Text

Amend 34 to License NPF-86,revising App a to Specify Composition of Sorc,Implement Station Qualified Reviewer Program & to Revise Time Which Nuclear Safety Audit Review Committee Must Issue Repts & Minutes
	ML20078H470
Person / Time
Site:	Seabrook
Issue date:	01/26/1995
From:	Mckee P; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20078H473	List: ML20078H470; ML20078H480; ML20078H495;
References
	NUDOCS 9502060221
	Download: ML20078H470 (50)
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~' *g UNITED STATES g

g NUCLEAR REGULATORY COMMISSION I

WASHINGTON, D.C. 20066 4 01 o%

/

NORTH ATLANTIC ENERGY SERVICE CORPORATION. ET AL*

DOCKET NO. 50-443 SEABROOK STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 34 License No. NPF-86 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by North Atlantic Energy Service Corporation, et al. (the licensee), dated Januo 14, 1994, as modified by letter dated October 17, 1994, com;I tes with the standards and requirements of the Atomic Energy of 1954, as amended (the Act), and the Comission's rules anc regulations set +

forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in comp 1hoce with the Comission's regulations; t

D.

The issuance of this amendment will not be inimical to the comon defense and se:urity or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

[

North Atlantic Energy Service Company (NAESCO) is authorized to act as agent for the: North Atlantic Energy Corporation, Canal Electric Company, The Connecticut Light and Power Company, Great Bay Power Corporation, Hudson Light and Power Department, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Power Company, New Hampshire Electric Cooperative, Inc., Taunton Municipal Light Plant, and The United Illuminating Company, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

9502060221 950126 PDR ADOCK 05000443 P

PDR

.,. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-86 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 34, and the Environmental Protection Plan contained in Appendix B are incorporated into Facility License No.

NPF-86. NAESCO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 2

NhillipF.kcKee, Director Project Directorate I-4 Division of Reactor Projects - I/II j

Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical Specifications Date of Issuance: January 26, 1995 l

o ATTACHMENT TO LICENSE AMENDMENT NO. 34 FACILITY OPERATING LICENSE NO. NPF-86 DOCKET N0. 50-443 Replace the following pages of Appendix A, Technical Specifications, with the attached pages as indicated.

The revised pages are identified by amendment number and contain vertic31 lines indicating the areas of change. Overleaf pages have been provided.

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6-5 6-5 6-6 6-6 6-7 6-7 6-8 6-8 6-8A 6-88 6-9 6-9 6-10 6-10 6-11 6-11 6-12 6-12 6-13 6-13 6-14 6-14 6-14A

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION E8K 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4 9-4 3/4.9.5 C0fWUNICATIONS.....................

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 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM.....

3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL..............

3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE P0OL...............

3/4 9-12 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM 3/4 9-13 3/4.9.13 SPENT FUEL ASSEMBLY STORAGE 3/4 9-16 FIGURE 3.9-1 FUEL ASSEMBLY BURNUP VS. INITIAL ENRICHMENT FOR SPENT FUEL ASSEMBLY STORAGE..........

3/4 9-17 3/4.9.14 NEW FUEL ASSEMBLY STORAGE 3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN 3/410-1 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 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN 3/4 10-5 1

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration 3/4 11-1 Dose..........................

3/4 11-2 Liquid Radwaste Treatment System............

3/4 11-3 Liquid Holdup Tanks 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS Dose Rate 3/4 11-5 Dose - Nobl e Gases...................

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

3/4 11-7 Gaseous Radwaste Treatment System 3/4 11-8 Explosive Gas Mixture - System.............

3/4 11-9 3/4.11.3 SOLID RADIOACTIVE WASTES................

3/4 11-10 3/4.11.4 TOTAL DOSE.......................

3/4 11-12 3/4.12 DADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM...................

3/4 12-1 SEABROOK - UNIT 1 ix Amendment No. 34

o -

lhD.LE LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PEE 3/4.12.2 LAND USE CENSUS 3/4 12-3 3/4.12 J INTERLABORATORY COMPARISON PROGRAM...........

3/4 12-5 3.0/4.0 BASES 3/4.0 APPLICABILITY...................:

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL....................

B 3/4 1-1 3/4.1.2 BORATION SYSTEMS....................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...............

B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHAKNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATIO B 3/4 2-3 3/4.2.5 DNB RARAMETERS.....................

B 3/4 2-4 3/4.3 INSTRUMENTATION i

3/4.3.1 and 3/4.3.2 REAC10R TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION.

B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.........,....

B 3/4 3-6

)).1,4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B 3/4 4-1 3/4.4.2 SAFETY VALVES B 3/4 4-1 3/4.4.3 PRESSURIZER B 3/4 4-2 3/4.4.4 RELIEF VALVES B 3/4 4-2 3/4.4.5 STEAM GENERATORS....................

B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.............

B 3/4 4-3 3/4.4.7 CHEMISTRY B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LINITS B 3/4 4-7 FIGURE E J/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE................

B 3/4 4-9 FIGURE B 3/4.4-2 (This figure number not used).........

B 3/4 4-10 l

i SEABROOK - UNIT 1 x

Amendment No. 19

INDEX 5.0 DESIGN FEATURES SECTION f.ag L3 REACTOR CORE i

i 5.3.1 FUEL ASSEMBLIES 5-9 5.3.2 CONTROL R00 ASSEMBLIES.................

5-9 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE 5-9 5.4.2 VOLUME....................

5-9 5.5 METEOROLOGICAL TOWER LOCATION 5-9

)

5.6 FUEL STORAGE 5.6.1 CRITICALITY 5-10 5.6.2 DRAINAGE........................

5-10 5.6.3 CAPACITY........................

5-10 1

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5-10 1

TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LINITS........

5-11 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY.......................

6-1 i

6.2 ORGANIZATION........................

6-1 j

6.2.1 0FFSITE AND ONSITE ORGANIZATIONS............

6-1 6.2.2 STATION STAFF 6-2 FIGURE 6.2-1 (This figure number is not used) 6-3 FIGURE 6.2-2 (This figure number is not used) 6-3 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION...........

64 6.2.3 IIEEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function........................

6-5 Composition 6-5 Responsibilities....................

6-5 Records 6-5

)

6.2.4 SHIFT TECHNICAL ADVISOR 6-5 6.3 TRAlhING..........................

6-5 SEABROOK - UNIT 1 xiii

INDEX 6.0 ADMINISTRATIVE CONTROLS SECTION EAGE 6.4 REVIEW AND AUDIT......................

6-6 6.4.1 STATION OPERATION REVIEW ComITTEE (50RC)

Function........................

6-6 Composition 6-6 Alternates.......................

6-6 Meeting Frequency 6-6 j

Quorum.........................

6-6 1

Responsibilities....................

6-7 Records 6-8 6.4.2 STATION QUALIFIED REVIEWER PROGRAM...........

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

6-8 Responsibilities....................

6-8 Records 6-BA Training and Qualification...............

6-8A 6.4.3 NUCLEAR SAFETY AUDIT REVIEW ComITTEE (NSARC)

Function........................

6-88 Composition 6-88 Alternates.......................

6-88 Consultants 6-88 Meeting Frequency 6-8B Quorum.........................

6-8B Review.........................

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

6-10 l

Records 6-11 6.5 REPORTABLE EVENT ACTION 6-11 6.6 SAFETY LINIT VIOLATION...................

6-11 6.7 PROCEDURES AND PROGRAMS 6-12 68 REPORTIE REQUIREMENTS 6.8.1 RIN INE REPORTS 6-14A Startup Report.....................

6-14A Annual Reports.....................

6-15 Annual Radiological Enviror. mental Operating Report...

6-15 Annual Radioactive Effluent Release Report.......

6-17 Monthly Operating Reports 6-18 CORE OPERATING LIMITS REPORT..............

6-18 6.8.2 SPECIAL REPORTS 6-19 SEABROOK - UNIT 1 xiv Amendment No. 34

INDEX 6.0 ADMINISTRATIVE CONTROLS SECTION E&GE 6.9 RECORD RETENTION......................

6-19 6.10 RADIATION PROTECTION PROGRAM 6-20 6.11 HIGH RADIATION AREA....................

6-20 6.12 PROCESS CONTROL PROGRAM (PCP) 6-21 6.13 0FFSITE DOSE CALCULATION MANUAL (00CM) 6-22 I

6.14 MAJOR CHANGES TO LIQUID. GASEOUS. AND SOLIO RADWASTE TREATMENT SYSTEMS 6-23 l

l l

i i

l SEABROOK - UNIT 1 xv Amendment No. 34 l

DEFINITIONS PROCESS CONTROL PROGRAM 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 processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State Regulations, burial ground requirements, and other requirements governing the disposal of radioactive waste.

PURGE - PURGING 1.26 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinament 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 POWCR 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.

l RATED THERMAL POWER 1.28 RATED THERMAL POWER shall be a total recctor 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 when the monitored parameter exceeds its Trip Setpoirt 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.

CONTAINMENT ENCLOSURE BUILDING INTEGRITY 1.31 CONTAlletENT ENCLOSURE BUILDING INTEGRITY shall exist when:

a.

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

The Containment Enclosure Emergency Air Cleanup System is OPERABLE, l

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. 34 u

1 DEFINITIONS SHUTDOWN MARGIN i

1.32 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l

l the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.33 The SITE BOUNDARY shall be that line beyond which the owned, nor leased, nor etherwl e controlled by the licensee., land is neither l

1 SLAVE RELAY TEST 1.34 A SLAVE RELAY TEST shall be the energization of each slave relay and l

verification of CPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION 1.35 SOLIDIFICATION shall be the conversion of wet wastes into a form that l

meets shipping and burial ground requirements.

SOURCE CHECK 1.36 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.37 A STAGGERED TEST BASIS shall consist of:

l a.

A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

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

THERMAL POWER 1.38 THElWWIL POWER shall be the total reactor core heat transfer rate to the l

reactor coolant.

. TRIP ACTUATING DEVICE OPERATIONAL TEST 1.39 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the l

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

SEABROOK - UNIT 1 1-6 Amendment No. 9

LIMITING SAFETY SYSTEM SETTINGS i

BASES I

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Undervoltaae and Underfrecuency - Reactor Coolant P-Busses,

The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DN8 as a result of complete loss of forced coolant flow. 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 trip of two or s. ore 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 tn reach the Reactor i

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

exceed 0.3 second. On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbir.e 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 Trin A Turbine tris initiates a Reactor trip. On decreasing power, the Reactor trip from tie 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 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 upon any signal which initiates a Safety Injection.

The ESF instrumentation channels that initiate a Safety Injection signal are l

shown in Table 3.3-3.

Reactor Trio System Interlocks 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 decreasing power, Source Range Level trips are automatically reactivated and high 'toltage is restored.

SEABROOK - UNIT 1 B 2-7 Amendment No. 34

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

Reactor Trio System Interlocks (Continued)

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pum bus undervoltage and underfrequency, pressurizer low pressure, pand 1

pressurizer high level. On decreasing power, the above listed trips are automatically blocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, i

the P-8 automatically blocks the above trip.

P-g On increasing power, P-g automatically enables Reactor trip on Turbine trip. On decreasing power, P-g automatically blocks Reactor trip on Turbine trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the l

Source Range high voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated. Provides input to P-7.

P-13 Provides input to P-7.

i 4

SEABROOK - UNIT 1 B 2-8

TABLE 4.3-1 (Continued).

TABLE NOTATIONS (Continued)

(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 Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circult(s).

(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.

(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.

SEABROOK - UNIT 1 3/4 3-13 Amendment No. 34

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR_ADERATION.

3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Set oints set consistent with the values shown in the Trip Setpoint column of Tab e 3.3-4.

AEELI. ABILITY: As shown in Table 3.3-3.

C ACTION:

With an ESFAS Instrumentation or Interlock Trip Setpoint trip less a.

conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value colemn of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.

b.

With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Allowable Value column of Table 3.3-4, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + S s TA Where:

Z = The value from Column Z of Table 3.3-4 for the affected

channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S - Either the "as measured" value (in percent span) of the :;ensor error, or the value from Column S (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.

c.

With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

SEABROOK - UNIT 1 3/4 3-14

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.

APPLICABILIJ1: MODES 1, 2, and 3.

ACTION:

i 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 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The provisions of Specification 3.0.4 are not applicable.

3 b.

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

less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , 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 Specification 6.8.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.

SURVEILLANCE 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 CHANNEL CALIBRATION.

SEABROOK - UNIT 1 3/4 3-49 Amendment No. 34

g TABLE 3.3-10 E

ACCIDENT MONITORING INSTRUMENTATION 8

TOTA!.

MINIMUM NO. OF CHANNELS E

INSTRUMENT CHANNELS OPERABLE 1.

Conteineent Pressure a.

Normal Range 2

1 b.

Extended Range 2

1 2.

Reactor Coolant Outlet Temperature - T. (Wide Range) 4 2

l l

3.

Reactor Coolant Inlet Temperature - T. (Wide Range) 4 2

4.

Reactor Coolant Pressure - Wide Range 2

1 5.

Pressurizer Water Level 2

1 h

6.

Steam Generator Pressure 2/ steam generator 1/ steam generator 7.

Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator 8.

Steam Generator Water Level - Wide Range 1/ steam generator 1/ steam generator 9.

Refueling Water Storage Tank Water Level 2

1

10. Reactor Coolant System.'iocooling Margin Monitor 2

1

11. Containment Building Water Level 2

1

12. Core Exit Thermocouples 4/ core quadrant 2/ core quadrant

13. Containment Post-LOCA Area Monitor 2

1 4

e m___

___ _ _ m.

__m___

m

' ~

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATIQN LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation.

AFPLICABILITY: MODES 1 and 2.*

ACTION:

With less than the above required reactor coolant loops in operation, be in at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SEVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

See Special Test Exceptions Specification 3.10.4 SEABROOK - UNIT 1 3/4 4-1

i REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS'AND COOLANT CIRCULATION HOT STANDBY l

l 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 1

Reactor Trip System breakers are open:*

a.

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

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.

l 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within I hour 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 System and immediately initiate corrective action to return the required reactor coolant loop to operation.

i

All reactor coolant pumps may be deenergized for up to I hour provided: (1) no operations are perinitted 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.

I SEABROOK - UNIT 1 3/4 4-2 Amendment No. 34

CONTAINMENT 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 i

'*st ifying that on a Phase "A" Isolation test signal, each Phase "A" l

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 l

c.

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

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

SEABROOK - UNIT 1 3/4 6-17 Amendment No. 34

l CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN MONITORS LIMITING ColeITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen monitors shall be OPERABLE.

APPLICABILITY: MODES I and 2.

ACTION:

a.

With one hydrogen monitor inoperable, restore the inoperable monitor 1

to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With both hydrogen monitors inoperable, restore at least one monitor i

to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l i

SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an ANALOG CHANNEL OPERATIONAL TEST at least once per 92 days, and at least once each refueling interval by l

performing a CHANNEL CALIBRATION using sample gas containing:

a.

One volume percent hydrogen, balance nitrogen; and b.

Four volume percent hydrogen, balanca nitrogen.

SEABROOK - UNIT 1 3/4 6-18 Amendment No. 30

fjh!T SYSTEMS TURBINE CYCLE r

AUXILIARY FEEDWATER SYSTEM LIMITING CofEITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps I

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 volume 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.

l ACTION:

a.

With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be l

in at least HOT STAN)BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With two emergency feedwater pumps inoperable restore at least one emergencyfeedwaterpumptoOPERABLEstatuswlthin12hoursand i

restore both emergency feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD $HUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

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

d.

With three auxiliary feedwater pumps inoperable, inmediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

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

a.

At least once per 31 days by:

1)

Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; 2)

Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER; and SEABROOK - UNIT 1 3/4 7-3 Amendment No. 34

^ '

PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.la.

(Continued),

3)

Verifying that valves FW-156 and FW-163 are OPERABLE for l

alignment of the startup feedwater purap to the emergency feedwater header.

b.

At least once per 92 days on a STAGGERED TEST BASIS by:

l 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 gpa 2)

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

I 3)

Verifying that the startup feedwater pump develops a discharge pressure of greater than or equal to 1375 psig at a flow of greater than or equal to 425 gpe; c.

At least once per 18 months during shutdown by:

l 1)

Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater System Actuation test signal; 2)

Verifying ths.t each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal; 3)

Verifying that with all manual actions, including power source and valve alignsent, the startup feedwater pump starts within the required elapsed time; and 4)

Verifying that each emergency feedwater control valve closes on receipt of a high flow test signal.

SEABROOK - UNIT 1 3/4 7-4 Amendment No. 30

SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual full-length shutdown and control rod drop time measurements provided; a.

Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and b.

The rod position indicator is OPERABLE during the withdrawal of the rods.*

J APPLICABILITY:

MODES 3, 4, and 5 during performance of rod drop time measurements.

ACTION:

With the Position Indication Systems inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers.

i SURVEILLANCE REOUIREMENTS 4.10.5 The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps when the rods are stationary.

This requirement is not applicable during the initial calibration of ne Digital Rod Position Indication System provided: (1) k is maintained less than or equal to 0.95, and (2) only one shutdown,,r, control rod bank o

is withdrawn from the fully inserted position at one time.

SEABROOK - UNIT 1 3/4 10-5 l

k PAGE LEFT INTENTIONALLY BLANK l

l l

l SEABROOK - UNIT 1 3/4 10-6 Amendment No. 34

REACTIVITY CONTROL SYSTEMS BASES i

3/4.1.2 80 RATION 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 ppe borated water from the boric acid storage tanks or a minimum contained volume of 24,500 gallons of 2000 ppe 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 Boron Thermal Regeneration System and the Reactor Water Makeup System in Modes 4, 5, l

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

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that:

(1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210, and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position Indication System does not indicate i

the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position.

The ACTION 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 measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation.

In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

SEABROOK - UNIT 1 B 3/4 1-3 Amendment No. 34

REACTIVITY CONTROL SYSTEMS j

BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued)

The maximum rod drop time restriction is consistent with the assumed rod

. drop time used in the safety analyses. Measurement with rods at their individual mechanical fully withdrawn position, T greater than or equal to 551*F and all reactor coolant pumps operating ens 7u es that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

The fully withdrawn position of shutdown and control banks can be varied between 225 and the mechanical fully withdrawn position (up to 232 steps),

inclusive. An engineering evaluation was performed to allow operation to the 232 step maximum. The 225 to 232 step interval allows axial repositioning to minimize RCCA wear.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more fre-quent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod (s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rahe (Westinghouse) to C. O. Thomas (NRC). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism.

In the event the plant is unable to verify the rod (s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a.

Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for this verification.

E SEABROOK - UNIT 1 B 3/4 1-4 Amendment No. 8 i

l

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES AGJMATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Sur-veillance Frequencies and Out of Service Times for the Reactor Protection In-strumentation System," and supplements to that report. Surveillance intervals and out of service times were determined based on raintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

(Implementation of quarterly testing of RTS is being postponed until after approval of a similar testing interval for ESFAS.) The NRC Safety Evaluation Report for WCAP-10271 was provided in a letter dated February 21, 1985, from C. O. Thomas (NRC) to J. J. Sheppard (WOG-CP&L).

The Engineered Safety Features Actuation System Instrumentation Trip Set-points specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To -a==adate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.

Opera-tion with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to acconmiodate this error. An optional provision has been included for j

determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other SEABROOK - UNIT 1 B 3/4 3-1 m---

+wr

--ma

~ '

INSTRUMENTATION 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 t

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. 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.

Ir.herent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.

Sensor and rack instrumentation util' zed 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.

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

Sensor response time verification may be demonstrated by either:

(1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time.

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

If they are, the signals are combined into logic matrices 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 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 1

SEABROOK - UNIT 1 B 3/4 3-2 Amendment No. 34 J

-c--

.-n

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

M ACTOR 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 Reactor 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 excess 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 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.

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

l 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 I gpe. 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.

SEABROOK - UNIT 1 B 3/4 4-5 Amendment No. 34

REACTOR COOLANT SYSTEM BASES I

3/4.4.8 SPECIFIC ACTIVITY (Continued)

The sample analysis for determining the gross specific activity and E can exclude the radiciodines because of the ' ow reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radiciodine level is to be determined every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radiciodine contribution would be approximately.15.

In a release of reactor coolant with a typical mixture of radioactivity, the actual radiciodine contribution would probably be about 205. The exclusion of radionuclides with half-lives less than 10 minutes from these determinations has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes' decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident

{

condition.

Based upon the above considerations for excluding certain radionuclides from the

- sample analysis, the allowable time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between sample taking and completing the initial analysis is based upon a typical time necessary to per-form the sampling, transport the sample, and wrform the analysis of about 90 minutes. After 90 minutes, the gross count s sould be made in a reproducible geometry of suple and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy.

It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample l

within typical counting basis following sampling of less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , about I day, about I week, and about 1 month.

l Reductag T to less than 500*F prevents the release of activity should a l

steam generator b rupture, since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.

The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

SEABROOK - UNIT 1 B 3/4 4-6 I

3/4.10 SPECIAL 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 or equal to 5% of RATED THERMAL POWER with the RCS T,y, 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-teristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications.

For i

instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not j

normally be allowed by Specification 3.1.3.6 and the RCS T may be below the minimum temperature of Specification 3.1.1.4 during the me,as,urement.

y 3/4.10.4 REACTOR COOLANT LOOPS l

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.

E4,10.5 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.

SEABROOK - UNIT 1 B 3/4 10-1 Amendment No. 34

,o 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 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with a zirconium alloy.

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.

CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 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.

t 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 5.4.2 The total water and steam volume of the Reactor Coclant System is 12,255 l

cubic feet an a nominal T, 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. 34

DESIGN FEATURES 5.6 FUEL STORAGE i

CRIT ICALITY 5 6.1.1 The spent fuel storage racks are designed and shall be maintained with:

A k, ted water, which includes margin for uncertainty inequivalent to a.

g unbora calculation methods and mechanical tolerances with a 955 probability at a 955 confidence level.

b.

A nominal 10.35 inch center-to-center distance between fuel assemblies placed in the storage racks.

5.6.1.2 The new fuel storage racks are designed and shall be maintained with:

a.

A k, equivalent to less than or equal to 0.95 when flooded with g

unborated water, which includes margin for uncertainty in calculational vethods and mechanical tolerances with a 955 probability at a 955 confidence level.

b.

A k, tion is assumed, which includes margin for uncertainty inequiva g

modera calculational rethods and mechanical tolerances with a d55 probability at a 955 confidence level.

c.

A nominal 21 inch center-to-center distance between fuel assemblies placed in the storage racks.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the peal below elevation 14 feet 6 inches.

CAPACITY 5.6.3 The spent fuel storage pool is uesigned and shall be maintained with a storage capacity limited to no more than 1236 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

l l

SEABROOK - UNIT 1 5-10 Amendment No. 6

,o ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1. The ISEG shall function to examine station operating characteristics, NRC issuances. Industry advisories, Licensee Event Reports, and other sources of station design and operating experience inforisation, including units of similar design, which may indicate areas for improving station safety. The ISEG shall make detailed recommendations for revised procedures, equipment f

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 er 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 fleid.

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 **duced as much as practical.

RECORDS 6.7.3.4 Records of activities performed by the ISEG shall be prepared, main-tahed, 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 Room Connander in the areas of thern,a1 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.%/ acement licensed training program for the station l

staff shall be maintained eder the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and the supplemental requirements specified in NUREG-1021, and shall include familiarization with relevant industry operational experience.

j l

Not responsible for sign-off function.

l SEABROOK - UNIT 1 6-5 Amendment No. 34 i

l

ADMINISTRATIVE CONTROLS 6.4 REVIEW AND AUDIT 2

6.4.1 STATION OPERATION REVIEW COP 94ITTEE (SORC)

EUNCTION 6.4.1.1 The SU,1C shall functior to advise the Station Manager on all natters related to nuclear safety.

COMPOSITION 6.4.1.2 The 50RC 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 Health Physics Operations l

Technical Support / Engineering Reactor Engineering The Station Manager shall serve as Chairman of the SORC and shall appoint the SORC members in writing. Members shall have a minimum of eight years power plant experience of which a minimum of three years shall be nuclear power experience. At least one member shall have an SR0 license for Seabrook Station.

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 FRE0VENCY 6.4.1.4 The SORC shall meet at least once per calendar month and ar convened by the SORC Chairman or one of his designated alternate (s).

l QUORUM 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 sufficient SORC members including alternates to equal at least 50 percent of the SORC composition.

SEABROOK - UNIT 1 6-6 Amendment No. 34

s-ADMINISTRATIVE CONTROLS RESPONSIBILITIES 6.4.1.6 The 50RC shall be responsible for:

a.

Review of:

(1) all proposed procedures required by Specification 6.7 and changes thereto, (2) all proposed programs required by Specification 6.7 and changes thereto, and (3) any other proposed procedures or changes thereto as-detemined by the Station Manager to affect nuclear safety. Procedures and programs required by Specification 6.7 that are designated for rev< ew and approval by the Station Qualifted Reviewer Program in accordance with Specification 6.4.2 do not require 50RC review.

b.

Review of all proposed tests and experiments that affect nuclear safety; c.

Review of all proposed changas to Appendix "A" Technical Specifications; d.

Review of all proposed changes or modifications to station systems or equipment that affect nuclear safety; Investigation of all violations of the Technical Specifications.,

e.

including the preparation and fomarding of reports covering evaluation 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 EVENTS; 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 recommended Security Plan changes to the NSARC; l

j.

Review of the Emergency Plan and implementing procedures and submittal of recommended Emergency Plan changes tr. the NSARC; l

k.

Review of any accidental, unplanned, or' uncontrolled radioactive release including the preparation of reports covering evaluation, recommandations, 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; 1.

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

Review of the Fire Protection orogram and implementing instructions and submittal of recommended Fire Protection Program changes to the l

NSARC.

SEABROOK - UNIT 1 6-7 Amendment No. 34

ADMINISTRATIVE CONTROLS 6.4.1.7 The SORC shall:

a.

Recosamnd 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 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Executive Director - Nuclear Production and the NSARC of disagreement between the S0RC 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 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 FUNCT[Qlf 6.4.2.1 The Station Manager may establish' a Station Qualified Reviewer Program whereby required reviews of designated procedures or classes of procedures required by Specification 6.4.1.6.a are performed by Station Qualified Reviewers and approved by designated 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 50RC.

RESPONSIBILITIES 6.4.2.2 The Station Qualified Reviewur Program shall:

a.

Provide for the review of designated procedures, programs, 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 procedures, programs, 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 Revieur(s) in affected disciplines, or by other persons designated by cognizant Managers or Directors as having specific expertise i

required to assess a particular procedure, progree or change.

Cross-disciplinary reviewers may function as a consiittee.

SEABROOK - UNIT 1 6-8 Amendment No. 34

t*

ADMINISTRATIVE CONTROLS 4

d.

Provide for a screening of designated procedures, programs 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

.unreviewed safety question does not exist. This screening will be performed by personnel trained and qualified in performing 10 CFR 50.59 evaluations.

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

the responsible Manager for approval or disapproval of procedures and programs considered under procedure or program was scree. Specification 6.4.1.6a and that the ned by a qualified, individual and found not to require a 10 CFR 50.59 evaluation.

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, program, or change will thou be forwarded with the 10 CFR 50.59 evaluation to 50RC for review.

6.4.2.4 Personnel recosmiended to be Station Qualified Reviewers shall be designated in writing by the Station Manager for each procedure, program, or class of procedure or program within the scope of the Station Qualified Reviewer Program.

6.4.2.5 Temporary procedure changes shall be made in accordance with Specification 6.7.3 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 procedures and programs performed under the Station Qualified Reviewer Program shall be documented in accordance with administrative procedures.

TRAINING AND QUALIFICATION 6.4.2.7 The training and qualification requirements af personnel designated as a Qualified Reviewer in accordance with the Station Qualifted Reviewer Program shall be in accordance with administrative procedures. Qualified reviewers shall have:

a.

A Bachelors degree in engineering, related science, cr technical discipline, and two years of nuclear power plant experience; OR b.

Six years of nuclear power plant experience; OR c.

An equivalent combination of education and experience as approved by a Department Supervisor.

SEABROOK - UNIT 1 6-8A Amendment No. 34 8

5 ADNINISTRATIVE CONTROLS 6.4.3 NUCLEAR SAFETY AUDIT REVIEW COMITTEE (NSARC)

EUNCTION 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 President on those areas of responsibility specified in Specifications 6.4.3.7 and 6.4.3.8.

C(MPOSITION l'

6.4.3.2 The NSARC shall be composed of at least five (5) individuals. The Chairman, Vice Chairman and members, including designated alternates, shall be appointed in writing by the Senior Vice President. Collectively, the individuals appointed to the NSARC should have experience and expertise in the following areas:

a.

Nuclear power plant operations, b.

Nuclear engineering, c.

Ck nistry ar.d radiochemistry, d.

Metallurgy, e.

Instrumentation and control, i

f.

Radiological safety, 2

g.

Mechanical and electrical engineering, and h.

Quality assurance practices.

f Each member shall meet the qualifications of ANSI 3.1-1978, Section 4.7.

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

President to serve on a temporary basis; however, no more than a minority shall participate as voting members in NSARC activities at any one time.

i CONSULTANTS 6.4.3.4 Consultants shall be utilized as determined by the NSARC to provide l

expert advice to the NSARC.

l MEETING FRE00ENry 6.4.3.5 The NSARC shall meet at least once per 6 months i 6 weeks.

l t

000RLM 6.4.3.6 The quorum of the NSARC necessary for the performance of the NSARC l

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 Chairmart, or his designated alternate, can participate as an NSARC member whan the Chairman is in attendance.

SEABROOK - UNIT 1 6-88 Amendment No. 34 L

e

1

/

ADMINISTRATIVE CONTROLS REVIEW 6.4.3.7 The tiSARC shall b6 responsible for the review of:

l a.

The safety evaluations for:

(1) changes to procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.5g, to verify that such actions did not constitute an unreviewed safety question; b.

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

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 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 aspect of design or operation of structures, systems, or components that could affect nuclear safety; and 1.

Reports and meeting minutes of the 50RC.

l SEABROOK - UNIT 1 6-9 Amendment No. 34

s g

ADMINISTRATIVE CONTROLS i

AUDITS 6.4.3.8 Audits of station activities shall be performed under the cognizance l

of the NSARC. The audits shall be performed within the specified time interval with a r.aximum allowable extension not to exceed 25% of the specified interval 1

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 conditions at least once per 12 months; b.

The perfoneance, training, and qualtfications 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 Ap p adix 8, 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 equipm6nt and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection i

consultant. An outside independent fire protection consultant i

shall be used at least every third year; g.

The Radiological Environmental Monitoring Progree and the results thereof at ' east once per 12 months; h.

The crFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months; i.

The Pd0 CESS CONTROL PROGRAN and implementing procedures for processing 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; i

k.

The Emergency Plan and icaplementing procedures at least once per 12 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 NSAP.C or the Senior Vice President.

SEABROOK - UNIT 1 6-10 Amendment No. 34

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ADMINISTRATIV[_C.ONTROLS ELLDBQS 6.4.3.9 Records of NSARC activities shall be prepared and distributed as l

indicated below:

a.

Minutes of each NSARC meeting shall be prepared and forwarded to the Senior Vice President within 30 working days following each

{

meeting; 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 working days l

follwing 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 t

positions responsible for the areas audited within 30 days after comoletion of the audit by the auditing organization.

fuS 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 REPORTA8LE EVENT shall be reviewed by the 50RC and the results l

of this review shall be submitted to the NSARC and the Executive l

Director-Nuclear froduction.

126 SAFETY LINIT VIOLATION The following actions shall be taken in the event a safety Limit is l

violated:

a.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within I hour. The Executive Director-Nuclear Production and the NSARC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; b.

A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SORC. 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 tree violation; and d.

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

SEABROOK - UNIT 1 6-11 Amendment No. 34

i.

ADMINISTRATIVE CONTROLS 6.7 PROCEDt**% AND PROGRAMS 6.7.1 Writi procedures shall be established, iaplemented, and maintained covering the activities referenced below:

The applicable procedures recommended in Appendix A of Regulatory a.

Guide 1.33, Revision 2, February 1978; b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement I 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; h.

Fire Protection Program implementation; and 1.

Technical Specification Improvement Prograin implementation.

6.7.2 The Station Manager may designate specific procedures and programs or classes of procedures and programs to be reviewed in accordance with the Station Qualified Reviewer Program in lieu of review by the 50RC. The review per the Qualified Reviewer Program shall be in accordance with Specification 6.4.2, 6.7.3 Procedures and programs listed in Specification 6.7.1 and changes thereto, shall be approved by the Station Manager or by cognizant Manager or Directors who are designated as the Approval Authority by the Station Manager, as specified in administrative procedures. The Approval Authority for each procedure and program or class of procedure and progra shall be specified in administrative procedures.

6.7.4 Each procedure of Specification 6.7.1, and changes thereto, shall be reviewed by the 50RC and shall be approved by the Station Manager, or be reviewed and approved in accordan::e with the Station Qualified Reviewer Program, prior to implementation. Each procedure of Specification 6.7.1 shall be reviewed periodically as set forth in administrative procedures.

6.7.5 Changes to procedures of Specification 6.7.1 may be made prior to SORC review prov ded:

a.

The intent of the original procedure is not altered; 4

b.

The change is aoproved by two members of the plant management staff, at least one of whom holds a Senior Operator license; and l

c.

The cl.&nge is documented, reviewed by the 50RC and approved by the Station Manager, or reviewed and approved in accordance with the Station Qualified Reviewer Program, within 14 days of implementation.

SEABROOK - UNIT 1 6-12 Amendment No. 34

,J ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRat$

6.7.6 The following programs shall be established, implemented, and l

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 A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

l l

1)

Identification of a sampling schedule for the critical I

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 nonitoring the discharge of the condensate pumps for evidence of condenser in-leakage, 4)

Procedures for the recording and aanagement of data, 5)

Procedures defining corrective actions for all off-control l

point chemistry conditions, and 6)

A procedure identifying:

(a) the authority responsible for the interpretation of the data, and to)d to initiatethe sequence and timing of administrative events require corrective action.

SEABROOK - UNIT 1 6-13 Amendment No. 34

s g, ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) l d.

Backun Method for Determinino Subcoolina Marcin A program that will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

l 1)

Training of personnel, and 2)

Procedures for monitoring.

e.

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

1)

Training of personnel, 2)

Procedures for sampling and analysis, and 1

3)

Provisions for maintenance of sampling &nd analysis equipment.

f.

Accident Monitorina Instrumentation l

A program which will ensure the ct.pability to monitor plant variables and systems operatinti status durin and following an accident. This program shall nelude those nstruments provided to indicate system cperating 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 tne following:

1)

Preventive maintenance and periodic surveillance of instrumentation, 2)

Preplanned operating procedures and ba:kup instrumentation tc be used if one or more monitoring instruments become inoperable, and j

3)'

/491nistrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

d

Seabrook has taken exception to the categorization of instrumentation provided in Regulatcry Guide 1.97, Revision 3.

The Seabrook exceptions are provided in FSAR Table 7.5-1, which has been reviewed by the NRC staff in SER Supplement ho. 5.

- SEA 8900K - UNIT 1 6-14 Amendment No. 34

\\

,i*

ADMINISTRATIVE CONTROLS 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:

Cl) receipt of an Operating License, (2) amendment to the license involving a pl anned increase in power level, (3) installation of fuel that has a different design or has been manufactured k/ a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the station.

l l

f I

I SEABROOK - UNIT 1 6-14A Amendment No. 34

J

ADMINISTRATIVE CONTROLS ANNUAL RADI0 ACTIVE EFFLUENT RELEASE IqQgI 6.8.1.4 A matine Annual Radioactive Effluent Release Report covering the operation of the station during the previous calendar year of operation shall be submitted by May 1 of each year.

The Annual Radioactive Effluent Release Reports shall include a summary of the cuantitles of radioactive 11guld and gaseous effluents and solid waste released from the station as outlined in Regulatory Guide 1.21

" Measuring, Evaluating andReportingRadioactivityinSolidvastesandReleasesof Radioactive, Materials in Liquid and taseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, dix B thereof'.

with data summarized on a quarterly basis following the format of Appen For solid wastes, i

the format for Table 3 in Appendix B shall be suppledented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61),

type of container SOLIDIFICATION agen(e.g.&

! SA, Type A, Type B, Lar t or absorbent (e.g., cement).ge Quantity) and The Annual Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual sumary may be either in the fons of an hour-byityl hour listing on magnetic tape of wind speed wind direction atmospheric stabil and precipitation (if measured or In the form of jo, int fr ncy distribu ons of wind speed wind directio)n,khe radiation doses due to the radioac and atmospheric stability.

Thls same report shall include a,n I

assessment of effluents released from the unit or station during the previous calendar year.

This same report shall also include an assessment of the radiation doses from radioactive liould and gaseous effluents to IWWERS 0F IdE PUBLIC due to their activities inside the SITE BOUNDARY durin All assumptions used in making these(Figure 5.1-3)i.e., g the report period.

asssssesnts,-

specific activity, exposure time and location, shall be included in these reports. The o

meteorological conditions concurrent with the time of release of radleactive materials in gaseous effluents, determining the cassous pathway doses.as det shall be used for The measurement,f radiation doses shall be performe3 in accordance with the assessment o asthodology and parameters in the 0FFSITE 00SE CALCULATION MMUAL (00CM).

The Annual Radioactive Effluent Release Report shall also include an assessa:ent of radiation doses to the likely most exposed MEMBER 0F THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources including doses from primary effluent pathways and direct radiation, for the, previous calendar year

)

I 1

In lieu of submission with the Annual Radioactive Effluent Release Report, the licensee has the option of retaining this suemary of required meteorological data on :ite in a file that shall be provided to the NRC upon request.

1 i

I SEABROOK - UNIT 1 6-17 Amendment No. 22

=-

g ADMINISTRATIVE CONTROLS I

ANNyAL RADIQACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 (Contir,jed) to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from 11guld 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 UNRES; JCTED AREAS of radioactive materials in gaseous and 11guld effluents made during the reporting period.

The Annual Radioactive Effluent Release Report shall include any changes 1

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 Ei' fluent Release Report shall also include the fe11owing: an explanation as to why the inoperability of 11guld 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 descriptica of the events leadinti to liquid hoidup tanks or gas storage tanks exceeding the limits of Specif' cation 3.11.1.4.

l S NTHLY OPEATIhG 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, Atta: Document Control Desk, with a copy to the NRC Reillonal Administrator, no later than the 15th of each month following the ca endar month covered by the report.

CORE OPERATING LIMITS REPORT l

6.8.1.6.a Core operating limits shall be established and docum6nted in the CORE l

OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

1.

Cycle dependent overpower AT and Overtemperature AT trip setpoint parameters and function modifiers for operation with skewed axial power profiles for Table 2.2-1 of Specification 2.2.1, 2.

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

SHUTDOWN MARGIN limit for MODE 5 for Specification 3.1.1.2, 4.

Moderator Temperature Coefficient 00L and E0L limits, and 300 ppe surveillance limit for Specification 3.1.1.3, I

SEABROOK - UNIT 1 6-18 Amenduent No. 34

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ML20078H470

Text

Attachment:

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