSARC necessary for the performance of the NSARC revie'w and audit functions of these Technical Specifications shall consist of the Chairman or Vice-Chairman and at least four NSARC members including alter-nates.

operation No ofmore than a minority of the quorum shall have line responsibility for the unit.

The Vice Chairman, or his designated alternate, can participate as an NSARC member when the Chairman is in attendance.

REVIEW f

6.4 The NSARC shall be responsible; for the . review of:

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

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

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

d. .

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

Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of. internal procedures or instructions having nuclear safety significance;

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

g. All REPORTABLE EVENTS;

h. All recognized indications of an unanticipated deficiency in some l aspect of design or operation of structures, systems, or comne. ants tr.at could affect nuclear safety; and

i. Reports and meeting minutes of the 50RC.

AUDITS l 6.4 Z.4 Audits of station activities shall be performed under the cognizance of l the 'N3 ARC. The audits shall be performed within the specified time interval l with a maximum allowable extension not to exceed 25% of the specified interval I

SEABROOK - UNIT 1 6-9 l I

a

,. t ADMINISTRATIVE CONTROLS B

AUDITS.

6.4 4 (Continued) provided the combined time interval for any three consecutive intervals shall not exceed 3.25 times the specified interval. These audits shall encompass:

a. The conformance of station operation to provisions contained within the Technical Specifications and applicable license conditiuns at least once per 12 months;

b. The performance, training, and qualifications of the entire station staff at least once per 12 months;

c. The results of actions taken to correct deficiencies occurring in station equipment, structures, systems, or method of operation that affect nuclear safety, at least once per 6 months; d.

The performance' of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months; -

e.

The fire protection programmatic controls including the' implementing >

procedures at-least once per 24 months by qualified licensee QA personnel;  ;

f. The fire protection equipment and program implementetion'at least i once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year; ,

g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;

h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months; ,

. .)

i. The PROCESS CONTROL PROGRAM and implementing procedures for ;oocassing e

I and packaging of radioactive wastes at least once per 24 months; '

j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months;

k. The Emergency Plan and implementing procedures at least once per 12 months; P

1. The Security Plan and implementing procedures at least once per 12 months; and

m. Any other area of station operation considered appropriate by the <

NSARC or the Senior Vice President.

SEABROOK - UNIT 1 6-10

', ADMZNZSTRATIVE CONTROLS RECORDS  !

6.4 h Records of NSARC activities shall be prepared.and distributed as

, indicated below:  :

a. Minutes of each NSARC meeting hall be prepared and forwarded to the Senior Vice President within ays following each meeting;

b. R,eports of reviews encompassed by Specification 6.4.2.7 shall be included in the minutes where applicable or forw rded under sepa-rate cover to the Senior Vice President withir days following completion of the review; and 3C)

c. Audit reports encompassed by Specification 6.4.2.8 shall be forwarded to the Senior Vice President and to the management. positions respon-sible for the areas audited within 30 days after completion of the audit by the auditing organization.

6. 5 REPORTABLE EVENT ACTION The following actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR'Part 50, and

b. Each REDORTABLE EVENT shall be reviewed by the SORC and the results of this review shall be submitted to the NSARC and the Executive Director-Nuclear Production.

6.6 SAFETY LIMIT VIOLATION The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Executive Director-Nuclear Production and the NSARC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;

b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the 50RC. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;

c. The Safety Limit Violation Report shall be submitted to the Commission, '

the NSARC, and the Executive Director-Nuclear Production within 14 days of the violation; and

d. Operation of the station shall not be resumed until authorized by the Commission.

t SEABROOK - UNIT 1 ,

6-11 r

4

. 'ADMfNISTRATIVE CONTROLS * '

6.7 PROCEDURES AND PROGRAMS 6.7.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;

b. The emergency operating orocedures required to. implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33;

c. Security Plan implementation;

d. Emergency Plan implementation;

e. PROCESS CONTROL PROGRAM implementation;

f. OFFSITE DOSE CALCULATION MANUAL implementation;

g. Quality Assurance Program for effluent and environmental  ;

monitoring; )

i

h. Fire Protection Program impismentation; and ~

emwe. %c%o g i. Technical Specificatiorr Improvement Program implementation. hq3l 3 'a~e %.,m 6.7.74 Each procedure of Specification 6.7.1, and changes thereto, shall e r%.x reviewed by the 50RC and shall be approved by the Station Manager, rior to h  %

impl ementati on.snCFeviewed-neModRattv- as -set-NPttr3h-edminm.rmWO WR,W c p,,c,1 cCccG wau 6.7.J vided: Tem rary/hangestoproceduresofSpecification6.7.1maybemade{ pro

a. The intent of the original procedure is not altered; <

b. The change is approved by two members of the plant mana em nt staff, at least one of whom holds a Senior Operator license (b'n the uni +"--

M ffecte@ and

c. The change is documented, reviewed by the 50R Q nd approved by the .

Station Manager, ithin 14 days of implementatisn.- --

6 22MW~ ~ *W 6.7.g The following progt s W u m r4 W;1 De establisnea, implemutEddiid maintain

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems inclu'de the RHR and containment spray, Safety Injection, chemical and volume control. The program shall include the following:

SEABROOK - UNIT 1 6-12

" .# 4 i

INSERT 3 6.7.2 The Station Manager may designate specific programs and procedures to be reviewed in accordance with the Station Qualified Reviewer Program in lieu of review by the SORC. The review per the Qualified Reviewer Program shall be in accordance with Specification 6.4.2.

6.7.3 Programs and procedures listed in Specification 6.7.1, and changes thereto, shall be approved by the Station Manager or by cognizant Managers or Directors who are designated as the Approval Authority by the Station Manager, as specified in administrative procedures. The Approval Authority for each program and procedure shall be specified in administrative procedures.

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ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7M.(Continued)

1) Training of personnel, and

2) Procedures for monitoring,

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

1) Training of personnel,

2) Procedures for sampling and analysis, and 3)

Provisions for meintenance of sampling and analysis equipment.

f. Accident Monitoring Inst umentatio F A program which will ensure the capability to monitor plant variables and systems operating status c'urirq and following an accident. This program shall include those i."tr"ments provided to indicate system operating status and futnish information regarding the release of radioactive materials (Category 2 and in Regulatory Guide 1.97, Revision 2) and instrumentation as defined s provide the following: ,

1) Preventive maintenance and per odic surveillance of instrumentation, .

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

3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable. i 6.8 REPORTING REQUIREMENTS ROUTINE REPORTS 6.8.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT

6. 8.1.1 be submitted A summary following: report of station startup and power escalation testing shall (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of )

fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the station. j l

• I:;h:ntaticaofthisspecificaticrIshalltakeeffect.;henplantgoesabove 5% pcwcr fcr th: first tie:. .

hSeabrookhastakenexceptiontoth in Regulatory Guide 1.97, Revisio 3 a egorization of instrumentation provided {

The Seabrook exceptions are provided in FSAR No. 5. Table 7.5-1, which has been re 1ewed by the NRC staff in SER Supplement SEABROOK - UNIT 1 6-14

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

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

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

The Annual Radioactive Effluent Release Report 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 ho don tanks or_ cas stor tanks exceeding the limits of Specification 3.11.1.4 MAW,Wy6sbd%V/i MONTHLY OPERATING REPORTS 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: 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 REPORT 6.8.1.6.a Core operating limits shall be established and documented in the CORE 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,

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

3. Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3.1.1.3, SEABROOK - UNIT 1 6-18 Amendment No. N, 22,

!!i. Retype of Proposed Channes -

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 this submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

Revision bars are provided in the right hand margin to designate a change in the text.

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INDEX s LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS  :

a SECTION PAGE. .

t 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4 9-4 3/4.9.5 COMMUNICATIONS . . . . . .... 3/4 9-5 .!'

3/4.9.6 REFUELING MACHINE ... ....... .. .. 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS ...... 3/4 9-7 3/4.9.8 RESIDUAL' HEAT REMOVAL AND COOLANT CIRCULATION ,

High Water Level . . .. . 3/4 9-8  ;[

Low Water Level ... ... . . .. 3/4 9-9 s l

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM . 3/4 9 110 3/4.9.10 WATER LEVEL - REACTOR VESSEL . .. 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL . . . . . . . . . . . . 3/4 9-12  :

3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM . 3/4 9 3/4.9.13 SPENT FUEL ASSEMBLY STORAGE . .. .... .. 3/4-9 FIGURE 3.9-1 FUEL ASSEMBLY BURNUP VS. INITIAL ENRICHMENT .;

3/4 9-17 FOR SPENT FUEL ASSEMBLY STORAGE . .

3/4.9.14 NEW FUEL ASSEMBLY STORAGE . . . . . 3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS . i 3/4.10.1 SHUTDOWN MARGIN ... ................ 3/4 10-1  ;

3/4.10.2 GROUP HEIGHT, INSERTION. AND POWER DISTRIBUTION LIMITS . 3/4 10-2:  :

3/4.10.3 PHYSICS TESTS .... . . . 3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS . . .. . . .. 3/4 10-4 a 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN . .. 3/4 10-5 ,

I i 3/4.11 RADICACTJVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS  ;

Concentration . . . , , 3/4 11-1  :

Dose . . . .. . . . ... 3/4 11-2  !

Liquid Radwaste Treatment System .. 3/4 11-3  ;

Liquid Holdup Tanks . . 3/4 11-4 i 3/4.11.2 GASEOUS EFFLUENTS Dose Rate .. . . . . , 3/4 11-5 Dose - Noble Gases . . .. ... .. .

.. '3/4 11-6 l Dose - Iodine-131 Iodine-133. Tritium, and Radioactive Material in Particulate Form . . . . 3/4 11-7 ,

Gaseous Radwaste Treatment System ... 3/4 11-8 '

Explosive Gas Mixture - System .. . .. 3/4 11-9 j 3/4.11.3 SOLID RADIOACTIVE WASTES . . . . 3/4 11 3/4.11.4 TOTAL DOSE . . . . .... 3/4 11-12 1 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM . . . . . 3/4 12-1 SEABROOK - UNIT 1 ix Amendment No. 6,

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

INDEX l 650 ' ADMINISTRATIVE' CONTROL $ j SECTION PAGE-6.4 REVIEW AND AUDIT . ... . . 6-6 o 1

6.4.1 - STATION OPERATION REVIEW COMMITTEE (SORC) .

i Function . . . . . .. .. 6-6 j

- Composition . . .. . . . 6-6 Alternates . . . . , ... . 6  ;

Meeting Frequency . . ... .. . . . 6-6  :

Quorum . . . . . .. . -. . . . . . 6 Responsibilities . . ..,. . .

6-6 Records . . , . ....,. . . . ' 6-8 .

6.4.2 STATION OUALIFIED REVIEWER PROGRAM . ... . . 6-8' a

Function . . . . . . . . . 6-8 Responsibilities . . .. ....,. 6-8 1,

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Records . . . . . w . . . ,, . . . 6-8A-Training and Qualification . . .... 6-8A_ .

6.4.3 NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC) .

Function . . . .......... . . . 6 8A Composition . . . . ...... . 6-8A Alternates . . . ...... . 6-9 Consultants . . . . ... .... ..... 6-9  :

Meeting Frequency . . . .. . . . . 9 -

QuorumE. . . . . . . . . . 6-9 e Review . . . . .. . , .6-9 .;

Audits .

6;10' l Records . . . .. .. . . . . 6 E.5 REPORTABLE EVENT ACTION . ..... . '6-11 6.6 SAFETY LIMIT VIOLATION . . ... .. . . . 6-11 l

6.7 PROCEDURES AND PROGRAMS . . . . , . . 6-12L {

6.8 REPORTING RE0UIREMENTS >

6.8.1 - ROUTINE REPORTS. . . . . . . 6- 14A .'  ;

Startup Report.. . . . . .. .......... 6-14A Annual Reports . . . . ... .... .. . 6-15'  ;

Annual Radiological Environmental Operating Report . . . 6-15 .i Annual Radioactive Effluent Release Report- . . . 6-17  :

Monthly Operating-Reports . .. . .... 6 t

. CORE OPERATING LIMITS REPORT , . . 18 J

6.8.2 SPECIAL REPORTS . . , . . . . . 6-19 j I

4- SEABROOK - UNIT 1 xiv Amendment No. 9. 22.

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INDEX  ;

l 6.0- ADMINISTRATIVE CONTROLS' SECTION PAGE {

6.9 RECORD RETENTION . . . . .. 6-19' 6.10 RADIATION PROTECTION PROGRAM . . . ... 6-20 l i

6.11 HIGH RADIATION AREA . . . . .. . 6-20 6.12 PROCESS CONTROL PROGRAM (PCP) .. .. . . 6-21 ,

6.13 0FFSITE DOSE CALCUL ATION MANUAL (0DCM) . . .. 6 6.14 MAJOR CHANGES TO LIQUID. GASEOUS. AND SOLID RADWASTE TREATMENT SYSTEMS . ... 6-23 *

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f SEABROOK - UNIT 1 xv Amendment No. l

-, .n DEFIN'ITIONS- ,

PROCESS CONTROL PROGRAM l

-1.25 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 f processing of actual or simulated wet solid wastes will be accomplished in such- i 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.

F PURGE - PURGING l 1.26 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.

I QUADRANT POWER TILT RATIO 1.27 OUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore 1 detector calibrated output to the average of the upper.excore-detector cali- -

brated outputs; or the ratio of_ the maximum lower excore detector calibrated l 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

1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to-the reactor coolant of 3411 Mwt.  !

REACTOR TRIP SYSTEM RESPONSE TIME  ;

1.29 The REACTOR TRIP SYSTEM RESPONSE TIME shall.be_the time interval from j when the monitored parameter exceeds its Trip Setpoint at the channel sensor?

until loss of stationary gripper coil voltage.

- REPORTABLE EVENT ,

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

-l CONTAINMENT ENCLOSURE BUILDING INTEGRITY  !

1.31 CONTAINMENT ENCLOSURE BUILDING INTEGRITY shall exist'when:

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

b. The Containment Enclosure Emergency Air Cleanup System is OPERABLE.  ;

and

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

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

SEABRO'i - UNIT 1 1-5 Amendment No. 7. 9. +

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.t j LIMITING SAFETY SYSTEM SETTINGS BASES i

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) i I

Undervoltaae and Underfreauency - Reactor Coolant Pumo Busses ,

The Undervoltage and Underfrequency Reactor toolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is i generated before the Low Flow Trip Setpoint is reached. Time delays are ,

incorporated in the Underfrequency and Undervoltage' trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor i trip breakers following the simultaneous trip of two or more reactor. coolant'  !

pump bus circuit breakers shall not exceed 1.5 seconds. For underfrequency. l the delay is set'so that the time required for a signal to reach the Reactor '

trip breakers after the Underfrequency Trip Setpoint is reached shall not  ;

exceed 0.3 second. On decreasing power the Undervoltage and Underfrequency . i Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber '

pressure at approximately 10% of full power equivalent): and on increasing power, the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are. -

reinstated automatically by P-7.

Turbine Trio

. A Turbine tri) initiates a Reactor trip. On decreasing power, the '[

Reactor trip from t1e Turbine trip is automatically blocked by P-9 (a power- L level of approximately 20% of RATED THERMAL POWER): and on increasing )ower,  !

the Reactor trip from the Turbine trip is reinstated automatically by 3-9. l Safety In.iection Inout from ESF ,

If a Reactor trip has not already been generated by the Reactor Trip -

System instrumentation. the ESF automatic actuation logic channels will initiate a Reactor trip u)on any signal which initiates a Safety Injection. i The ESF instrumentation clannels that initiate a Safety Injection signal are.

shown in Table 3.3-3. j Reactor Trio System Interlocks j The Reactor Trip System interlocks perform the following functions: ,

P-6 On increasing power. P-6 allows the manual block of the Source i Range trip (i.e. , prevents premature block of Source Rarige trip). t On decreasing power, Source Range Level' trips are automatically  ;

reactivated and high voltage is restored.  !

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SEABROOK - UNIT 1 B 2-7 Amendment No.

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TABLE 4.3-1 (Continued) I

-TABLE NOTATIONS (Continuedl (12) Number not used.

(13) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the -l- ,

- OPERABILITY of the undervoltage and shunt trip circuits for the' Manual- '

Reactor Tri) Function-. The test shall also verify the OPERABILITY of the  ;

Bypass Brea ter trip circuit (s). i

.i (14) local manual shunt trip prior to placing breaker in ' service' .

(15) Automatic undervoltage trip.  ;

(16) Each channel shall be tested at least every 92 days on a STAGGERED' TEST BASIS. j ,

(17) These channels also provide inputs to ESFAS. Comply with the applicable MODES and surveillance frequencies of Specification 4.3.2.1 for any-por-  :

tion of the channel required to be OPERABLE by Specification 3.3.2.  ;

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SEABROOK - UNIT 1 3/4 3-13 Amendment No, p2, 47,  ;

,~ _ - _, , , , . r,-_. _ _ , .- y

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. , - ..d- l JNSTRUMENTATION: ,

MONITORING INSTRUMENTATION l ACCIDENT MONITORING INSTRUMENTATION ,

LIMITING CONDITION FOR OPERATION j t

3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3 .

shall be OPERABLE.

- ' APPL ICABILITY: MODES-1. 2. and 3.

ACTION:

a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table  ;

3.3-10, restore the inoperable channel (s) to OPERABLE status within i 7 days. or be in at least HOT STANDBY within the~next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in -

at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. .The provisions >

of Specification 3.0.4 are not applicable.

b. With the number of OPERABLE accident monitoring instrumentation channels except the containment POST-LOCA high range area monitor, j -

3 less than the Minimum Channels ~ OPERABLE requirements of Table -i 3.3-10. restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and-  ;

in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The pro-  ;

visions of ~ Specification 3.0.4 are not applicable.

c. With the number of OPERABLE channels for the containment Post-LOCA high range area monitor less than required by the Minimum Channels- ,

OPERABLE. requirements, initiate an alternate _ method of monitoring:

the appropriate-]arameter(s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and either restore i the inoperable clannel(s) to OPERABLE status within 7 days or .

l prepare and submit a Special Report to the Commission. pursuant to- ,

Specification 6.8.2. within 14 days that provides actions taken,  !

cause of the inaperability, and the plans and schedule for restoring i the channels te; OPERABLE status. l I

SURVEILLANCE REQUIREMENTS j 4.3.3.6 Each accident monitoring instrumentation channel shall be . demonstrated ~

OPERABLE j

a. Every 31 days by performance of a CHANNEL CHECK. and 'l

b. Every 19 months by performance of a CHANNEL CALIBRATION.

SEABROOK - UNIT 1 3/4 3-49 Amendment No.

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REACTOR COOLANT SYSTEM  ;

REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

- HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when thei Reactor Trip System breakers are open:*  :

a. Reactor Coolant Loop A and its associated steam generator and .;

reactor coolant pump.  ;

~

b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump, l m

c. Reactor Coolant loop C and its associated steam generator and '

reactor coolant pump. and

d. Reactor Coolant Loop D and its associated steam generator and  ;

reactor coolant pump.  ;

APPLICABILITY: MODE 3. ,

ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the recuired loo]s to OPERABLE status within-72 hours or be in HOT SHUTDOWh within t1e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position; within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the -

Reactor Trip System breakers.

c. With no reactor coolant loop in operation, suspend.all operations involving a reduction in boron concentration-of the Reactor Coolant i System and immediately initiate corrective action-to return the  :

required reactor coolant loop to operation. _

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• All reactor coolant pumps may be deenergized for'up to I hour.provided: (1) no .I operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.

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

. - ~. . . . - -

TCONTAINMENT SYSTEMS.

CONTAINMENT ISOLATION VALVES

-SURVEILLANCE REQUIREMENTS 4'6.3.2 Each containment isolation valve shall be demonstrated OPERABLE during

. shutdown at least once per 18 months by: l -

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A"' -

Isolation valve actuates to its isolation' position,

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" Isolation valve actuates to its isolation position.'and

c. Verifying that on a Containment Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation '

position.

i-4.6.3.3 The isolation time of each power-operated or automatic containment isolation v61ve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. i i

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i SEABROOK - UNIT 1 3/4 6-17 Amendment No. 44 4.

+ ,r- e - - - . = s- , - . -- - --, - - . , -

PLANT SYSTEMS TURBINE CYCLE P AUXILIARY FEEDWATER SYSTEM j LIMITING CONDITION FOR OPERATION ,

3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. One motor-driven emergency feedwater pump, and one startup feedwater aump capable of being powered from an emergency bus and capable of '

]eing aligned to the dedicated water volume in the condensate storage tank, and r

b. One steam turbine-driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2. and 3. l ACTION: 1

a. With one auxiliary feedwater pump inoperable, restore the required ,

auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With two emergency feedwater pumps inoperable restore at least one emergency feedwater pump to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and restore both emergency feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in .

COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

c. With one emergency feedwater pump and the startup feedwater Jump inoperable. restore both emergency feedwater pumps to OPERAB_E status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and all three pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. With three auxiliary feedwater pumps inoperable. immediately ,

initiate corrective action to restore at least one auxiliary '

feedwater pump to OPERABLE status as soon as possible.

SJRVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1) Verifying that the motor-driven emergency feedwater pump develops a discharge pressure of greater than or equal to 1460 psig at a flow of greater than or equal to 270 gpm;

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

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SEABROOK - UNIT 1 3/4.10-6 Amendment No.  !

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REACTIVITY CONTROL SYSTEMS l BASES 3 I

'3/4.1.2 BORATION SYSTEMS (Continued)

. The boron capability required below 200*F is sufficient to provide a r' 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- i storage tanks or a minimum contained volume of 24.500 gallons of 2000 ppm borated water from the RWST.

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

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

The limitations on OPERABILITY of isolation provisions for the Boren i Thermal Regeneration System and the Reactor Water Makeup System in Modes'4 4 .5, i

~

and 6 ensure that the boron dilution flow rates cannot exceed the value assumed in the transient analysis. ,

l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1)' acceptable power  !

distribution limits are maintained. (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of, rod misalignment on associated 1 accident analyses are limited. OPERABILITY of the control rod Josition indicators is required to determine control rod positions and t1ereby ensure-  :

compliance with the control rod alignment and insertion limits. Verification  !

that the Digital Rod Position Indicator agrees with the demanded position within- l 1 12 steps at 24, 48,120. and 228 steas withdrawn for the Control Banks and 18.  !

210. and 228 steps withdrawn for the Slutdown 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 i position.  !

The ACTION statements which permit limited variations from the basic. I requirements are accompanied by additional restrictions which ensure that the  ;

original design criteria are met. Misalignment of a rod requires' measurement of j peaking factors and a restriction in THERMAL' POWER. These restrictions provide 'j 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. .,

1 l

SEABROOK - UNIT 1 B 3/4 1-3 Amendment No. 9.

. .. 1

^

POWER DISTRIBUTION LIMITS f

BASES'

-1 3/4.2.5 DNB PARAMETERS-The limits on the DNB-related parameters assure that each of the parameters is maintained within the normal steady-state envelope'of operation assumed in the transient and accident analyses . The limits are consistent with .

the initial FSAR assumptions and have been analytically demonstrated adequatelto -

maintain a minimum'DNBR of 1.30 throughout each analyzed transient. . Operating -

procedures include allowances for measurement and indication uncertainty so that the limits of 594.3 F for T m and 2205 psig for pressurizer pressure are not exceeded

[ -

The measurement error of 2.4%' for RCS total flow rate is based upon per- i forming a precision heat balance and using the result to normalize the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be .

detected could bias the result from the precision heat balance in a noncon- 9 servative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is applied. Any fouling which might bias the RCS flow rate '

measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall-be taken before performing subsequent precision heat balance measurements',.i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow-rate measurement or the venturi shall be cleaned. to eliminate the fouling.

The 12-hour periodic surveillance of these parameters through instrument  :

readout is sufficient to ensure that the parameters are ' restored within their i limits following load changes and.other expected transient operation. i The periodic surveillance of indicated RCS flow is sufficient to' detect f only flow degradation which could lead to operation outside the specified limit.

.I 4

1

)

j SEABROOK - UNIT 1 B 3/4.2-4 Amendment No. 9. 42.

1

..- ~

.MSTRUMENTATION 1

BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 2.2-1. l Z + R S 5 TA. the interactive effects of the errors in the rack. and the sensor, and the "as measured" values of the errors are considered. Z' as.specified in-

. 1 Table 3.3-4. in percent span. is the statistical summation of errors assumed in.

the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance.is the' difference, in percent span: R or Rack Error is the "as measured" deviation. in the percent ~ 1 span, for the affected channel from the specified Trip.Setpoint. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration  :

point or the value specified in Table 3.3-4. in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor ~. an l- '

increased rack drift factor.-and provides a threshold value for. REPORTABLE EVENTS.

The methodology to derive the Trip Setpoints is based upon combining all?

of the uncertainties in the channels. Inherent to the determination of the Trip .

Setpoints are the magnitudes of these channel uncertainties. Sensor and rack .~

instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in i excess of the Allowable Value exhibits the behavior that the rack has.not met its allowance. Being-that there is a small statistical chance that this'will-happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasioncl. may be indicative of more serious problems and should warrant further investigation.

The measurement of response time at the specifi'e d frequencies.provides '

assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with-response times indicated as not applicable. Response time may ee demonstrated' by any series of sequential. overlapping or total channel test measurements >

provided that such tests demonstrate the total channel response .ime as defined.

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

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

exceeded. If they are the signals are combined into logic matrices ' sensitive  :

to combinations. indicative of various accidents. -events, and transients. Once the required logic combination is completed the system sends actuation signals ~;

to those Engineered Safety Features components whose aggregate function best: >

serves the requirements of the' condition. As an example, the following actions .

4 may be initiated by.the Engineered Salety Features Actuation System to mitigate the consequences of a steam line' break or loss-of-coolant accident: (1) Safety  ;

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

Y n , , - ,,a ..m- , , --

. _ _ - _ - -.- - -I

REACTORC00lLANTSYSTEM BASES 3 2/4.4.7 CHEMISTRY  !

The limitations on Reactor Coolant System chemistry ensure that corrosion l of the Reactor Coolant System is minimized and reduces the potential for Reactor i Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection. I to ensure the structural integrity of the Reactor Coolant System over the life -  !

of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that o]eration may be continued with contaminant concentration levels in excess of -

t7e Steady-State Limits, up to the Transient Limits for the specified limited l time intervals without having a significant effect on the structural integrity  ;

of the Reactor Coolant System. The time interval permitting continued operation '

within the restrictions of the Transient Limits provides time for taking  !

corrective actions to restore the contaminant concentrations to within the Steady-State Limits. ,

1 The Surveillance Requirements provide adequate assurance that concentra- l tions in excess of the limits will be detected in sufficient time to take cor- a rective action.

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure-  !

that the resulting 2-hour doses at the SITE-BOUNDARY will not exceed an appro-priately.small fraction of.10 CFR Part 100 dose guideline values following a- i steam generator tube rupture accident in conjunction with an assumed steady-  !

state reactor-to-secondary steam generator leakage rate of 1 gpm. The values j for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Seabrook site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

i l

q SEABROOK - UNIT 1 B 3/4 4-5 Amendment No I

a

S t,. ..

9 3/4.10 SPECIAL TEST EXCEPTIONS r BASES 3/4.10.1 SHUTDOWN MARGIN ,

This special test exception provides that a minimum amount of control rod- .

worth is immediately available for reactivity control when tests are performed  !

for control rod worth measurement. This special test exception is required to- 1 permit the periodic verification of the actual versus predicted core reactivity j condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 GROUP HEIGHT. INSERTION. AND POWER DISTRIBUTION LIMITS  ;

This special test exception permits individual control rods to be posi--

tioned outside of their normal group heights and insertion limits during.the -  :

performance of such PHYSICS TESTS as those required ia: (1) measure control  !

rod worth and (2) determine the reactor stability inuex and damping factor under xenon oscillation conditions. 1 3/4.10.3 PHYSICS TESTS This specr test exception permits PHYSICS TESTS to be performed at'less 1 than or_ equal to 6% of RATED THERMAL POWER with the RCS T,, slightly lower than  :

normally allowed so that the fundamental nuclear characteristics of the core-  !

and related instrumentation can be verified. In order for various charac- i teristics to be accurately measured, it is at times necessary to operate- l outside the normal restrictions of these Technical Specifications. For  ;

instance, to measure the moderator temperature coefficient at BOL. it is  ;

necessary to position the various control rods at heights which may not l normally be allowed by Specification 3.1.3.6 and the RCS T,, may be below the  ;

minimum temperature of Specilication 3.1.1.4 during the measurement. j 3/4.10.4 REACTOR COOLANT LOOPS '!

- . 1 This special test exception permits reactor criticality under no flow conditions and is required to perform certain STARTUP and PHYSICS TESTS while- ,

at low THERMAL POWER levels. t 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN l This special test exception permits _the Position Indicatic Systems _to be' l inoperable during rod drop time measurements. The exception is required since:  !

the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped, This induced- t voltage-is small compared to the normal voltage and. therefore, cannot be- a observed if the Position Indication Systems remain OPERMLE. j r

.I SEABROOK - UNIT 1 B 3/4 10-1 Amendment No. j

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

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 52.0 psig and a temperature of 296 F.-

5.3 REACTOR CORE FUEL ASSEMBLIES i

5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a-nominal active fuel length of 144 inches. The initial core loading shall have  :

a maximum enrichment of 3.15 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 5.0 weight percent U-235.

j CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-  :!

length control rod assemblies shall contain a nominal 112 inches.of absorber  !

material. The nominal values of absorber material shall be 80% silver.15%  :

indium. and 5% cadmium. All control rods shall be clad with stainless steel '

tubing.  ;

i 5,4 REACTOR COOLANT SYSTEM  :

DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be. maintained: . ;i

a. In accordance with the Code requirements specified in~Section;5.2  !

of the FSAR. with allowance for normal degradation pursuant to the a applicable Surveillance Requirements.  ;

i

b. For a pressure of 2485 psig, and

c. For a temperature of 650 F. except for the pressurizer which is -

680 F. 3 i

VOLUME .

5.4.2 The total water and steam volume of the Reactor Coolant System'ic 12.255 .;

cubic feet at a nominal T m of 588.5 F. j 5.5 METEOROLOGICAL TOWER LOCATION j 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.  ;

a

~

SFMR00K - UNIT 1 5-9 Amendment No. 6,

, . .. I ADMINISTRATIVE CONTROLS 1

6.2.3 INDEPENDENT-SAFETY ENGINEERING GROUP (ISEG)

FUNCTION [

6.2.3.1 The ISEG shall function to examine station operating characteristics.

NRC issuances, industry advisories. Licensee Event Reports, and other sources-of station design and operating experience information, including units of similar design. which may indicate. areas for improving station safety. The ISEG shall make detailed recommendations for revised procedures, equipment .;

modifications, maintenance activities operations activities, or other means of  ;

improving station safety to the Senior Vice President.

COMPOSITION

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

RESPONSIBILITIES 6.2.3.3 The ISEG shal1 be responsible for maintaining surveillance of station activities to provide independent verification

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

RECORDS ,

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

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Control Rocn Commander in the areas of thermal hydraulics, reactor engi-~  ;

neering, and plant analysis with regard to the safe operation of the station.

6.3 TRAINING >

t 6.3.1 A retraining and replacement licensed training program for the . station -l staff shall be maintained under the direction of the Training Manager and shall l meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 aid the supplemental requirements specified in-NUREG-1021, and shall  :

include familiarization with relevant industry operational . experience. '

• Not responsible for sign-off function.

I i

SEABROOK -UNIT 1 6-5 Amendment No. .]

.- e e ,.e-

c .. .

~ ADMINISTRATIVE CONTROLS 6.4' REVIEW AND AUDIT 6.4.1 STATION OPERATION REVIEW COMMITTEE (SORQ FUNCTION l

1.1 The 50RC shall function to advise the Station Manager on all matters related to nuclear safety. '

COMPOSITION 6.4.1.2 The SORC shall. as a minimum, be composed of the Chairman and nine individuals who collectively have experience and expertise in the following i areas: 3 Nuclear Power Plant Administrative Controls Mechanical Maintenance Electrical Maintenance Instrumentation & Control '

Chemistry Health Physics Operations Technical Support / Engineering Reactor Engineering -

The Station Manager shall serve as Chairman of the 50RC and shall appoint'the  !

SORC members in writing. ,

ALTERNATES 6.4.1.3 All alternate members shall be appointed in writing by the SORC Chairman to serve on a temporary basis and shall have qualifications equivalent to those of the members.

MEETING FREQUENCY 6.4.1.4 The 50RC shall meet at least once per calendar month and as convened  !

by the 50RC Chairman or one of his designated alternate (s). l OUORUM 6.4.1.5 The quorum of the 50RC necessary for the performance of the SORC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or one of his designated alternate (s) and at least four  ;

SORC members including alternates.

RESPONSIBillTIES 6.4.1.6 The 50RC shall be responsible Tor: 9

a. Review of: (1)'all p oposed 6.7 and changes thereto. (2) all procedures proposed required programsby. Specification-required by l

Specification 6.7 and changes thereto, and (3) any other proposed orocedures or changes thereto as determined by the Station Manager in aficct nucle 6r safety. Procedures and orograms r quired by Saecification 6.7 that are desianated for ' review ana approval by t1e Station Qualified Reviewer Program in accordance with Specification 6.4.2 do not require SORC review.

b. Review of all proposed tests and experiments that affect nuclear l safety; SEABROOK - UNIT 1 6-6 Amendment No.

,. - ~ .. 1

}

ADMINISTRATIVE CONTR0l.S RESPONSIBILITIES l 6.4.1.6 (Continued) ,

c. Review of all proposed changes to Appendix "A" Technical Specifications: '

d. Review of all proposed changes or modifications to station systems l or equipment that affect nuclear safety; i

e. Investigation of all violations of the Technical Specifications- ' ,

including the preparation and forwarding of reports covering -

evaluation and recommendations to prevent recurrence. to the ,

Executive Director - Nuclear Production and to the-Nuclear Safety -;

Audit Review Committee (NSARC): j

f. Review of all REPORTABLE EVENTS:

g. Review of station operations to detect potential hazards to ' nuclear safety h.

Performance of special reviews, investifations or anal ses andreportsthere

1. Review of the Security Plan and implementing procedures and submittal of recommended Security Plan changes to the-NSARC: 1 i i

j. Review of the Emergency Plan and im]lementing procedures and -

submittal of recommended Emergency 31an changes to the NSARC: l

k. Review of any accidental, unplanned, or uncontrolled radioactive . -t release inck u.ag the preparation of- reports covering evaluation, l recommendatiMs end disposition of the corrective action to 4

]revent recur ance and the forwarding of these reports to the  ;

Executive Director - Nuclear Production and to the.NSARC: 1 1 Review of chances to the PROCESS CONTROL PROGRAM. OFFSITE DOSE CALCULATION M.WJAL, and the Radwaste Treatment System; and  !

m. Review of the Fire Protection Program and implementing instructions-and submittal of recommended Fire Protection Program changes to the. l- -

NSARC. .

6.4.1.7 The SORC shall: i

a. Recommend in writing to the Station Manager approval.or. disapproval i of items considered under Specification 6.4.1.6a. through d:  !>

b. Render determinations.in writing with regard to whether or not each . .

item considered under Specification 6.4.1.6a. through e. t constitutes an unreviewed safety question; and j

c. Frovide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Executive ..

Director - Nuclear Production and the NSARC of disagreement between the 50RC and the Station Manager however, the Station Manager shall.  !

have responsibility for resolution of such disagreeirierits pursuorit i to Specification-6.1.1. q

'?

r SEABROOK - UNIT 1 6-7 Amendment _No.

-. ..e ., . -

. . . , a.

ADMINISTRATIVE CONTROLS RECORDS 6.4.1.8 The SORC shall maintain written minutes of. each SORC meeting that ' at .

a minimum, document the results of all SORC activities performed under the  !

responsibility provisions of these Technical Specifications. Copies shall be-  :

provided to the Executive Director-Nuclear Production and the NSARC.

6.4.2 STATION OUALIFIED REVIEWER PROGRAM FUNCTION >

6.4.2.1 The Station Manager may establish a Station Qualified Reviewer Program whereby required reviews of designated programs and procedures required by Specification G.4.1.6.a are performed by Station Qualified Reviewers and approved by desi! s ed managers. These reviews are in. lieu of reviews by the 50RC. However. procedures which require a 10 CFR 50 59 evaluation must be  !

reviewed by the SORC. 1 RESPONSIBll.ITIES j 6.4.2.2 The Station Qualified Reviewer Program shall: ,

a. Provide for the review of designated programs, procedures, and .:

changes thereto by a Qualified Reviewer (s) other than the-  !

individual who prepared the procedure program, or change.

b. Provide for cross-disciplinary review of programs, procedures. and-changes thereto when organizations other than the_ preparing -

organization are affected by the procedure program, or change.  ;

c. Ensure cross-disciplinary reviews are performed by a Qualified Reviewer (s) in affected disciplines, or by other persons _ designated-  ;

by cognizant Managers or Directors as having specific expertise required to assess a particular program, procedure er change.

Cross-disciplinary reviewers may function as a committee _. l

d. Provide for a screening of designated programs, procedures and 1 changes thereto to . determine if an evaluation should be performed -,

in accordance with the provisions of 10 CFR 50.59 to verify that an unreviewed safety question does not exist. This screening will.be i]

performed by personnel trained and qualified in performing 10 CFR-  !

50.59 evaluations. )

l

e. Provide for written recommendation by the Qualified Reviewer (s) to- )

the responsible Manager for approval or disapproval of programs and  !

procedures considered under Specification 6.4.1.6a .and that the program or procedure was screened by a qualified individual and found not to require a 10 CFR 50.59 evaluation.

SEABROOK UNIT 1 6-8 Amendment No.

. .?

ADMINISTRATIVE CONTROLS 6.4.2.3 If the responsible manager determines that a new program, procedure or change thereto requires a 10 CFR 50.59 evaluation that Manager will. ensure the required evaluation is performed to determine if the new procedure.. '

program, or change _ involves an unreviewed safety question. The new procedure, rogram, or change will then be forwarded with the 10 CFR 50.59 evaluation to ORC for review.

6.4.2.4 Personnel recommended to be Station Qualified Reviewers shall be designated in writing by the Station Manager for pro the scope of the Station Qualified Reviewer Program. grams and procedures within 1

6.4.2.5 Changes to procedures made prior to SORC review shall be made in r accordance with Specification 6.7.5 with the exception that changes to  ;

procedures for which reviews are assigned to Qualified Reviewers will be reviewed and approved as described in Specification 6.4.2.2.

RECORDS 6.4.2.6 The review of programs and procedures performed under the Station Qualified Reviewer Program shall be documented in accordance with '

administrative procedures.

l TRAINING AND OUALIFICATION f

6.4.2.7 The training and qualification requirements of personnel designated as-a Qualified Reviewer in accordance with the Station Qualified Reviewer Program  ;

shall be 5 accordance with administrative procedures.

6.4.3 Nuu EAR SAFETY AUDIT REVIEW COMMITTEE (NSARC)

FUNCTION 6.4.3.1 The NSARC shall function to provide independent review and audit of designated activities. The NSARC shall report to and advise the Senior Vice i President on those areas of responsibility specified in Specifications 6.4.3.7 and 6.4.3.8. ,

COMPOSITION  :

6.4.3.2 The NSARC shall be composed of at least five (5) individuals, The a Chairman. Vice Chairman and members, includinc designated alternates, shall be appointed in writing by the. Senior Vice Presif ent. Collectively, the '

individuals appointed to the NSARC should have experience and expertise in the.

following areas: '

a. Nuclear power plant operations. '

b. Nuclear e gineering,

c. Chemistry and radiochemistry.

~d. Metallurgy.

e. Instrumentation and control,

f. Radiological safety, j

g. Mechanical and electrical engineering, and  ;

h. Quality assurance practices. 1 Each member shall meet the qualifications of ANSI 3.1-1978.-Section 4.7.

-SEABROOK - UNIT 1 6-8A Amendment No.

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

ADMINISTRATIVE CONTROLS l

ALTERNATES ,

6.4.3.3 All alternate members shall be appointed in writing by the Senior Vice l -

President: however, no more than a minority shall participate as voting members in NSARC activities at any one time. ,

CONSULTANTS  ;

6.4.3.4 Consultants shall be utilized as determined by the'NSARC to provide- I expert advice to the NSARC.

MEETING FREQUENCY 6.4.3.5 The NSARC shall meet at.least once per 6 months 6 weeks, OUORUM 6.4.3.6 The quorum of the NSARC necessary for the performance of the NSARC .

review and audit functions of these Technical Specifications shall consist of the Chairman or Vice-Chairman and at least four NSARC members including alter- ,

nates. No more than a minority of the quorum shall have line responsibility .

for operation .of the unit. The Vice Chairman, or his designated alternate, can participate as an NSARC member when the Chairman is in attendance.

REVIEW 6.4.3.7 The NSARC shall be responsible for the review of: I

a. The safety evaluations for: (1) changes to procedures, equipment, ,

or systems and (2) tests or experiments completed under the  :

provision of 10 CFR 50.59 to verify that such actions did not ,

constitute an unreviewed safety question:

b. Proposed changes to procedures. ecuipment, or -systems that involve. l an unreviewed safety question as cefined in 10 CFR 50.59:

c. Proposed tests or experiments that involve an unreviewed safety question as defined in 10 CFR 50.59:

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

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

license reguirements, or of internal procedures or instructions  !

having nuclear safety significance.  ;

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

g. All REPORTABLE EVENTS:  ;

e

h. All recognized indications of an unanticipated deficiency in some- 1 aspect of design or operation of structures, systems. or components j that could affect nuclear safety; and

i. Reports and meeting minutes of the SORC.

SEABROOK - UNIT 1 6-9 Amendment No.

_ _ _ . _ _ ._ _ .V - _,

ADMINISTRATIVE CONTROLS j

. AUDITS l

6.4.3.8 Audits of station activities shall be performed under the cognizance. l l of the NSARC. The audits shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the specified interval-provided the combined time interval for any three consecutive intervals shall i not exceed 3.25 times the specified interval. These audits shall encompass:

a. The conformance of station operation to 3rovisions contained within -

the Technical Specifications and applica>1e license conditions at i least once per 12 months: l

b. The performance, training and qualifications of the entire station ,

staff at least once per 12 months: '

c. The results of actions taken to correct deficiencies occurringLin station equipment. structures, systems, or method of operation.that affect nuclear safety, at least nnce per 6 months:

d. The performance of activities required by the'0perational Quality 'l Assurance Program to meet the criteria of Appendix B.10 CFR Part i 50 at least once per 24 months: l

e. The fire protection programmatic controls including the  ;

implementing procedures at least once per 24 months by-qualified ,

licensee QA personnel:  !

f. The fire protection equipment and program implementation at least -  !

once per 12 months ut111 zing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year:

l

g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months; ,

h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months: jP

i. The PROCESS CONTROL PROGRAM and implementing procedures for 3 processing and packaging of radioactive wastes-at least once per 24 1 months:  ;

j. The performance of activities required by the Quality Assurance Program for effluent and environmental- monitoring at least once per.

12 months: ,

k. The Emergency Plan and implementing procedures at least once per 12- l months: -

1. The Security Plan and implementing procedures at least once per 12- ]

months; and  ;

m. Any other area of station operation considered appropriate by the ,

NSARC or the Senior Vice President.

j 4

SEABROOK - UNIT 1 6-10 Amendment No'. j

' ADMINISTRATIVE CONTROLS RECORDS 6.4.3.9 Records of NSARC activities shall be prepared and distributed as I indicated below:

a, . Minutes of each NSARC meeting shall be prepared and forwarded to

.the Senior Vice President within 30 daysLfollowing each meeting: I

b. Reports of reviews encompassed by Specification 6.4.2.7 shall be included in the minutes where applicable or forwarded under.sepa-rate cover to the Senior Vice President within 30 days following l completion of the review: 'and

c. Audit reports encompassed by Specification 6.4.2.8 shall be forwarded to the Senior Vice President and-to the management positions responsible for the areas audited within 30 days,after completion of the audit by the auditing organization.

6.5 REPORTABLE EVENT ACTION The following actions shall be taken for REPORTABLE EVENTS:

a. The-Commission shall be notified and a re) ort submitted pursuant to the requirements of Section 50.73 to 10 C:R Part 50, and

b. Each REPORTABLE EVENT'shall be reviewed by the 50RC and the results of this review shall be submitted to the NSARC and the Executive Director-Nuclear Production.

6.6 SAFETY LIMIT VIOLATION The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as;soon as aossible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Executive Director-iuclear Prodection and the NSARC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;

b. A Safety Limit Violation Re) ort shall be prepared. The report shall be reviewed by the S01C.. This report shall describe:

(1) applicable circumstances preceding the violation. -(2) effects of the violation upon facility components, systems, or . structures, and (3) corrective action taken to prevent recurrence:

c. The Safety Limit Violation Report shall be submitted to the Commission, the NSARC. and the Executive Director.-Nuclear Production within 14 days of the violation; and.

d. 0)eration of the station shall not be. resumed until authorized by t7e Commission.

SEABROOK - UNIT 1 6-11 Amendment No.

ADMINISTRATIVE' CONTROLS 6.7 PROCEDURES AND PROGRAMS

'6.7.1 Written procedures shall be established. implemented, and maintained covering the activities referenced below: '

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33. Revision 2. February 1978;

b. The emergency operating procedures required to implement the .

requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33:

c. Security Plan implementation:

d. Emergency Plan implementation: -;

e. PROCESS CONTROL PROGRAM implementation;

f. OFFSITE DOSE CALCULATION MANUAL implementation: .i

g. Quality. Assurance Program for effluent and environmental monitoring:

h. Fire Protection Program implementation; and

i. Technical Specification Improvement Program implementation.

6.7.2 The Station Manager may designate specific 3rograms and procedures to be reviewed in accordance with the Station Qualified Reviewer Program in lieu of.

review by the 50RC. The review accordance with Specification 6.per 4.2. the Qualified Reviewer Program shall be in ,

6.7.3 Programs and procedures listed in S)ecification 6.7.1, and changes -

thereto, shall be approved by the Station ianager or by cognizant Managers or '

i Directors who are designated as the Approval Authority by the Station Manager, as specified in administrative procedures. The Approval Authority for each a program and procedure shall be specified in administrative procedures.

6.7.4 Each procedure of S)ecification 6.7.1 ano changes thereto, shall be reviewed by the SORC and s1all be approved by the Station Manager, or be reviewed and approved in accordance with the-Station Qualified Reviewer Program, prior to implementation.

6.7.5 Changes to procedures of Specification 6.7.1 may be made prior to '$0RC _ l- I review provided:

a. The intent of the original procedure is not altered;

b. The change is approved by two members of the plant management--  :

staff, at least one of whom holds a Senior Operator license; and-l.:

. The change is documented, reviewed by the SORC and approved by the Station Manager, or reviewed and ap  !

Station Qualified Reviewer Programwithin proved 14indays accordance of with the  ;

implementation.

i SEABROOK - UNIT 1 6-12 Amendment No.

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ADMINISTRATIVE-CONTROLS PROCEDURES AND PROGRAMS 6.7.6 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of-systems outside containment that could contain highly _ radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the RHR and-containment spray, Safety Injection, chemical and volume control. -The program shall~ include the following: ,

1) Preventive maintenance and periodic visual inspection requirements, and

2) Integrated leak test requirements for each system at refueling cycle intervals or less.

b. In-Plant Radiation Monitorina A program that will ensure the capability to accurately determine ,

the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1) Training of personnel.  ;

2) Procedures for monitoring, and

3) Provisions for maintenance of sampling and analysis  ;

equipment, ,

c. Secondary Water Chemistry i A program for monitoring of secondary water chemistry-to inhibit steam generator tube degradation. This program shall include:

1) Identification of a sampling schedule for the critical' variables and control points for these variables.

2)- Identification of the procedures used to measure the values of the critical variables.

3) Identification of process sampling points, which shall ,

include monitoring the discharge of the condensate' pumps for evidente of condenser in-leakage. '

4) Procedures for the recording and management of data,.  :

5) Procedures defining corrective actions-for all off-control  !

point chemistry conditions, and  :

6) A prccedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events ~ required to initiate corrective action.

SEABROOK - UNIT 1 6-13 Amendment No.

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'hDMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) j

~d . Backuo Method far Determ'nina Subcoulina Marain  !

A program that will ensure the capability to-accurately monitor the 1 Reactor Coolant System subcooling margin. .This program shall include the following:

1) Training of personnel, and

2) Procedures for monitoring. l

e. Post-Accident Samnlina A program that will ensure the capability to-obtain arid analyze ,

reactor coolant, radioactive iodines and ) articulates in plant i gaseous effluents, and containment atmosplere sam)les under -  :

accident _ conditions. The program shall. include t7e following:

1) Training of personnel.

2) Procedures for sampling and analysis, and

3) Provisions for maintenance of sampling and analysis equipment,

f. Accident Monitorina Instrumentation l A program which will ensure the capability to monitor plant >

variables and systems operating status during and following an accident. This program shall include those instruments provided to indicate system operating status and furnish information regarding the release of radioactive materials (Category-2 and 3 instrumentation as defined in Regulatory Guide 1.97, Revision 3)* l and provide the following:

1) Preventive maintenance and periodic surveillance of-instrumentation,

2) Preplanned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and *

3) Administrative procedures for returning inoperable _.

instruments to OPERABLE status as soon as. practicable.

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• Seabrook: has taken exception to the categorization of instrumentation provided in Regulatory Guide 1.97, Revision 3. The Seabrook exceptions are provided in a FSAR Table 7.5-1. which has been reviewed by the NRC staff in SER Supplement i No. 5. '

SEABROOK - UNIT 1 6-14 Amendment No.~

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

noted.  !

STARTUP REPORT 6.8.1.1 A summary report of station startup and power escalation testing shall'  ;

be submitted following: (1) receipt of an Operating License. (2) amendment to the license involving a planned increase in power level. (3) installation of  ;

fuel that has a different design or has been manufactured by a different fuel supplier _. and (4) modifications that may have significantly altered the- ,

nuclear, thermal, or hydraulic performance of the station.

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SEABROOK - UNIT 1- 6-14A Amendment No.

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.: .4 :j 6D.MINISTRATIVE CONTROLS ANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 (Continued) to show conformance with 40 CFR Part 190. " Environmental Radiation Protection '

Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluent.s are given in Regulatory-Guide 1.109. Rev. 1. October 1977.

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

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The Annual Radioactive Effluent Release Report shall' include any changes l made during the reporting period to the PROCESS CONTROL PROGRAM and the 0DCM.

pursuant to Specifications 6.12 and 6.13, respectively, as well. as any major. '

change to Liquid. Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.14. It shall also include a listing of new locations for dose -

calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2. L The Annual Radioactive Effluent Release Re) ort shall also include the  ;

following: an explanation as to why the inoperaaility 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 limits.

of Specification 3.11.1.4. -

MONTHLY OPERATING REPORTS l

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: 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 REPORT 6.8.1.6.a Core operating limits shall be established and documented in the~  :

CORE 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.  ;

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

3. Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm I surveillance limit for Specification 3.1.1.3, i I

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-l Amendment No.'9, 23 SEABROOK - UNIT 1 6-18

4 IV. Safety Evaluation of License Amendment Reauest 93-20 Proposed Channes

• The purpose of License Amendment Request 93-20 is to propose changes to the Seabrook Station Technical Specifications to redefine the composition of the Station Operation Review Committee (SORC) based en experience and expertise and to delete the requirement that no more than two - t alternate SORC members shall participate as voting members; to allow implementation of a ,

Station Qualified Reviewer Program for the review and approval of designated Station procedures; to delete the requirement for periodic procedure reviews; to revise the requirement for issuing .

Nuclear Safety Audit Review Committee minutes from within fourteen days to within 30 days following each meeting; W to make numerous editorial changes. Editorial changes delete items that only applied prior to the plant exceeding 5% power or until completion of the first refueling outage, revise system / equipment names to be consistent with those used elsewhere.in the Seabrook Station Technical Specifications, clarify several minor inconsistencies involving Nuclear Safety Audit Review Committee function, composition, and use of attemates, correct the value used for Reactor Coolant System volume due to removal of the hot leg and cold leg resistance temperature detector manifolds, and delete the Bases for a Technical Specification requirement -

that was not included in the final draft of the Seabrook Station Technical Specifications.

The proposed change in SORC composition replaces specific position titles for SORC membership with the requirement that the SORC be comprised of individuals with specified experience and expertise. The required areas of expertise include all areas pertinent to plant operation, including Nuclear Power Plant Administrative Controls, Mechanical Maintenance, Electrical Maintenance, Instrumentation & Control, Chemistry, Health Physics, Operations, Technical Support / Engineering, and Reactor Engineering. The personnel selected will meet the requirements of ANSI N18.1c as stated in the Updated Final Safety Analysis Report. Each of these individuals will be appointed in writing by the Station Manager, who has the ultimate responsibihty for ensuring safe plant operation. This proposed change will continue to ensure that the SORC is capable of performing its safety oversight function and will eliminate the requirement to submit a Technical Specification -

change when organizational changes are made or titles are changed.

The proposed addition of a Station Qualified Reviewer Program will permit review of designated '

programs and procedures, that are required by Technical Specification 6.4.1.6a, to be performed -  ;

by Station Qualified Reviewers and approved by designated managers. These reviews would be in lieu of reviews by the SORC. Those programs and procedures that would be reviewed by a Station Qualified Reviewer in lieu of SORC will be designated in writing by the Station Manager.

Station' Qualified Reviewers and managers authorized to approve procedures will:also be designated in writing by the Station Manager. Procedures which require a 10CFR50.59 evaluation -

will continue to be reviewed by SORC. Qualified reviewers will be someone other than the individual who prepared the procedure, and procedures will receive a cross-disciplinary review.  ;

The purpose of this program is to reduce the amount of material that is currently being reviewed by SORC by eliminating less significant items that do not affect nuclear safety. This administrative change will continue to ensure that procedures are reviewed by experienced personnel who are quahfied to determine whether a program, procedure or change is technically adequate, and that '

any program, procedure or change that requires a 10 CFR 50.59 evaluation is reviewed by the SORC. .

The proposed deletion of the requirement to perform periodic procedure reviews will'not degrade  ;'

the quality of plant procedures. Seabrook Station administrative procedures specify that Station procedures will be reviewed at least every two years to determine if changes are necessary or-desirable. However, North Atlantic has determined that the biennial procedure reviews require significant expenditure of resources, impose a significant administrative burden, and do not1 significantly contribute to the quality of plant procedures. Most procedure changes are identified '

dunng the performance of other activities, such as plant design changes and operating experience reviews, and when the procedure is being used to perform plant evolutions. These existing l

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programmatic requirements will ensure that procedures are reviewed and revised when required, and make biennial procedure reviews redundant and unnecessary.

The proposed change to the current requirement in Technical Specification 6.4.2.9 to forward Nuclear Safety Audit Review Committee (NSARC) meeting minutes and reports to the Senior Vice President within 14 days following each meeting will not result in a reduction of plant safety. The current requirement places an unnecessary administrative burden on Seabrook Station staff since any significant safety problems found during NSARC review would be reported to senior management by the NSARC Chairman prior to the issuance of the meeting minutes or reports, with the meeting minutes used to transmit routine matters to the Senior Vice President and provide documentation of matters discussed during each meeting for future reference. The proposed change to forward the minutes or reports within 30 days following each meeting would have no effect on plant systems or operation and would not adversely impact nuclear safety.

The proposed editorial changes included in License Amendment Request 93-20 do not change any plant operating parameters or_ design features. In the case where the required time for a signal to reach the reactor trip breakers was changed from 1.2 seconds to 1.5 seconds (Bases for Technical Specification 2.2.1), the value is provided only as background in the Bases and the new value of 1.5 seconds is the correct value as listed in the Updated Final Safety Analysis Report (UFSAR). The other changes delete items that only applied prior to the plant exceeding 5% power .,

or until completion of the first refueling outage, revise system / equipment names to be consistent  ;

with those used elsewhere in the Seabrook Station Technical Specifications, correct the value used for Reactor Coolant System volume, clarify several minor inconsistencies involving NSARC function, composition, and use of altemates, and delete the Bases for a Technical Specification which was not included in the final draft of the Seabrook Station Technical Specifications. ,

All of the above changes are administrative in nature. There are no associated modifications to .

plant systems, structures, or components, nor is the design, function, or method of operation of :i any plant system, structure, or component altered.

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l V. Determination of Slanificant Hazards for License Amendment Request 93-20 Proposed )

Chanaes )

l (1) The proposed changes do not involve a significant increase in the probability or consequences of ,

an accident previously evaluated.

These changes are administrative in nature. They do not involve any modifications to plant systems and do not alter the method of operation of any plant equipment. The most significant '

changes involve (i) redefining the Station Operation Review Committee (SORC) composition based on required experience and expertise instead of by_ position title; (ii) implementing a Station -

Qualified Reviewer Program for the review of plant procedures that do not involve a 10 CFR 50.59 evaluation: (iii) deleting the requirement to perform periodic reviews of plant procedures; and (iv) '

changing the required time for preparing and forwarding minutes of Nuclear Safety Audit Review Committee (NSARC) meetings from within 14 days to within 30 days.

Redefining SORC composition based on experience and expertise will not result in a degradation of this committee's ability to perform its safety function. The Station Manager retains the responsibility for ensuring the safe operation of the plant. Personnel who are members of SORC will be individuals who have proven they possess the maturity, judgement, and technical knowledge in their area of expertise to meet the SORC responsibilities as defined in the Technical Specifications and will be designated in writing by the Station Manager. The required areas of expertise include all areas pertinent to plant operation, including Nuclear Power Plant t Administrative Controls, Mechanical Maintenance, Electrical Maintenance, Instrumentation &

Control, Chemistry, Health Physics, Operations, Technical Support / Engineering, and Reactor. ,

Engineering. This will ensure that the SORC's ability to advise the Station Manager on all matters related to nuclear safety meets or exceeds its present capabilities. f Implementing a Station Qualified Reviewer Program will not result in a degradation of the current level of procedure review. SORC will retain the responsibility for reviewing any document for ,

which a 10 CFR 50.59 evaluation is required. The Station Qualified Reviewer Program will be limited to reviewing procedures that do not affect nuclear safety. Personnel selected to be Station

• Qualified Reviewers will possess the technical experience and expertise to provide a thorough technical review as required by administrative procedures. These personnel and the managers authorized to approve these procedures will be designated in writing by the Station Manager.

Procedures or classes of procedures that can be reviewed per the Station Qualified Reviewer Program will be specified in writing by the Station Manager. Procedures will receive appropriate cross-disciplinary reviews when necessary.  ;

Deleting the requirement to perform periodic procedure reviews will not result in a degradation of plant procedures necessary for safe plant operation. The basis for deleting this requirement is that other programmatic requirements will ensure that procedures are reviewed and revised when ,

required (such as when plant ' design changes are implemented, when corrections or enhancements are identified during operating experience reviews, or when deficiencies are noted ,

during procedure implementation). These programmatic requirements make biennial procedure .

review redundant and unnecessary.

Extending the time allowed for preparing and forwarding the minutes of NSARC meetings to the  ;

Senior Vice President from within 14 days to within 3C days will not reduce plant safety. _The -

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NSARC is responsible for providing an independent rev4w and audit of activities specified in the Technical Specifications to ensure there are no unreview id safety questions resulting from Station  !

and engineering activities. The NSARC minutes serve es a means of documenting the reviews l performed by the NSARC and the determination as to whe her an unreviewed safety question was found to exist. If an unreviewed safety question, or any otner significant concern, were detected, the NSARC Chairman would be expected to promptly bring the matter to the attention of the Senior Vice President, without waiting for the issuance of the meeting minutes. Similarly, the  ;

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4 ee extension of the time for forwarding reports of NSARC reviews that are not forwarded with the NSARC meeting minutes will not impede prompt communication of significant concern to the Senior Vice President.

Editorial changes delete items that applied prior to the plant exceeding 5% power or until completion of the first refueling or second refueling outages, revise system / equipment names to be consistent with those used elsewhere in the Seabrook Station Technical Specifications, update the value in the Bases for Technical Specification 2.2.1 for the time required for a signal to reach the Reactor trip breakers, correct the value used for Reactor Coolant System volume due to removal of the hot jeg and cold leg resistance temperature detector _ manifolds, clarify several minor inconsistencies involving NSARC function, composition, and use of alternates, and delete the Bases for a Technical Specifiestion requirement that was not included in the final draft of the Seabrook Station Technical Specifications. None of these changes affects the design or operation of any plant system, structure or component. Since the changes proposed in this License Amendment Request do not revis=, existing plant structures, systems or components nor do they change the way the plant is operated or change the response of the plant to any transient or condition the proposed changes do not involve' a significant increase in the probabillty or consequences of an accident previously evaluated.

(2) The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed Technical Specification changes do not change the design or function or any plant structure, system or component, nor do they introduce any new failure modes. As stated above, -

the changes affecting SORC and NSARC will not degrade these committees' ability to perform their nuclear safety oversight function. The implementation of a Station Qualified Reviewer Program and the deletion of the requirement for periodic reviews of plant procedures will not degrade the quality of plant procedures. There are no modifications to plant structures, systems or components associated with these proposed changes, and the operation of plant equipment and systems remains unchanged. Since the changes proposed in this License Amendment Request -

do not revise existing plant structures, systems or components nor do they change the manner in which the plant is operated and in which it will respond to any design basis accident.the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.

(3) The proposed changes do not result in a significant reduction in the margin of safety.

The changes proposed in this License Amendment Request do not affect the ability of any system to perform its safety related function. As described above, these proposed changes are administrative in nature. They do not change any plant operating parameters or design features, and do not reduce the level of effectiveness of any existing administrative controls With the exception of correcting several minor editorial errors, there has been no change to the bases for any Technical Specification. Since none of the assumptions in the Technical Specification Bases are affected by the proposed changes presented in this License Amendment Request the margin of safety whico exists in the current Technical Specifications is not reduced.

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1 VI. Proposed Schedule for License Amendment issuance and Effectiveness North Atlantic requests NRC review of License Amendment Request 93-20 and issuance of a Ucense amendment having immediate effectiveness by July 14,1994.

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.w Vll. Environmental Impact Assessment-North Atlantic has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed changes do not involve a'significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, North Atlantic concludes that the proposed change meets the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmente! .npact Statement.

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