Text

Amend 9 to License NPF-86,revising Tech Spec by Relocating Cycle Specific Core Operating Limits from Tech Spec to Core Operating Limits Rept
	ML20092M261
Person / Time
Site:	Seabrook ; (NPF-86-A-009, NPF-86-A-9)
Issue date:	02/18/1992
From:	Anthony Mendiola; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20092M262	List: ML20092M261; ML20092M264;
References
	NUDOCS 9202270228
	Download: ML20092M261 (56)
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t' UNITED STATES n

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,i NUCLEAR REGULATORY COMMISSION o,.

WASHINoTON, D C. 20%6

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PUBLIC SEPVICE COMPANY OF NEW HAMPSHIRE. ET AL.*

DOCKET NO. 50-443 SEABROOK STATION. UNIT NO. 1 AMENDMENT-TO FACILITY OPERAT1HG LICENSE Amendment No. 9 License No. NPF-86 1.

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

A.

The application for amendment filed by the Public Service Company of New Hampshire (the licensee), acting for itself and as agent and representative of the 11 other utilitiet listed below and hereafter referred to as licensees, dated September 4, 1991, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and 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)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter 1; D.

The issuance of this amendment will not be inimical to the common defense and security 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 Comission's regulations and all applicable requirements have been 4-satisfied.

Pt.blic Service Company of New Hampshire is authorized to act as agent for the:

Canal-Electric Company, The Connecticut Light and Power Company, EUA Power Corporation, Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Power Company, New Hampshire Electric Cooperative, Inc., Taunton Municipal Light Plant, The

-United Illuminating Company, and Vermont Electric Generation and Transmission Cooperative, Inc., and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

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.- 2.

- Accordingly, the license is amended by changes to the Technicel

-Specifications as-indicated in the attachment to this license

. amendment, and paragraph F.C.(?) of Fccility Operating License No, NPF-06 is hereby amended to rced as follows:

(?)'TechnicalSpecifications The Technical Specifications contained in Appendix A, at revised through Amendment tio. 9, and the Environr.cntal Protection Plan contained in Appendix B are incorporated into facility License No, flPF-86. PSNH shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3, This license amendment is effective as of itt date of issuance and shall be implemented within f0 days of receipt of this letter.

FOR THE NUCLEAR REGULATORY COMMISS10f' l' the J. flendiolo, Acting Director P

t Directorate I-3 Divisior cf Reactor Projects - 1/11 Office of Nuclear Peactor Regulat~ ;n

Attachment:

-Changes to the Technical Specifications Date of Issuance:

February 18, 1992 i.

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ATTACHt4ENT TO LICENSE At4Etir:gNT NO. 9 FACILITY OPERATit;G LICENSE 1:0. NPF-86 00CLET P0. 60 443 Replace the fc110 wing pages of the Aprtndix A Technical Specifications with the attacted pages. The revised pepes are identified by trendment number and contain vertical lines indicating the area of change. Approp.iate over1bp pages have also been provided.

Recove Insert i

i itt iii xiv xiv 1-2 1-2 1-3 1-3 1-4 1 -4 1-5 1-5 1-6 1-6 1-7 1-7 3/4 1-1 3/4 1-1 3/4 1-3 3/41-3 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-8 3/4 1-8 3/4 1-10 3/4 1-10 3/4 1-12 3/4 1-12 3/4 1-15 3/4 1-15 3/4 1-16 3/4 1-16 3/4 1-21 3/4 1-21 3/4 1 22 3/4 1-22 3/4 1-24 3/4 1-23 3/4 ?-1 3/4 P-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/42-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/42-7 3/4 2-7 3/4 2-0 3/4 2-8 E 3/41-1 P 3/4 1-1 E 3/4 1-2 B 3/4 1-?

B 3/4 1-3 B 3/4 1-3 B 3/4 P-1 B 3/4 2-1 B 3/4 ?-3 B 3/4 2-3 6-1B 6-18 6-18A*

6-18B*

6-180*

Denotes new page

INDEX 1.0 -DEFINITIONS SECTION FAHE 1.1 ACTI0N........................................................

1-1 1.2 ACTUATION LOGIC TEST..........................................

1-3 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................

1-1 1.4 AXIAL FLUX DIFFERENCE..........................................

1-1 1.5 CHANNEL CALIBRATION,..........................................

1-1

1. 6 CHANNEL CHECK.................................................

1-1 1.7 CONTAINMENT INTEGRITY.........................................

1-2 1.8 CONTROLLED LEAKAGE............................................

1-2

1. 9 CCRE ALTERATION...............................................

1-2 1.10 CORE OPERATING LIMITS REP 0RT.................................

1-2 1.11 DIGITAL CHANNEL OPERATIONAL TEST..............................

1-2 1.12 00SE EQUIVALENT I-131........................................

1-3 1.13 E - AVERAGE DISINTEGRATION ENERGY............................

1-3 1.14 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................

1-3 1.15 FREQUEMCY N0TATION...........................................

1-3 1.16 GASEOUS.RADWASTE TREATMENT SYSTEM............................

1-3 1.17 IDENTIFIED LEAKAGE...........................................

1-3 1.18 MASTER RELAY TEST............................................

1-4 1.19 MEMBER (S) 0F THE PUBLIC......................................

1-4 1

1.20 0FFSITE DOSE CALCULATION MANUAL..............................

1-4 1.21 OPERABLE - OPERABILITY.......................................

1-4

~ 1. 22 OPERATIONAL MODE - M0DE......................................

1-4 1.23 PHYSICS TESTS................................................

1-4 1.24 PRESSURE BOUNDARY LEAKAGE....................................

1-4 1.25 PROCESS CONTROL PR0 GRAM......................................

1-5 1.26 PURGE - PURGING..............................................

1-5 1.27_ QUADRANT POWER TILT RATI0....................................

1-5 1,28 RATED THERMAL P0WER..........................................

1-5 1,29 REACTOR TRIP SYSTEM RESPONSE TIME............................

1-5 1.30 REPORTABLE EVENT................................

1-5 1.31 CONTAINMENT ENCLOSURE BUILDING INTEGRITY.....................

1-5 1.32 SHUTDOWN MARGIN..............................................

1-6 1.33 SITE B00NDARY................................................

1-6 1.34 S LAV E R E LAY T E S T............................................

1-6 1.35-SOLIDIFICATION...............................................

1-6 1.36 SOURCE CHECK.................................................

1-6 1.37 STAGGERED TEST BASIS.........................................

1-6 1.38 -THERMAL P0WER................................................

1-6 1,39-TRIP ACTUATING DEVICE OPERATIONAL TEST.........

1-6 1.40 UNIDENTIFIED LEAKAGF,..

1-7 1,41 UNRESTRICTED AREA......

1-7 1.42 VENTILATION EXHAUST TRE.0FFl.T SYSTEM.........................

1-7 1.43 VENTING......................................................

1-7 TABLE 1.1 FREQUENCY N0TATION......................................

1-8 TABLE 1.2 OPERATIONAL M0 DES.......................................

1-8 l

l SFABROOK - UNIT 1 i

AM2NDMENT NO. 9 l

l

INDEX

2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 5 SECTION PAGE 2,1 SAFETY LIMITS.................................................

2-1 2.1.1 REACTOR C0RE..................................

21 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

2-1 FIGURE 2.1-1 REACTOR CORE SAFE 1Y LIMIT - FOUR LOOPS IN OPERATION..

2?

22 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...............

2-3 1

TABLE 2.2-1 REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS....

2-4 2.0 BASES t

u.1 SATETY LIMITS 2.1.1 RiiACTOR C0RE................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

B 2-2 i

2.2 L74tTING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS...............

B 2-3 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIkEMENTS 3/4.0---APPLICABILITY...............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg Greater Than 200'F................

3/4 1-1 Shutdown Margin - T,yg Less Than or Equal to 200'F.......

3/4 1-3 Moderator Temperature Coefficient........................

3/4 1-4 Minimum Temperature for Criticality......................

3/4 1-6 SEABROOK - UNIT 1 ii

j INDEX r

LIMITING CONDITIONS,f0R OPERI. TION AND SURVEILL ANCE REQUIREMENTS SECTION

_P.A_G E 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown....................................

3/4 1 7 Flow Paths

Operating...................................

3/4 1 8 Charging Psmp - Shutdown.................................

3/4 1-9 Charging Pseps - Operating...............................

3/4 1 10 Borat ed Wa,er Sources - Shutdown.........................

3/4 1-11 Borated Weter Sources - Operating........................

3/4 1-12 Isolation of Unborated Water Sources - Shutdown..........

3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLlES Group Ho1ght.............................................

3/4 1-15 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00...................

3/4 1-17 Position Indication Systems - Operating..................

3/4 1-18 Position Indication System - Shutdown....................

3/4 1-19 Rod Drop Time...................

3/4 1-20 Shutdown Rod Insertion Limit.............................

3/4 1-21 Control Rod Insertion Limits.............................

3/4 1-22 1

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE....................................

3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z).....................

3/4 2-4 n

3/4,2,3 NUCLEAR ENTHALPY RISE HOT CHANNEt FACTOR,................

3/4 2 8 3/4.2.4 QUADRANT POWER TILT RAT!0................................

3/4 2-9 3/4.2.5 DNB PARAMETERS...........................................

3/4 2 10 3/4.3 INSTRUMEN,TATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................,.

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

3/4 3-2 t

SEABROOK - UNIT 1.

iii Amendment No. 9

o INDEX LIMITING CONDIVIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-2 (This table number is not used)

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................

3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................

3/4 3 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................

3/4 3-16 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUnTION SYSTEM i

-INSTRUMENTATION TRIP SETP0!NTS...........................

3/4 3-24 l

TABLE 3.3-5 (This table number is not used) i TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i

INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-31

3/4.3.3 MONITORING INSTRUMENTATION I

Radiation Monitoring For Plant Operations................

3/4 3 36 i

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS.....................................

3/4 3-37 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.....................

3/4 3-39 l

Movable Incore 0etecte-s................................

3/4 3-40 l

Seismic 1 Instrumentation..................................

3/4 3-41 TABLE'3.3-7 SEISMIC HONITORING INSTRUMENTATION....................

3/4 3-42

. TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............................................

3/4 3 - E Meteorological Instrumentation...........................

3/4 3-44

-TABLE 3.3 8 ' METEOROLOGICAL MONITORING INSTRUMENTATICN.............

3/4 3-45 Remote Shutdown System...................................

3/4 3-46 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM................................

3/4 3-47 Accident Monitoring Instrumentation.....................,

3/4 3-49'

-TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION..................

3/4 3-50 TABLE 3.3 (This table-number is.not used)......................

- 3/4-3-53 Radioactive Liquid Effluent Monitoring Instrumentation...

3/4 3-55 TABLE 3.3-12 RA010 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-56 SEABROOK - UNIT 1 iv k

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m.~.~.--,,.,y_., -.Ah_-...,

_-.,y,,m__,,v,.,-%.r,--,,-~.mw-,,,,,.-,I

INDEX 5.0 DESIGN FEATURES SECTION PAGE

5. 3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES.............................................

5-9

5. 3. 2 CONTROL R0D ASSEMBLIES......................................

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

59 5.4.2 V0LUME......................................................

5-9 5.5 HETEOROLOGICAL TOWER LOCATION.................................

5-9

5. 6 FUEL STORAGE 5.6.1 CRITICALITY.................................................

5-10 5.6.2 DRAINAGE....................................................

5-10 5.6.3 CAPACI'"..................

5-10 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...........................

5-10 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIM 115..................

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

6-1 6.2 ORGANIZATION................................................

6-1 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......................

6-4 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

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

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

6-5 Respons.ollities..........................................

6-5 Records...................................................

6-5 6.2.4 SHIFT TECHNICAL ADVIS0R...................................

6-5 6.3 TRAINING...........................................

6-5 SEABROOK - UNIT 1 xiii

INDEX 4

6.0 ADMINISTRATIVE CONTROLS SECTION PAGE 6.4 REVIEV AND AUDIT............................................

6-6 i

6.4.1 STATION OPERATION REVIEW CLMMITTEE ($0RC)

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

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

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

6-6 Meeting frequency.........................................

6-6 Quorum....................................................

6-6 Responsibilitie',..........................................

6-6 Records...................................................

6-8 6.4.2 NUCLEAR SAFF(Y AUDIT REVIEW COMMITTEE (NSARC)

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

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

6-8 Alternates................................................

6-3 Consultants...............................................

6-8 Meeting Frequency.........................................

6-9 Quorum................................

6*9 Review....................................................

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

6-9 Records...................................................

6-11 6.5 REPORTABLE EVENT ACTI0N.....................................

6-11 6.6 SAFETY LIMIT VIOLATION......................................

6-11 6.7 PROCEDURES AND PR0 GRAMS.....................................

6-12 6.8 REPORTING REQUIREMENTS 6.8.1 ROUTINE REP 0RTS...........................................

6-14 Startup Report............................................

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

6-15 Annual Radiological Environmental Operating Report........

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

6-17 Monthly Operating Reports.......................

6-18 CORE OPERATING LIMITS REP 0RT..............................

6-18 6.8.2 SPECIAL REP 0RTS...........................................

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

6-19 6_.10. RADIATION PROTECTION PR0 GRAM.....................

6-20 SEABROOK - UNIT 1 xiv Amendment No. 9

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1. 0 'Ef!NITIONS

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Yne defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications, gTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required bnder designated conditions.

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output.

The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.

ANALOC CHANNEL OPERAT_10NAL TEST

1. 3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions.

The ANALOG CHANNEL OPERA 110NAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.

AX1AL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFTERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CAllBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK

1. 6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

SEABROOK - UNIT 1 1-1

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DEFINITIONS CONTAINMENT INTEGRITY

1. 7 CONTAINMENT INTEGRITY shall exist when:

a.

All peretrations required to be closed during accident conditions are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve systes., or 2)

Closed by manual valves, blind flanges, 3r deactivated automatic valves secured in their closed positions, b.

All equipment hatches are closed and sealed, c.

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

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

The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE, CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant puinp seals.

CORE ALTERATION 1.9 CORE ALTEPATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT provides core operating limits for the current operating reload cycle.

The cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.8.1.6.

Plant operation within these operating limits is addressed in individual t specifications.

DIGITAL CHANNEL OPERATIONAL TEST 1.11 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exero'ising the digital l

computer hardware using data base manipulation and/or injecting simulated process data to verify OPERABILITY of alarm and/or trip functions.

The Digital Channel Operational Test definition is only applicable to the Radiation Monitoring Equipment.

Amendment No. 9 SEABROOK - UNIT 1 1-2

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l DOSE EQUIVALENT I-131 1.12 DOSE EQUlVALENT l-131 shall be that concentration of 1-131 (microcurie / gram) l which alone would produce the same thyroid dose as the quantity and isotopic mixture of i-131.1-132,1-133,1-134, and 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109, Revision 1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1."

T - AVERAGE DISINTEGRATION ENERGY 1.13 E shall be the average (weighted in proportion to the concentration of l

each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample with half-lives greater than 10 minutes.

ENGINEERED SArETY FEATURES RESPONSE TIME 1.14 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their requir,ed values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance l

Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM 1.16 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and l

installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:

a.

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

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with tl.6 operation of Leakage Detection Systems or not to be PRESSURE BOUNDAn LEAKAGE, or c.

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

SEABROOK - UNIT 1 1-3 Amendment No. /, 9 l

l

DEFINITIONS MASTER RELAY TEST 1.18 A MASTER RELAY TEST shall be the energization of each master relay and l

verification of OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (S) 0F THE PUBLIC 1.19 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-l tionally associated with the plant.

This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

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

OFFSITE DOSE CALCULATION MANUAL 1.20 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall-contain in Part A the l

radiological effluent sampling and analysis program and radiological-environ-mental monitoring program.

Part B of the ODCM shall contain the methodology and parameters used in the calculation of offsite Coses due to radioactive gaseous _and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radio. logical Monitoring Program.

OPERABLE - OPERABILITY A system, subsystem train, component or device shall be OPERABLE or 1.21 haveOPERABILITYwhenitIscapableofperformingitsspecifiedfunction(s),

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

OPERATIONAL MODE - MODE 1.22 An OPERATIONAL MODE (i.e.

MODE) shall correspond to any one incicsive

-l combination of core reactivity condition,-power level, and avere.ge reactor coolant temperature specified in Table 1.2.

i PHYSICS TESTS 1.23;-PHYSICS TESTS shall be those tests performed to measure the fundamental

. nuclear characteristics of the reactor core and related instrumentation:

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

PRESSURE BOUNDARY LEAKAGE-1.24 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube

-leakage)-through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

SEABROOK - UNIT 1 1-4 AmendmentNo./,9 4

OEFINITIONS PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, l

sampling, analyses, tests, and determinations to be made te 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 l

detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lows excore detector calibrated a

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 l

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 l

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 these conditions specified in l

Section 50.73 of 10 CFR Part 50.

CONTAINMENT ENCLOSURE BUILOING INTEGRITY 1.31 CONTAINHENT ENCLOSURE BUILDING INTEGRITY shall exist when:

a.

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

The Containment Enclosure filtration System is OPERABLE, and c.

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

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

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DEFINITIONS j

.l SHU100WN MARGIN i

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

the reactor is subcritical or would be subcritical from its present condition l

assuming all full length rod clasier assemblies (shutdown and control) are fully inserted except for the single rod. cluster assembly of highest reactivity l

worth which is assumed to be fully withdrawn.

i SITE BOUNDARY 1.33 The SITE BOUNDARY shall be that line beyond which the land is neither

[

owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST

-1.34 A SLAVE RELAY TEST shall be the energization of each slave relay ano I

verification of OPERABILITY _of each relay.

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

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

meets shipping and burial ground requirements, j

SOURCE CHECK 1.36 A SOURCE CHECK shall be the qualitative assessement of channel response l

when the channel sensor is exposed to a source of increased radioactivity.

i 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 r

b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval, THERMAL POWER 1,38 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPErJ f!ONAL TEST f

1.39' A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating-Device and_ verifying OPERABILITY of alarm. interlock and/or

. trip functions The. TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment,- as necessary, of the Trip Actuating Device such that it actuates

-at the required Setpoint within'the required accuracy.

_ SEABROOK - UNIT 1 1-6 Amendment No. //, 9 i

^

_. _. _ _-__ _ _ 1. _,_ z_. _ _

l DEFIN!TIONS UNIDENTIFIED LEAKAGE 1.40 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.41 An UNRESTRICTED AREA shall be any area at or beyond the SITE B0UNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATHENT SYSTEM 1,42 A VENTILATION EXHAUST TREATHENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particu-lates from the gaseous exhaust stream prior to the release to the environment.

Such a system is not considered to have any effect on noble gas effluents.

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

VENTING 1,43 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in s

.3 a manner that replacement air or gas is not provided or required during VEN(ING.

Vent, used in system names, does not imply a VENTING process.

AmendmentNo.[9 SEABROOK - UNIT 1 1-7

I TABLE 1.1 i

FRE0VENcy NOTATION NOTATION FREQUENCY f

i 5

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

I D

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

W At least once per 7 days.

H At least once per 31 days.

Q At least once per 92 days.

SA-At least once per 184 days.

j R.

At least once per 18 months.

S/U Prior to each reactor startup.-

N. A.

Not applicable.

P Completed prior to each release.

TABLE 1.2 OPERATIONAL MODES

}

REACTIVITY

% RATED AVERAGE COOLANT MODE _

CON 0! TION. k THERMAL-POWER

TEMPERATURE

-l df 1.

POWER OPERATION _

> 0.99

> 5%

> 350'F l

2.

STARTUP

~

I 0.99

< 5%

I 350'F 3.

HOT STAN08Y I 0.99

~0 I 350'F 5.

COLD SHUTDOWN

< 0.99 0-

' 350'F > T,yg > 200'F

[

4.

HOT SHUTDOWN

< 0.99 0

< 200'F

6. - REFUELING **

$ 0.95 0

5140*F

'* Excluding oecay heat.

- Fuel in the reactor vessel with the vessel head clot.ure bolts less than fully tensioned or with the head removed.

i SEABROOK --UNIT 1 1*8

__..__.__.,_..t__--..____._._.___..._.-._.

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORAT10N CONTROL SHUTDOWN MARGIN - T,yg GREATER THAN 200'f LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHU100WN MARGIN for four-loop operation shall be greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABIL11Y:

MODES 1, 2*, 3, and 4.

ACTION:

With the SHUTDOWN MARGIN less than the limiting value, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHU100WN MARGIN is restored.

SURVEILLANCE REQUIREMENTS i

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

a.

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

If the inoperable control rod is immovable or untrippable, the above required SHU100WN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b.

When in MODE 1 or MODE 2 with k,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; When in MODE 2 with k,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to c.

achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6; d.

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

See Special Test Exceptions Specification 3.10.1.

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

1 i

f REACTIV11Y CONTROL SYSTEMS BORAT10N CONTROL i

SHUTDOWN MARTIN - T,y GREATER 1HAN 200'F SURVEILLANCE REQUIREMENTS i

4.1.1.1.1 (Continued) e.

When in MODE 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor Coolant System boron concentration, i

2)

Control rod position, 3)

Reactor Coolant System average temperature.

4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

4.1.1.1.2 lhe overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% Ak/k at least once per 31 Effective Full Power Days (EFPD)..This comparison shall con 6iger at least those factors stated in Specification 4.1.1.1 le., above, The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel. loading.

1 I

t i

b SEABROOK - UNIT 1 3/4 1-2 i

REACTIVITY CONTROL SYSTIMS BORAT10N CONTROL SHUTDOWN MARGIN - T, LESS THAN Ok EQUAL TO 200'r LIMITING CON 0lTION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit spe-cified in the CORE OPERATING LIMITS REPORT (COLR).

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

APPLICABILITY:

MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than the limit specified in the COLR or the Reac-tor Coolant System boron concentration less than 2000 ppm boron, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN and boron concentration are restored.

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

k a.

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

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

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

1)

Reactor Coolant System boron concentration, 2)

Control rod position, 3)

Reactor Coolant System average temperature, 4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

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

c j

REACTIVITY CONTROL SYSTEMS I

BORATION CONTROL MODERATOR TEMPERATURE COEFFICIENT

- LIMITING CONDITION FOR OPERATION f

i 3.1.1.3-The moderator temperature coefficient (MTC) shall t' within the limits.specified in the CORE OPERATING LIMITS REPORT (COLR).

The maximum t

upper limit shall be less positive than 0 ak/k/'f.

APPLICABILITY:

Beginning of cycle life (BOL) limit - MODES 1 and 2*.only**.

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

ACTION:

a.

With the MTC more positive than the BOL limit specifled in the COLR,-

l operation. in MODES 1 and 2 may proceed provided:

i

- 1.

Control rod withdrawal limits are established and maintained 1

- sufficient to restore the MTC to less positive than the BOL i

limit specified in the COLR, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ..These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies tht-the MTC has been restored to within its limit.for the all rsds

)

withdrawn condition; and

. 3.

A Special Report is prepared and submitted to the Commission, s

pursuant to Specification 6.8.2, within 10 days, describing the-value of-the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restorIngthepositiveMTCtowithinitslimitforthea?lrods withdrawn condition.

b.

_ With the.MTC more negative than the EOL limit specified in the COLR,

-l be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i "With k,ff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

l 1

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

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e.w,w y mo r,

w.. c+,qv--v,-

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--ev3,,-wi,yp.my,v-+,-.,-,p.-y-w-wwegwww.w,-,p.---,

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REACTIVITY CONTROL SYSTEMS BORATION CONTROL HODERATOR TEMPERATURE COEFFICIENT SURVE1LLANCE REQUIREMENTS 4.1.1.3 The HTC shall be determined to be within its limits during each fuel cycle as follows:

a.

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

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.

In the event this comparison indicates the HTC is more negative than the 300 ppm surveillance limit specified in the COLR, the HTC shall be remeasured, and compared to the EOL HTC limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

a SEABROOK - UNIT 1 3/4 1-5 knendment No. 9

?

REACTIVITY CONTROL SYST[MS BORATION CONTROL MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONCITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be greater than or equal to 551'r.

APPLICABillTY:

MODES 1 and 2* **.

t ACTION:

With'a Reactor Coolant System operating loop temperature (T,yg) less than 551'f, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

1 SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 551*F:

a.

Within 15 minutes prior to achieving reactor criticality, and b.

At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 561*F with the T,yg-1 ref I

Deviation Alarm not reset.

4 i

With k,ff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

SEABROOK - UNIT 1 3/4 1-6

-mm-

..v,

RFACTIVifY CONTROL SYSTEMS

.c 3/4.1.2 BORAT10N SYSTEMS FLOW PATHS SHUTDOWN LIM! TING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:

a.

A flow path from the boric acid tanks via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.5a. is OPERABLE, or b.

The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water-storage tank in Specificat, ion 3.1.2.5b. is OPERADLE.

APPLICABILITY:

MODES 4, 5, and 6.

ACTION:

With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source suspend all operations involving CORE ALTERATIONSorpositivereactivItychanges.

SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or other-wise secured in position, is in its correct position.

SEABROOK - UNIT 1 3/4 1-7 i

i REACTIVITY CONTROL SYSTEMS f

BORATION SYSTL15-f FLOW PATHS - OPERATING o

l LIMITING CONDITION FOR OPERATION

{

l 3.1.2.2 At least two of the following three boron injection flow paths shall

[

be OPERABLE:

i a.

lThe flow path from the boric acid tanks via a boric acid transfer i

pump and a charging pump to the Reactor Coolant System (RCS), and l

b.

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

APPLICABILITY:

MODES 1. 2, and 3*

ACTION:

i With only one of the above required boron injection flow paths to the RCS i

OPERABLE, restore at least two boron injection flow paths to the RCS to i

OPERA 8LE' status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAND 8Y and borated to a SHUTDOWN MARGIN equivalent to at least the limit specified in the CORE OPERATING LIMITS REPORT (COLR) for the above MODES at 200'F within the i

next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , j

SURVE!LLANCE REQUIREMENTS a

i 4.1.2.2 At least two of the above required flow paths shall be' demonstrated 1

OPERABLE:

f At least_ once per 31 days by) verifying that each valve (manual, a.

power-operated, or automatic in the flow path that is not locked sealed,orotherwisesecuredinposition-is.initscorrectpositIon; b.

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

- At least once per 18 months by verifying that the flow p6th required by. Specification 3.1.2.2a. delivers at least 30 gpa to the RCS.

j

'*The provisions of Specifications 3.0.4 and 4.0.4'are not applicable for entry into MODE 3 for the centrifugal charging. pump. declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored-to OPERABLE status within.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the tamperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

s SEABROOK - UNIT 1 3/4 1-8 Amendment No. 9

,ew.,.~,..n,.,.n-

. *N..e,-

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,-e,.,,,-,nm-.--w.,--,',_..

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REACTIVIVY CONTROL SYSTEMS BORAT10N SYSTEMS CHARGINC PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABillTY:

H0 DES 4, 5, and 6.

AC110N:

With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SilRVEILL ANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across the pump of greater than or equal to 2480 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All charging pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position vithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering H0DE 4 from HODE 3 or prior to the temperatu e of one or more of the RCS cold legs decreas-ing below 325'f, whichever comis first, and at least once per 31 days there-after, except when the reactor vessel head is removed.

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

SEABROOK - UNIT 1 3/4 1-9

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY:

H00ES 1, 2, and 3.*

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTOOWN MARGIN equivalent to at least the limit specified in the CORE OPERATING LIMITS REPORT (COLR) for the above H00ES at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in 001.0 SHUT 00WN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2480 psid is developed when tested pursuant to Specification 4.0.5.

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

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

SEABROOK - UNIT 1 3/4 1-10 Amendment No. 9 l

REACTIVITY CON 1ROL SYSTEMS 80 RAT 10N SYSTEMS BORATED WATER SOURCES - SHUTDOWN L]MITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water voiume of 6,500 gallons, 2)

A minimum boron concentration of 7000 ppm, and 3)

A minimum solution temperature of 6b'F.

b.

The ref ueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 24,500 gallons, 2)

A minimum boron concentration of 2000 ppm, and 3)

A minimum solution temperature of 50'F.

APPLICABILITY:

MODES 5 and 6.

~

s ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERA 0LE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and 3)

Verifying the boric acid storage tank solution temperature when it is the source of borated water, b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature, SEABROOK - UNIT 1 3/4 1-11

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS 80 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum the following borated water sources shall be OPERABLE asrequiredbySpecifIcation3.1.2.2:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 22,000 gallons, 2)

A minimum boron concentration of 7000 ppm, and 3)

A minimum solution temperature of 65'F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 4/7,000 gallons, 2)

A minimum boron concentration of 2000 ppm, 3)

A minimum solution temperature of 50'F, and 4)

A maximum solution temperature of 9B'F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the system to OPERABLE-status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least the limit specified in the CORE OPERATING LIMITS REPORTS (COLR) for the above MODES at 200'F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLO SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SEABROOK - UNIT 1 3/4 1-12 Amendment No. 9

~

\\

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.

APPLICABill1Y:

MODES 1* and 2*

ACil0N:

a.

With one or more full-length rods inoperable because of being immov-able as a result of excessive f riction or mechanical interf erence or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.3.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more thaq i 12 steps (indicated position), POWER OPERATION may continue provided that within I hour:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

Tne rod is declared inoperable and the remainder of the rods in the group with the incperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.3.5 lhe THERMAL POWER level shall be restricted pursuant to Sped fication 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable an:' the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

A reevaluation of each accident analysit of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation t.nder these conditions; b)

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

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

SEABROOK - UNIT 1 3/4 1-15 Amendment No. 9

REACTIVITY CONTROL SYSfEMS MOVABLE CONTROL ASSEMB'lES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 ACTION b.3 (Continued) c)

A power distribution map is obtained from the movable incore detectors and F (Z) and f are verified to be q

g within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d)

The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within tae next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the Hign Neutron Flux Trip Setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER.

c.

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

1.

Within I hour, the remainder of the rods in the bank (s) with the inoperable rods are aligned to within i 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6.

The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and 2.

The inoperable rods are restored to OPERABLE status within 72

hours, d.

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

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except ouring time intervals when the rod position deviationmonitorjjfinoperable;thenverifythegrouppositionsatleastonce per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

SEABROOK - UNIT 1 3/4 1-16 Amendment No. 9

REACTIVITY CONTROL SYST!MS HOVADLE CONTROL ASSEMBLIES SHUTOOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn # as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

H0 DES la and 2* **.

ACTION:

With a maxirnum of one shutdown rod not fully withdrawn #, except for surveillance testing pursuant to Specification 4.1,3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a.

Fuliv withdraw the rod, or b.

Declare the t'od to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn # as speci-fled in the COLR:

a.

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

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

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

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

The fully withdrawn position is defined as the interval within 225 to the mechanical fully withdrawn position, inclusive.

SEABROOK - UNIT 1 3/4 1-21 AmendmentNo.d,9

REACTIVITY CONTROL SYSTEMS-MOVADLE CONTROL ASSEMBLIES CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 -The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODES 1* and 2* **.

ACTION:

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

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

b.

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

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

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

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

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

SEABROOK - UNIT 1 3/4 1-22 Amendment No._9

^

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,.1 :.

i.

SEABROOK - UNIT 1 3/4 1-23 Amendment No.

,9

= _,

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION F01 OPERATION 3.2.1 The indicated AXIAL FLUX OIFFERENCE (AFD) shall be maintained within the target band (flux difference units)-about the target flux difference as speci-fied in the CORE OPERATING LIMITS REPORT (COLR):

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

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

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

ACTION:

a.

With the indicated AFD outside of the required target band specified in the COLR and with THERMAL PbWER grtster than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:

1.

Restore the indicated AFD to within the target band limits, or 2.

Reduce THERMAL POWER to less than 90% ot' RATED THERMAL POWER.

b.

With the indicated AFD outside of the required target band specified in the COLR for more ti.in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits specified in the-COLR and with THERMAL POWER less than 90% but l

equal to or greater than 50% of RATED THERMAL POWER, reduce:

1.

THERMAL POWER to less than 50% of RATED THERMAL POWER within l

30 minutes, and i

2.

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

See Special Test Exceptions Specification 3.10.2.

- Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits specified in the COLR.

A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ' operation may be accumulated with the nFD outside of the required target band as specified in the COLR during testing without penalty deviation.

SEABROOK - UNIT 1 3/4 2-1 Amendment No. 9

POWER OISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE a

LIMITING CONDITION FOR OPERATION

3.2.1 ACTION

(Continued) c.

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

of RATED THERMAL POWER.

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

a.

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

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

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band.

Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

a.

One-minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.

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

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

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full-Power Days by either determining the target, flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.

Amendment No. 9 SEABROOK - UNIT 1 3/4 2-2

t..

e.

,., + +

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P SEABROOK - UNIT 1 3/4 2-3 Amendment No. 9

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (2) shall be limited by the following relationships:

9 F (Z) $ f K(Z) for P > 0.5 q

P TP F (Z) $ F q

.5 THERMAL POWER

, and Where:

P = f.AIED THERMAL POWER RTP Fq = the Fq limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR), and K(Z) = the normalized F (Z) as a function of core height q

as specified in the COLR.

APPLICABILITY:

MODE 1.

ACTION:

With F (Z) exceeding its limit:

9 a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the q

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

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased, provided F (Z) is demonstrated through incore mapping to be 9

within its limit.

SEABROOK - UNIT 1 3/4 2-4 Amendment No. 9

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~

l SEABROOK - UNIT 1 3/4 2-5 Amendment No. 9

POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-Fg SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F

shall be evaluated to determine if F (Z) is within its limit by:

xy q

a.

Using the movable incore detectors to obtain a power distribution

..iap at any THERMAL POWER greater than 5% of RATED THERMAL POWER, b.

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

Comparing the F computed (Fx) btained in Specification 4.2.2.2b.,

xy above, to:

1)

The F limits for RATED THERMAL POWER (FRTP) for the appropriate xy x

measured core planes given in Specification 4.2.2.2e. and f.,

below, and 2)

The relationship:

RTP F

=F gy pp y_p g x

l Where F is the limit for fractional THERMAL POWER operation P

f express as a function of F

, PF is the Power Factor xy Multiplier for F specified in the COLR and P is the l

xy fraction of RATED THERHAL POWER at which F was measured.

xy d.

Remeasuring F according to the following schedule:

xy RTP 1)

When F is greater than the F limit for the appropriate x

measured core plane but less than the F relationship, additional RTP power distribution maps shall be taken a d F compared to F x

x and F either:

xy a)

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

POWER or greater, the THERMAL POWER at which F*Y was last determined, or b)

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

whichever occurs first.

SEABROOK - UNIT 1 3/4 2-6 Amendment No. 9 K

a 6

,7 POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-Fg SURVEILLANCE REQUIREMENTS 4.2.2.2d. (Continued) l C

RTP 2)

When the F is less than or_ equal to the F limit for the x

x appropriate measured core plane, additional power distribution mapsshallbetakenandFfcomparedtoF andFfatleast RTP x

x x

once per.31 EFPD.:

e.

The F limits-for RATED THERMAL POWER (F P) shall be provided for xy all core planes containing B nk "D" control rods and all unrodded core planes in the CORE OPERATING LIMITS REPORT per Specification 6.8.1.6;-

. f.

The F limits of Specification 4.2.2.2e., above, are not applicable xy in-the following core planes regions as measured in percent of core height from the bottom of the fuel:

1)

. Lower core region from 0 to 15%, inclusive, 2)

-Upper _ core region from 85 to 100%, inclusive, 3)-

Grid plane regions at 17.8't 2%, 32.1 1 2%, 46.4 t 2%, 60.6 1 2%,

and 74.9 t 2%, inclusive, and s

4)_

Core plane regions within t 2% of core height (t 2.88 inches) about the bank demand position of the Bank "D" control rods, C exceedingFf,theeffectsofF g.-

With F n F (Z) shall be evaluated 4

x xy q

to determine if F (Z):is within its limits.

q 4.2.2.3 When F (Z) is measured'for other than F

-determinations, an overall 9

xy measured F (Z) _shall be obtained from a power distribution map and increased q

by_3% to account for manufacturing tolerances and further increased by 5% to-

' account for measurement uncertainty.

SEABROOK - UNIT 1 3/4 2-7 Amendment No. 9 g

gr

,w, v

,-w

POWER DISTRIBUTION LIMITS

-3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION RTP 3.2.3 F shall be less than FAH [1.0 + PF3g (1-P)).

H Where:

P=

THERMAL POWER

, and RATED THERMAL POWER RTP

=theFfH limit at RATED THERMAL POWER (RTP),

F3g specified in the CORE OPERATING LIMITS REPORT (COLR), and AH=thePowerFactorMultiplierforFhspecifiedinthe PF COLR.

APPLICABILITY:

MODE 1.

ACTION:

With Fh exceeding its limit:

a.

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

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

above;THERMALPOWERmaythenbeincreased,providedFhis demonstrated through incore mapping to be within its limit.

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

4.2.3.2 F

shall be demonstrated to be within its limit prior to operation H

above 75% RATED THERMAL PCWER after each fuel loading and at least once per 31 EFPD thereafter by:

a.

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

b.

Using the measured value of F which does not include an allowance g

for measurement uncertainty, i

SEABROOK - UNIT 1 3/4 2-8 Amend..ent No. 9

)

4 3/4.1 REACTIVITY CONTROL' SYSTEMS BASES 3/4.1.1 BORATIONCONTROJ 3/4.1.1.1 and 3/4.1.1.2 -SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that:

(1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

SHUTOOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg.

The most restrictive condition occurs at EOL, with T,yg at no-load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown.

In the analysis of this eccident, a minimum SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) is required to control the reactivity transient.

Accordingly, the SHUT 00WN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions.

With T,yg less than 200' F, the reactivity transients resulting from a postulated steam line break cooldown are minimal.

A SHUTDOWN MARGIN as specified in the COLR and a boron concentration of greater than 2000 ppm are required to permit sufficient time for the operator to terminate an inadvertent boron dilution event with T,yg less than 200' F.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value-of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.

These corrections

.. involved-subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting end of cycle life (EOL) HTC value as specified in the COLR.

The 300 ppm surveillance limit MTC value as specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value as specified in l

the COLR.

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

o REACTIVITY CONTROL SYSTEMS BASES B0 RATION CONTROL 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (Continued)

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551'F.

This limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

NDT 3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation.

The components required tn perform this function include:

(1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

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

The boration capabil-ity of either flow path is sufficient to provide a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT from expected operating conditions after xenon decay and cooldown to 200'F.

The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 22,000 gallons of 7000 ppm borated water from the boric acid storage tanks or a minimum contained volume of 477,000 gallons of 2000 ppm borated water from the refueling water storage tank (RWST).

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

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

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

SEABROOK - UNIT 1 B 3/4 1-2 Amendment No. 9

)

REACTIVITY CONTROL SYSTEMS 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 isolatirn provisions for the Boron 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 MOVA8LE CONTROL ASSIMBLIES The specifications of this section ensure that:

(1) acceptable power.distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, and 220 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.

Since the Digital Rod Position Indication System does not indicate the actual shutdown rod-position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with-demanded position.

The ACTION 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 1

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.

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

\\.

REACTIVITY CONTROL SYSTEMS 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 indivi-dual mechanical fully withdrawn position, T greater than or equal to 551 F 3yg and all reactor coolant pumps operating ensures 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.

Tha 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 i.

D.

Rahe (Westinghouse) to C. O. Thomas (NRC).

This may be by verification of 'h 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.

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

3/4.2 POWER DISTRIBUTION LIMITS i

BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the minimum DNBR in the core greater than or equal to 1,30 during normal operation and in short-term transients, and'(2) limiting the f'ssion gas release, fuel pellet temperature, and cladding mechanical preperties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

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

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat g

flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of g

the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z)

Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFO) assure that the F (Z) upper bound g

envelope of the F limit specified in the CORE OPERATING LIMITS REPORT (COLR) q times the normalized axial peaking factor is not exceeded during either normal operation or in_the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The-full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER redstions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent returr. to SEABROOK - UNIT 1 B 3/4 2-1 Amendment No. 9

POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AXIAL FLUX DIFFERENCE (Continued)

RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited.

Accordingly, a 1-hour penalty deviation limit cumulative duriag the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation ot:tside of the target band but within established limits while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ' actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs en alarm message when the penalty deviation accumulates beyond the limits of 1 hout and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that:

(1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA, the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position; b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; c.

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

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

SEABROOK - UNIT 1 B 3/4 2-2

_ POWER DISTRIBUTION LIMITS BASES-3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

N F

will-be maintained within its limits provided Conditions a. through 3g

d. above are maintained.

The relaxation of F as a function of THERMAL POWER g

allows changes in the radial power shape for all permissible rod insertion limits.

Fuel rod bowing reduces the value of DNBR.

Credit is available to offset this reduction in the generic margin.

The generic margins, totaling 9.1% DNBR completely offset any rod bow penalties.

This margin includes the following:

a.

Design limit DNBR of 1.30 vs. 1.28, b.

Grid spacing (K,) of 0.046 vs._0.059, c.

Thermal diffusion coefficient of 0.038 vs. 0.059, d.

DNBR multiplier of 0.86 vs. 0.88, and e.

Pitch reduction.

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

When an F measurement is taken, an allowance for both experiment'al error g

and manufacturing tolerance must be made.

An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.

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

The F

limit f r RATED THERMAL POWER (FRTP)q as provided in the CORE OPERATING xy x

LIMITS REPORT per Specification 6.8.1.6 was determined from expected power control maneuvers over the full range of burnup conditions in the core.

When RCS F is measured, no additional _ allowances are necessary prior to g

comparisonwiththeestablishedlimitofameasurementerrorof4%forFhhas been allowed for in determination of the design DNBR value.

3/4.2.4 QUADRANT POWER TILT RATIO The purpose of this specification is to detect gross changes in core power distribution between monthly incore flux maps.

During normal operation the QUADRANT POWER TILT RATIO is set equal to zero once acceptability of core peaking factors has been established by review of incore maps.

The limit of 1.02 is established as an indication that the power distribution has changed enouph to warrant further investigation.

SEABROOK - UNIT 1 B 3/4 2-3 Amendment No. 9 t.

POWER DTSTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS Ine limits on the ONB-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 adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3'F for T,yg and 2205 psig for pressurizer are not exceeded.

The measurement error of 2.1% for RCS total flow rate is based upon per-forming a precision heat balance and using the result to 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-servative manner.

Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is applied.

Any fouling hich might bias the RCS flow rate measurement greater than 0.1% can be detec; a 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.

SEABROOK - UNIT 1 B 3/4 2-4

ADMINISTRATIVE CONTROLS S_EMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT

6.8.1.4. Routine Semiannual Radioactive Effluent Release Reports covering the operation of the station during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year.

The period of the first report shall begin with the date of initial criticality.

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

For Jolid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories:

class of solid wastes (as defined by 10 CFR Part 61), type of centainer (e.g., LSA, Type A, Type B, large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days af ter January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year **,

This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured),

or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability,*** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period.

All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports.

The meteorological conditions concurrent with t.he time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses.

The assessment of radiation doses-shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CH).

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

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

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

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

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

SEABROOK - UNIT 1 6-17

ADMINISTRATIVE ___ CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 (Continued) to chow conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating t

the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

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

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCH, pursuant to Specifications 6.12 and 6.13, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.14.

It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.

The Semiannual Radioactive Effluent Release Reports shall also include the following:

an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.9 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 or 3.11.2.6, respectively.

e 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, f

Washington, D.C. 20555, Attn:

Document Control Desk, with a copy to the NRC L

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 REFORT prior to each reload cycle, or prior to any romaining 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.

SHUT 00WN 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. 9 x

f ADMINISTRATIVE CONTROLS 6.8.1.6.a (Continued) 4.

Shutdown Rod Insertion limit for Specification 3.1.3.5, 5.

Control kod Insertion limits for Specification 3.1.3.6, 6.

AXIAL FLUX DIFFERENCE limits and target band for Specification 3.2.1, RTP RTP 7.

Heat Flux Hot Channel Factor, F, K(Z), Fxy, and the Power Factor q

Multiplier for F for Specification 3.2.2, xy RTP 8.

Nuclear Enthalpy Rise Hot Channel Factor, F and the Power Factor 3g MultiplierforFhforSpecification3.2.3.

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

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

1.

WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology" July 1985 (W Proprietary)

Methodology for Specifications:

SHUTDOWN MARGIN limit for MODES 1, 2, 3 and 4 3.1.1.1 SHUTDOWN MARGIN limit for MODE 5 3.1.1.2 Moderator Temperature Coeffici'nt 3.1.1.3 Shutdown Rod Insertion Limit 3.1.3.5 Control Rod Insertion Limits 3.1.3.6 Nuclear Enthalpy-Rise Hot Channel Factor 3.2.3 2.

WCAP-11596-P-A, " Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores" June 1988 (W Proprietary)

Methodology for Specifications:

3.1.1.1 - SHUTDOWN MARGIN limit for MODES 1, 2, 3 and 4 3.1.1.2 - SHUTDOWN MARGIN limit for MODE 5~

3.1.1.3

- Moderator Temperature Coefficient 3.

WCAP-8385-P-A, " Power Distribution Control and Load Following Proce-dures Topical Report," September 1974 (W Proprietary)

Methodology for Specifications:

3.1.3.5 - Shutdown Rod Insertion Limit 3.1.3.6

- Control Rod Insertion Limits 3.2.1

- AXIAL FLUX DIFFERENCE 4.

WCAP-7811. " Power Distribution Control of Westinghouse Pressurized Water Reactors," December 1971 (W Proprietary)

SEABRGJK - UNIT 1 6-18A Amendment No. 9

A,PMINISTRATIVE CONTROLS 6.8.1.6.u.

(Continued)

Methodology for Specificatiens-3.1.3.5

- Shutdown Rod Insertion Limit 3.1. 3. 6 - Control Rod Insertion Limits 5.

Letter, T.M. Anderson to K. Kneil (Chief of Core Performance Branch, NRC), January 31, 1980,

Attachment:

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

3.2.1 AXIAL FLUX DIFFERENCE 6.

NUREG-0800, Standard Review Plan, US Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981, Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev.

2 July 1981 Methodology for Specification:

3.2.1 - AXIAL FLUX DIFFERENCE 7.

WCAP-7308-L, " Evaluation of Nuclear Hot Channel Factor Uncertainties" December 1971 (W Proprietary)

Methodology for Specification:

3.2.2 - Heat Flux Hot Channel Factor 8.

WCAP-8622, " Westinghouse ECCS Evaluation Model, Octcber, 1975 Version," November 1975 (W Proprietary)

Methodology for Specification:

3.2.2 - Heat Flux Hot Channel Factor 9.

WCAP-9220, " Westinghouse ECCS Evaluation Model, February 1978 Version," February 1978 (W Proprietary)

Methodology for Specification:

3.2.2 - Heat Flux Hot Channel Factor

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

Methodology for Specification:

3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor

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

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

I SEABROOK - UNIT 1 6-16B Amendment No. 9 l

p ADMINISTRATIVE CONTROLS 6.8.1.6.b.

(Continued)-

Methodology for Specifications:

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

3.1.3.5 Shutdown Rod Insertion Limit 3.1. 3. 6 Control Rod Insertion Limits

' 3. 2.1 AXIAL FLUX DIFFERENCE-3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor

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

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

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5-6.8.1.6.c.

The core operating limits shall.be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident enalysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT for each reload t.yc.le, including any mid-cycle revisions or supplements thereto,.shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

1 l

l l

1 SEABROOK - UNIT 1 6-18C Amendment No. 9

ML20092M261

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ML20092M261
Site:	Seabrook (NPF-86-A-009, NPF-86-A-9)
Issue date:	02/18/1992
From:	Anthony Mendiola Office of Nuclear Reactor Regulation
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