Text

Proposed Tech Specs,Reflecting Elimination of cycle-specific Parameter Limits to Incorporate Guidance in Generic Ltr 88-16 & SER for Amend 19 to NEDE-24011-P-A
	ML20235Q840
Person / Time
Site:	Brunswick
Issue date:	02/20/1989
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20235Q839	List: ML20235Q835; ML20235Q840; ML20235T388; ML20235T402;
References
	NUDOCS 8903030051
	Download: ML20235Q840 (81)
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, >l BRUNSWICK STEAM ELECTRIC PLANT,-UNITS l'AND.2.: '

M NRC DOCKETS. 50-325 & 50 324L ,

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OPERATING LICENSES DPR-71 & DPR-62 SUPPLEMENT-TO REQUEST.FOR LICENSE AMENDMENT s

ELIMINATION.0F, CYCLE-SPECIFIC PARAMETER LIMITS: ..

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(BSEP-1-104) u INDEX DEFINITIONS SECTION PACE 1.0 DEFINITIONS ACTI0N........................................................... 1-1 AVERACE PLANAR EXP0SURE............................. ............ 1-1 AVERACE PLANAR LINEAR HEAT CENERATION RATE....................... 1-1 CH ANNE L C ALI B RATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 1 CHANNEL CHECK.................................................... 1-1 CHANNEL FUNCTIONAL TEST.......................................... 1 CORE ALTERATION.................................................. 1-2 CORE MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO..... 1-2 CORE OPERATING LIMITS REP 0RT..................................... 1-2 l CRITICAL POWER RATI0............................................. 1-2 DOSE EQUIVALENT I-131............................................ 1-2 E AVERACE DISINTEGRATION ENERCY.................................. 1-2 EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME................ 1-3 FRACTION OF RATED THERMAL P0WER.................................. 1-3 FREQUENCY N0TATION................................................ 1-3 CASEOUS RADWASTE TREATMENT SYSTEM................................ 1-3 IDENTIFIED LEAKAGE............................................... 1-3 ISOLATION SYSTEM RESPONSE TIME................................... 1-3 LIMITING CONTROL ROD PATTERN..................................... 1-3 LOGIC SYSTEM FUNCTIONAL TEST..................................... 1-4 MAXIMUM AVERACE PLANAR HEAT CENERATION RATE RATIO................. 1-4 MEMBER (S) OF THE PUBLIC........................................... 1-4 MINIMUM CRITICAL POWER RATI0..................................... 1-4 ODYN OPTION A.................................................... 1-4 ODYN OPTION B.................................................... 1-4 0FFSITE DOSE CALCULATION MANUAL (0DCM)........................... 1-4 OPERABLE - OPERABILITY........................................... 1-4 .

OPERATIONAL CONDITION............................................ 1 I PHYSICS TESTS..................................................... 1-5 ).

PRESSURE BOUNDARY LEAKAGE........................................ 1-5 PRIMARY CONTAINMENT INTECRITY.................................... 1-5

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DEFINITIONS-

.SECTION, o 4 1.0 DEFINITIONS (Continued), PAGE PROCESS. CONTROL-PROGRAM (PCP) ................................... 1-6.

PURCE - PURCING ..............'................................... 1 .

4 RATE D . TH ER MAL POW ER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -6 REACTOR PROTECTION' SYSTEM RESPONSE TIME ..........................'l-6' REFEREN C E LEVE L Z ERO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 REPORTABLE EVENT ................................................ 1 ROD DENSITY...................................................... 1-6 SECONDARY CONTAINMENT INTEGRITY ................................. 1-7~ lI SHUTDOWN MARCIN ........'..........................................,1-7

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SITE BOUNDARY .................................................... 1-7

' SOLIDIFICATION ................................................... 1-7 SOURCE CHECK ............................................ ........ 1-7

' SPIRAL RELOAD ................................................... 1-7.

SPIRAL-UNLOAD ................................................... 1-8 STACCERED TEST BASIS ............................................ 1-8 THERMA $'P0WER ................................................... 1-8 UNIDENTIFIED LEAKAGE ............................................ 1-8

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UNRESTRICTED AREA ............................................... 1-8 VENTILATION EXHAUST TREATMENT SYSTEM ............................ 1-8 VENTING ......................................................... 1-9 FREQUENCY NOTATION , TABLE 1.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-10 OPERATIONAL CONDITIONS, TABLE 1.2 ............................... 1-11 l BRUNSWICK - UNIT 1 II Amendment No.

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INDEX

w -SAFETY LIMITS AND' LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE

'2.1 SAFETY LIMITS g . Thermal Power (Low Pressure or Low Flow)......................... 2-1 L

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, Thermal Power (High Pressure and High F1ow)...................... 2 Reactor Coolant System Pressure................................... 2-1 Reactor Vessel Water Leve.1........................................ 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection' System Instrumentation-Setpoints...............'2-3 BASES '

2.1 SAFETY LIMITS, Thermal Power-(Low Pressure or Low F1ow).......................... B 2-1 Thermal Power (High Pressure and High F1ow)....................... B 2-2 Reactor Coolant System Pressure................................... B 2-3 Reactor' Vessel Water Level........................................ B 2-3 2.2 Limiting Safety System Settings Reactor Protection System Instrumentation Setpoints............... B 2-4 l 1

BRUNSWICK - UNIT 1 III Amendment No.

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INDEX LIMITING' CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION' PAGE

' 3 / 4 . 0 - AP P LI CAB I L I TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 0- 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 S HUT DOWN MARG I N . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 1 - l '

3/4.1.2 REACTIVITY AN0MALIES....................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control. Rod Operability.................................... 3/4 1-3 Control Rod Maximum Scram Insertion Times.................. 3/4 1-5 Control

Rod Average Scram Insertion Times.................. 3/4 1-6 Four Control Rod Group Insertion Times..................... 3/4 1-7 Control Rod Scram Accumulators............................. 3/4 1-8 Control Rod Drive Coupling................................. 3/4 1-9 Control. Rod Position Indication............................ 3/4 1-11 Control Rod Drive Housing Support.......................... 3/4 1-13 l

3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer........................................ 3/4 1-14 ,

Rod Sequence Cont rol Sys t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1-15 Rod Bl ock Moni t o r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.............................. 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1' AVERAGE PLANAR LINEAR HEAT GENERATION RATE................. 3/4 2-1 3/4.2.2 AP RM S ET P0 I NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 2 - 2 3/4.2.3 MINIMUM CRITICAL POWER RATI0............................... 3/4 2-3 l

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BRUNSWICK - UNIT 1 IV Amendment No.

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INDEX BASES SECTION PACE 3/4.0 AP P LI C AB I LI TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 0- 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARCIN...................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.............'...................

. B 3/4 1-1 3/4.1.3 CONTROL R0DS......................................... B 3/4 1-1 3/4.1.4 CON TROL ROD PROGRAM CONTR0LS . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM........................ B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATING RATE. . . . . . . . . . . B 3/4 2-1 3/4.2.2 APRM SETP0INTS....................................... B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0......................... B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION. . . . . ,. . . . . . . B 3 /4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION........ ......... B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION................................... B 3/4 3-2 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION................................... B 3/4 3-2 3/4.3.5 MONITORING INSTRUMENTATION........................... B 3/4 3-2 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION................................... B 3/4 3-6 3/4.3.7 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM................................. B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES................................. B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 4-1 BRUNSWICK - UNIT 1 X Amendment No.

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4 INDEX ADMINISTRATIVE CONTROLS-a '

SECTION PAGE 6.5.4 ' CORPORATE NUCLEAR SAFETY.SECTION Function......................'......................... 6-13 Organization........................................... '6-13 Review................................................. 6-14 Records................................................- 6-15

' 6. 5. 5' CORPORATE. QUALITY ASSURANCE AUDIT PROGRAM n

Functio'............................................... 6-16 Audits................................................. '6-16 Records................................................ 6-17 Authority...............................'............... 6-17

. Per7onnel.............................................. 6-17 6.5.6 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM............- 6-18 6.6 REPORTABLE EVENT ACTI0N.................................... ~6-18 a6. 7 SAFETY LIMIT VIOLATION..................................... 6-18

"$.8 PROCEDURES AND PR0 CRAMS......................i............. 6-19 6.9 REPORTINC REQUIREMENTS Routine Reports........................................ 6-20 Startup Reports........................................ 6-20 i

Annual Reports......................................... 6-21 j 1

Personnel, Exposure and Monitoring Report............... 6-21 l l

Annual Radiological Environmental Operating Report..... 6-22

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Semiannual Radioactive Effluent Release Report......... 6-23 l Monthly Operating Reports.............................. 6-24 Special Reports........................................ 6-25 Core Operating Limits Report........................... 6-25 i

BRUNSWICK - UNIT 1 XV Amendment No.

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INDEX Lllh ADMINISTRATIVE CONTROLS SECTION ' PACE 6.10- RECORD RETENTION.......................................... 6-26 6.11 RADIATION PROTECTION PR0 CRAM.............................. 6-28 6.12 HICH RADIATION-AREA....................................... 6-28 6.13 0FFSITE DOSE CALCULATION MANUAL (0DCM)..................... 6 6.14 PROCESS CONTROL PROCRAM (PCP).............................. 6-29 6.15- MAJOR. CHANCES TO LIQUID, CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS..........................' 6-30

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BRUNSWICK - UNIT 1 XVI Amendment No.

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1.0 DEFINITIONS The following terms are defined so that uniform inte'rpretation of these specifications may be achieved. The defined terms appear in capitalized type and are applicable throughout these Technical Specifications.

' ACTION ACTIONS are those additional requirements specified as corollary statements to each' specification and shall be part of the specifications.

AVERACE PLANAR EXPOSURE The AVERACE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of. fuel rods in the fuel bundle at that height.

AVERACE PLANAR LINEAR HEAT CENERATION RATE The AVERACE PLANAR LINEAR HEAT CENERATION RATE (APLHCR) shall be applicable to e 9pe:ific planar height and is equal to the sum of the heat generation rate perl unit length cf fuel rod for'all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at that height.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment as necessary of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL ~ TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, I comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

BRUNSWICK - UNIT 1 1-1 Amendment No.

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, DEFINITIONS CHANNEL FUNCTIONAL TEST (Continued) b.. Bistable channels - the injection of a simulated signal into the channel sensor to' verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core with the vessel head removed'and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative location.:

L CORE MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO-The, CORE MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO (CMAPRAT)

.shall be the largest value of the MAPRAT that exists.in the core.

CORE OPERATING LIMITS REPORT The CORE OPERATINC' LIMITS REPORT is the unit-specific document that provides core operating limits for the current reload _ cycle. These cycle-specific core operating limits.shall'be determined for each reload cycle in accordance with Specifications 6.9.3.1, 6.9.3.2, 6.9.3.3, and 6.9.3.4. Plant' operation within these' core operating. limits is addressed in individual specifications.

CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated, by application of an NRC approved correlation, to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be concentration of I-131, uci/ gram, which.alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.. The following is defined equivalent to 1 uCi of I-131 as determined from Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites": I-132, 28 uci; I-133,'3.7 uCi; I-134, 59 uCi; I-135, 12 uCi.

E -AVERACE DISINTEGRATION ENERGY E shall be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes with half lives greater than 15 minutes remking up at least 95% of the total non-iodine activity in the coolant. '

BRUNSWICK - UNIT 1 1-2 Amendment No.

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DEFINITIONS EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME -

The EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its l safety function (i.e., the valves cravel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FRACTION OF RATED THERMAL POWER The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

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

CASEOUS RADWASTE TREATMENT SYSTEM <

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A CASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to ;

reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the ,

purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or

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

ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable.

LIMITING CONTROL ROD PATTERN A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a limiting value for APLHCR or MCPR. l BRUNSWICK - UNIT 1 1-3 Amendment No.

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DEFINITIONS LOCIC SYSTEM FUNCTIONAL TEST ;

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A LOCIC SYSTEM FUNCTIONAL TEST shall be a test of all relays and contacts of a {

logic circuit, from sensor output *o activated device, to ensure that I components are OPERABLE.

MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO The MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO (MAPRAT) for a bundle shall be the maximum ratio of the APLHCR at a specific height in the bundle divided by the exposure dependent APLHCR limit for that specific l height.

MEMBER (S) OF THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the. plant. This category does not include employees of the utility, 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 recreational, 3 occupational or other purposes not anisociated with the plant. '

MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

ODYN OPTION A ODYN OPTION A shall be analyses which refer to minimum critical power ratio limits which are determined using a transient analysis plus an analysis uncertainty penalty.

ODYN OPTION B ODYN OPTION B shall be analyses which refer to minimum critical power ratio limits determined using a transient analysis which includes a requirement for 20% scram insertion times to reduce the analysis uncertainty penalty.

OFFSITE DOSE CALCULATION MANUAL (ODCM) l The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a manual which contains the current methodology and parameters to be used to calculate offsite doses ;

resulting from the release of radioactive gaseous and liquid effluents; the 1 methodology to calculate gaseous and liquid effluent monitoring j instrumentation alarm / trip setpoints; and, the requirements of the environmental radiological monitoring program.

i OPERABLE - OPERABILITY

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

BRUNSWICK - UNIT 1 1-4 Amendment No.

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DEFINITIONS OPERABLE - OPERABILITY'(Continued) l- Implicit in this definition shall be the assumption that all necessary attendant instrumentation, contr'ols, normal and emergency electric power sources,-cooling or seal water, lubrication or other auxiliary-equipment that l 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 CONDITION An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor cora and related instrumentation and are 1) described in Section 14 of the Updated FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. l PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolatable fault in a reactor coolant system component body, pipe wall, or vessel wall.

PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY shall exist when:

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

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

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.1, or

b. All equipment hatches are closed and sealed.

c. Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

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

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

BRUNSWICK - UNIT 1 1-5 Amendment No.

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' DEFINITIONS l PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formula, sampling, j analyses, tests and determinations to be made to ensure that the processing and packaging of' solid radioactive wastes based on demonstrated processing of i actual'or simulated wet' solid wastes will be accomplished in such a way as to l assure compliance with 10.CFR Part 20, 10 CFR Part 71, and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

PURCE - PURCING PURCE or PURCINC is the controlled process of discharging air or gas from'a confinement to maintain' temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is.

required to purify the containment.

RATED THERMAL POWER RATED THERMAL POWER'shall be total reactor core heat transfer rate to the reactor coolant of 2436 MWt.

REACTOR PROTECTION SYSTEM RESPONSE TIME i

REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the vessel zero point. This REFERENCE LEVEL ZERO is approximately mid point on the top fuel guide and is the sing'.e reference for all specifications of vessel water level.

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

ROD DENSITY ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

BRUNSWICK - UNIT 1 1-6 Amendment No.

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I DEFINITIONS k

SECONDARY CONTAINMENT INTECRITY '

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SECONDARY CONTAINMENT INTEGRITY shall exist when;

a. . All automatic reactor building ventilation system isolation valves or dampers"are OPERABLE or secured in the isolated position.

Lb' The standby gas treatment system is OPERABLE pursuant to Specification 3.6.6.1.

c. At least or2e door in each access to the reactor building is closed.

d. The sealing mechanism associated with each penetration (e.g., welds, bellows, er 0-rings) is OPEEABLE.

SHUTDOWN MARCIN SHUTDOWN MARCIN shall be'the amount of reactivity by.which the reactor would be suberitical assuming that all control rods capable of insertion are fully inserted except for the analytically determined highest' worth rod which is

. assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68'F, and Xenon free.

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

Figure 5.1.3-1.

SOLIDIFICA_ TION SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of chennel response when the channel sensor is exposed to radiation.

1 SPIRAL RELOAD A SPIKAL EELOAD is the reverse of a SPIRAL UNLOAD. Except for fuel bundles

, around each of the four SRMs,, the fuel in the interior of~the core, symmetric to the SRMs, is loaded first. Up to four fuel bundles may be loaded around each of the four SRMs.

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i BRUNSWICK - UNIT 1 1-7 Amendment No.

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DEFINITIONS SPIRAL UNLOAD' LA ' SPIRAL UNLO'AD is a core unload performed by first' removing the fuel from the outermost control cells (four bundles surrounding a control blade). Unioading

= continues in a spiral fashion by removing fuel from the outermost periphery to the interior of the core, symmetric about the SRMs, except for fuel-bundles around each of the four SRMs. Up to four fuel bundles may be left around each SRM to maintain adequate count rate. i STACCERED TEST BASIS A STACCERED TEST BASIS shall consist of:

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

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

component at the beginning of each subinterval. '

THERMAL POWER THERMAL POWER shall be the total reactor core h'est transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKACE shall be all leakage which is not IDENTIFIED LEAKAGE.

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

VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is 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 particulate from the gaseous exhaust stream prior to the release to the j environment. Such a system is not considered to have any effect on noble gas I

effluents.. Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

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BRUNSWICK - UNIT 1 1-8 Amendment No.

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VENTINC.is'the' controlled' process of. discharging' air.or gas.from a confinement to maintain . temperature, pressure, humidity, concentration or other operating -

-condition,-in such;a manner that replacement air or gas is not provided.or

" required.: Vent, used in system names',' does not imply a VENTINC process.

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1 BRUNSWICK - UNIT 1 1-9 Amendment No.

.. _a _ ;, .

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'(BSEP-1-104)-

l TABLE 1.1 '

FREQUENCY NOTATION-w

, , 3 fi NOTATION' FREQUENCY?

A S At least cnce per:12 hours.

D 'At' lea'st once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .-

LW At lbast'once per 7 days.-

SM At least'once per'16 days.

.H At.least'once per 31 days.

Q "At least;once per.92' days.

-SA At least once per 184 days.

s ,

A; At least once per 366 days.

R At least once.per 18 months (550 days).:

S/U Prior to each reactor startup.

P Prior to each' release.

'NA '

Not applicable..

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< 1 BRUNSWICK - UNIT 1 1-10 Amendment No.

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

(BSEP-1-104)

TABLE 1.2 OPERATIONAL CONDITIONS AVERACE OPERATIONAL MODE SWITCH . REACTOR COOLANT CONDITIONS POSITIONS TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown #'*** > 212*F

4. COLD SHUTDOWN Shutdown #'##'*** $ 212*F

5. REFUELINC* Shutdown or Refuel **'I 1 212*F

The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

- See Special Test Exceptions 3.10.1 and 3.10.3.

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled provided that the one-rod-out interlock is OPERABLE.

BRUNSWICK - UNIT 1 1-11 Amendment No.

(BSEP-1-104) 2.1 SAFETY LIMITS BASES 2 .'O Th'e fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity limit.is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that -1 the MINIMUM CRITICAL POWER RATIO (MCPR) is no less than the Safety Limit CPR l of' Specification 2.1.2. The Safety Limit MCPR represents a conservative !

margin relative to the conditions required to maintain fuel cladding i integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER (Low Pressure or Low Flow)

The use of the NRC approved CPR correlation is not valid for all critical power calculations at pressures below 800 psia or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power andflowsyillalwaysbegreaterthan4.5 psi. Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. 3Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 800 psia is conservative.

BRUNSWICK - UNIT 1 B 2-1 Amendment No. !

9 .: :

-(BSEP-1-104) ]

l SAFETY LIMITS l

BASES l

2.1.2 THERMAL POWER (High Pressure and High Flow)

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reacter -3 operation, the thermal and hydraulic conditions resulting in a departure from {

nucleate boiling have been used to mark the beginning of the-region where fuel j damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical i

. power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating i

state and in the procedures used to calculate the critical power, result in an uncertainty in.the value of the critical power. Therefore, the fuel cladding 3

J integrity safety' limit is defined as the critical power ratio in the limiting !

fuel assembly for which more than 99.9% of the fuel rods in the core are i expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

l The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using an approved critical power .i' correlation. Details of the fuel cladding integrity safety limit calculation are provided in Reference 1.

Uncertainties used in the determination of the fuel cladding integrity safety limit and t.he bases of these uncertainties are presented in Reference 1.

The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution

'having the greatest number of assemblies at the highest power levels. The worst distribution in Brunswick Unit I during any fuel cycle would not be as severe as the distribution used in the analysis. The pressure safety limits are arbitrarily selected to be the lowest transient overpressure allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1. l Reference i

1. NEDE-240ll-P-A, " General Electric Standard Application for Reactor J Fuel," latest approved revision. l l

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BRUNSWICK - UNIT 1 B 2-2 Amendment No.

4J .,

3: (BSEP-1-104) n, o SAFETY LIMITS ng: ,

BASES (Continued) 2.1.3 REACTOR COOLANT SYSTEM PRESSURE

.The Safety Limit for the reactor coolant system pressure has been selected such that it-is at a pressure below which it can be shown that the integrity of the system is not endangered. However, the pressure safety. limit-is set high enough such that'no. foreseeable circumstances'can cause the system pressure to rise'to this~ limit. The pressure safety limit is also selected to be the lowest transient overpressure allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III and USAS Piping Code, Section B 11.1.

2.1.4 REACTOR VESSEL WATER LEVEL With fuel _in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of' decay heat. If the water level should drop below the top of the active fuel during this period, the ability to remove decay heat is reduced.

This reduction:in cooling capability could lead to elevated cladding temperatures and clad perforation in-the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.

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i BRUNSWICK - UNIT 1 B 2-3 Amendment No.

_-____--__--_a

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.(BSEP-1-104)

'2.2 LIMITING SAFETY SYSTEM SETTINGS

-BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS i

f The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

1. . Intermediate Range Monitor, Neutron Flux - High The'IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5-decade 10-tange instrument. The trip setpoint of 120

. divisions is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setroint is also' ranged up. Range 10 allows the IRM instruments to remain on scale at higher power levels to provide for additional overlap and also permits calibration at these higher powers.

The most significant source of reactivity change during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required. protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR. The most severe. case involves an initial condition in which the reactor is just subcritical and the IRMs are not yet on scale. Additional conservatism was taken in this analysis-by assuming the IRM chaenel closest to the rod being withdrawn is bypassed. The results of this analysis show that the reactor is shut down and peak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above the Safety Limit MCPR of Specification 2.1.2. Based on this analysis, the IRM provides protection.against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between ,

the setpoint and the Safety Limits. This margin accommodates the anticipated ]

maneuvers associated with power plant startup. Effects of increasing pressure l at zero or low void content are minor, cold water from sources available j during startup is not much colder than that already in the system, temperature I coefficients are small, and control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform control rod 1

i BRUNSWICK - UNIT 1 B 2-4 Amendment No. l u

, j t

(BSEP-1-104) 2.2 LIMITING SAFETY-SYSTEM SETTINGS BASES (Continued)

2. Average Power Range Monitor (Continued)

I withdrawal is the most probable cause of significant power increase. Because -f the flux distribution' associated with uniform rod withdrawals does not involve i high local peaks and because several rods must be moved to change power by a ,

significant amount, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod 3 withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% APRM trip remains active until the mode switch is placed in the Run .i position.

The APRM flow-biased trip system is calibrated using heat balance data I taken during steady state conditions. Fission chambers provide the basic input to the system and, therefore, the monitors respond directly and quickly to changes due to transient operation; i.e., the thermal power of the fuel will be less than that indicated by the neutron flux due to the. time constants of the heat transfer. Analyses demonstrate that with only the 120% trip setting, none of the abnormal' operational transients analyzed violates the fuel safety limit and there is substantial margin from fuel damage.

Therefore, the use of the flow-referenced trip setpoint, with the 120% fixed setpoint as backup, provides adequate margins of safety.

The APRM trip setpoint was selected to provide adequate margin for Safety Limits and yet allows operating margin that reduces the-possibility of unnecessary shutdowns. The flow-referenced trip setpoint must be adjusted by the specified formula in Specification.3.2.2 in order to maintain these margins. l

3. Reactor Vessel Steam Dome Pressure-High High Pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating, will also tend to increase the power of the reactor by compressing voids, thus adding reactivity. The trip will quickly reduce the neutron flux counteracting the pressure increase by decreasing heat generation. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the 1 BRUNSWICK - UNIT 1 B 2-5 Amendment No. f

l '.

(BSEP-1-104) 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

-i

3. Reactor Vessel Steam Dome Pressure-High (Continued) pressure measurement compared to the highest pressure that occurs in the system during a transient. This setpoint is effective at low power / flow conditions when~the turbine stop valve closure is bypassed.. For a turbine trip under these conditions, the transient analysis indicates a considerable margin to the thermal hydraulic limit.

4. Reactor Vessel Water Level-Low, Level #1 The reactor war.er level trip point was chosen far enough below the normal operating level to avoid spurious scrams but high enough above the fuel to assure that there is adequate water to account for evaporation losses and displacement of cooling following the most severe transients. This setting was also used to develop the thermal-hydraulic limits of power versus flow.

5. Main Steam Line Isolation Valve-Closure The low pressure isolation of the main steamline trip was provided to give protection against rapid depressurization and resulting cooldown of the reactor vessel. Advantage was taken of the shutdown feature in the run mode which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low pressures does not ..

occur. Thus, the combination of the low pressure isolation and isolation i valve closure reactor trip with the mode switch in the Run position assures the availability of neutron flux protection over the entire range of the Safety Limits. In addition, the isolation valve closure trip with the mede switch in the Run position anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.

6. Main Steam Line Radiation - High The Main Steam Line Radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected, a scram is initiated to reduce the continued failure of fuel cladding. At the same time, the Main Steam Line Isolation Valves are closed to limit the release of fission products. The trip setting is high enough above background radiation level to prevent spurious scrams, yet low enough to promptly detect gross failures in the fuel cladding.

BRUNSWICK - UNIT 1 B 2-6 Amendment No.

. . 1 (BSEP-1-104) k LIMITING SAFETY SYSTEM SETTINCS 1 BASES (Continued) "

1

7. Drywell Pressure, High '

High pressure in the drywell could indicate a break in the nuclear ;

process systems. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant.

The trip setting was selected as low as possible without causing spurious trips.

8. Scram Discharge Volume Water Level-High The scram discharge tank receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this tank fill up to a point where there is insufficient volume to accept the displaced water, control rod movement would be hindered. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods when they are tripped.

9. Turbine Stop Valve-Closure I

The turbine stop valve closure trip anticipates the pressure, neutron i flux, and heat flux increases that would result from closure of the stop valves. With a trip setting of 10% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.

10. Turbine Control Valve Fast Closure, Control Oil Pressure - Low ,

1 The reactor protection initiates a scram signal after the control valve l hydraulic oil pressure decreases due to a load rejection exceeding the '

capacity of the bypass valves or due to hydraulic oil system rupture. The i turbine hydraulic control system operates using high pressure oil. There are several points in this oil system where upon a loss of oil pressure, control valves closure time is approximately twice as long as that for the stop valves, which means that resulting transients, while similar, are less severe q than for stop valve closure. No fuel damage occurs, and reactor system l pressure does not exceed the safety relief valve setpoint. This is an !

anticipatory scram and results in reactor shutdown before any significant '

increase in pressure or neutron flux occurs. This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first-stage pressure.

BRUNSWICK - UNIT 1 B 2-7 Amendment No.

- , e (BSEP-1-104)

REACTIVITY CONTROL SYSTEMS

' ROD BLOCK MONITOR LIMITINC CONDITION FOR OPERATION 3.1.4.3 Both Rod Block Monitor (RBM) channels shall be OPERABLE.'

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.

ACTION:

a. With one RBM channel inoperable, POWER OPERATION may continue provided that either:

1. 'The inoperable RBM channel is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. The redundant RBM is demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at' least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable RBM is restored to OPERABLE status, and the inoperable RBM is restored to OPERABLE status within 7 days, or

3. THERMAL POWER is limited such that MCPR will remain above the Safety Limit MCPR'of Specification 2.1.2, assuming a single error that results in complete withdrawal of any single control rod that is capable of withdrawal.

Otherwise, trip at least one rod block monitor channel.

b. With both RBM channels inoperable, trip at least one rod block monitor channel within one hour.

SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and during the OPERATIONAL CONDITIONS'specified in Table 4.3.4-1.

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BRUNSWICK - UNIT 1 3/4 1-17 Amendment No. j o

(BSEP-1-104) ]

l 1

.3/4.2. POWER DISTRIBUTION LIMITS i

3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE j LIMITING CONDITION FOR OPERATION 3.2.1 During power ' operation, the AVERACE PLANAR LINEAR HEAT CENERATION RATE (APLHCR) for each. type of fuel as a function of axial location and AVERACE PLANAR EXPOSURE shall not exceed limits based on applicable APLHCR limit values that have been approved for the respective. fuel and lattice type and determined by the approved methodology described in CESTAR-II. When hand calculations are required, the APLHCR for each type of fuel as a function of AVERACE PLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) of each type of fuel shown in the applicable figures of the CORE OPERATING LIMITS REPORT.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than-or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHCR-exceeding the limits specified in.the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and continue corrective action so that APLHCR is within the required limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

s SURVEILLANCE REQUIREMENTS .__

4.2.1 All APLHCRs shall be verified to be equal to or less than the limits specified in Specification 3.2.1:

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

b. Within 12. hours after completion of a THERMAL' POWER increase of at least 15% of RATED THERMAL-POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHCR.

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J BRUNSWICK - UNIT 1 3/4 2-1 Amendment No.

j 1

(BSEP-1-104) i

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l POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS l i

LIMITING CONDITION FOR OPERATION 3.2.2 The flow-biased APRM scram trip setpoint (S) and rod block trip set point (SRB) shall be established according to the following relationship:

S $ (0.66W + 54%) T SRB $ (0.66W + 42%) T where: S and S RB are in percent of RATED THERMAL POWER.

W = Loop recirculation flow in percent of rated flow.

T = FRACTION OF RATED THERMAL POWER (FRTP) divided by CORE MAXIMUM AVERACE LINEAR HEAT CENERATION RATE RATIO (CMAPRAT)

(T is applied only if less than 1.0.)

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ;

ACTION:

With S or SRB exceeding the allowable value, initiate corrective action within 15 minutes required and limits continue within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months scorrective or reduceaction so thatPOWER THERMAL S and SRB are to less within than 25%the of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and CHAPRAT shall be determined, the value of T calculated, and the flow-biased APRM scram trip and rod block trip setpoints verified to be within the above limits or adjusted, as required:

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with CHAPRAT greater than or equal to FRTP.

BRUNSWICK - UNIT 1 3/4 2-2 Amendment No. l

i I

'(BSEP-1-104)

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO-LIMITING CONDITION FOR OPERATION 1

3.2.3.1 1..2 MINIMUM CRITICAL POWER RATIO (MCPR),'as a function of core flow and cycle average er.posure, shall be equal to or greater than the MCPR limit times the Kg specified in the CORE OPERATING LIMITS REPORT. The MCPR limits for ODYN OPTION A and ODYN OPTION B analyses, used in the above determination, shall be specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER ACTION:

With MCPR, as a function of core flow and cycle average exposure, less than the applicable MCPR limit specified in the. CORE OPERATING. LIMITS REPORT, inir.iate corrective action within 15 minutes and .estore MCPR to within the applicable limit within 4-hours or reduce' THERMAL POWER'to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.3.1 MCPR, as a function of core flow and cycle average exposure, shall be determined to.be equal to or greater than the applicable MCPR limit of Specification 3.2.3.1:

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating in a LIMITING CONTROL ROD PATTERN for MCPR.

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

(BSEP-1-104)

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)

LIMITING CONDITION FOR OPERATION 3.2.3.2 For the OPTION B MCPR limits provided in the CORE OPERATING LIMITS REPORT to be used, the cycle average 20% (Notch 36) scram time (Tay,) shall be less than or equal to the OPTION B scram time limit (TB), where t,y, and T B are determined as follows:

n

=

[ N; t; i=1 T,y, n N. ' " *#8

[

i=1 i = Surveillance test number, n = Number of surveillance tests performed to date in the cycle (including BOC),

N; = Number of rods tested in the i th surveillance test, and ti = Average scram time to notch 36 for surveillance test i l N 1/2 i T

B " "

I*b' n N. ' " *#*'

i i[=1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including BOC),

th Ni = Number of rods tested in the i surveillance test-N; = Number of rods tested at BOC, u = 0.813 seconds (mean value for statistical scram time distribution from ,

de-energization of scram pilot valve solenoid to pickup on notch 36),

o = 0.018 seconds (standard deviation of the above statistical distribution).

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER.

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

j' j ' ,l-~

(BSEP-1-104)

POWER DISTRIBUTION LIMITS LIMITINC CONDITION FOR OPERATION (Continued)

, ACTION Within twelve hours after determining that t,y, is greater than tB, the ,

operating limit MCPRs shall be either:

a. Adjusted.for each fuel type such that the operating lim.*t ACPR is the maximum of the non pressurization transient MCPR operating limit specified in the CORE OPERATING LIMITS lEPORT or the adjusted pressurization transient MCPR operatir.g limits, where the' adjustment is made by:

+ ""* ~

MCPR = MCPR (MCPR - MCPR option ad j.usted option B T A

~

B Ption A B).

where: T A = 1.05 seconds, control rod average scram insertion time limit to notch 36 per Specification 3.1.3.3, MCPRoption A = Specified in the CORE OPERATING LIMITS REPORT, MCPRoption B " Specified in the CORE OPERATING LIMITS REPORT, or

b. The OPTION A MCPR limits specified in the CORE OPERATING LIMITS REPORT.

SURVEILLANCE REQUIREMENTS 4.2.3.2 The values of T and i shall be determined and compared each time a scram test isperforme8I*Ther$quirementforthefrequencyofscramtime testing shall be identical to Specification 4.1.3.2.

I BRUNSWICK - UNIT 1 3/4 2-5 Amendment No.

{ $:

(BSEP-1-104) !

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~l INSTRUMENTATION f 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4 The control rod withdrawal block instrumentation shown in Table 3.3.4-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4-2.

APPLICABILITY: As shown in Table 3.3.4-1.

ACTION:

a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4-2, declare the channel inoperable until the channel is restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

b. With the requirements for the minimum number of OPERABLE channels not satisfied for one trip system, POWER OPERATION may continue provided that either:

1. The inoperable channel (s) is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. The redundant trip system is demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable channel is restored to OPERABLE status, and the inoperable channel is restored to OPERABLE status within 7 days, or

3. For the Rod Block Monitor only, THERMAL POWER is limited such that MCPR will remain above the Safety Limit MCPR of ,1 Specification 2.1.2, assuming a single error that results in .l complete withdrawal of any single control rod that is capable of ,

withdrawal.

4. Otherwise, place at least one trip system in the tripped condition tithin the next hour.

c. With the requirements for the minimum number of OPERABLE channels not satisfied for both trip systems, place at least one trip system in the tripped condition within one hour.

d. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5.

SURVEILLANCE REQUIREMENTS 4.3.4 Each of the above requir'ed control rod withdrawal block instrumentation i channels shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK, CHANNEL CALIBRATION, and a CHANNEL FUNCTIONAL TEST during the OPERATIONAL ,

CONDITIONS and at the frequencies shown in Table 4.3.4-1. l I

BRUNSWICK - UNIT 1 3/4 3-39 Amendment No.

s *

(BSEP-1-104) i REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) potential effects of the rod ejection accident are limited. The ACTION statements permit variations from the basic' requirements but at the same time.

impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem; therefore, with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the non-fully-inserted position are consistent with the SHUTDOWN MARCIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shut down for investigation and resolution of the problem.

The control rod system is analyzed to bring the reactor suberitical at a rate fast enough to prevent the MPCR from 'vecoming less than the Safety Limit j HCPR of Specification 2.1.2 during the limiting power transient analyzed in 1 Section 14.3 of the FSAR. This analysis shows that the negative reactivity rates resul' ting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than the Safety Limit MCPR of Specification 2.1.2. The occurrence of scram times longer than those specified should be viewed as an indication of a systemic problem with the rod drives and, therefore, the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion i

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

V >

(BSEP-1-104) j 3/4.2 POWER DISTRIBUTION LIMITS BASES -

The specifications of this section assure that the-peak cladding temperature

.followingthepostulateddesignbasisloss-of-coolantaccidentwillnotexceed the 2200 F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effects-of fuel pellet densification. i 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE

.The limiting values for APLHCR when conformance to the operating limit is

, performed by hand calculation are provided in the.CORS.0PERATING LIMITS REPORT for each fuel type and, when required for the most limiting lattice for multiple lattice fuel bundle types.

This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed' the limits specified in 10 CFR 50.46 and that the fuel design analysis limits specified in NEDE-24011-P-A-(Reference 1) will not be exceeded.

Mechanical Design Analysis: NRC approved methods (specified in Reference 1) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 1. No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the fuel design analysis APLHCR limit.

LOCA Analysis: A LOCA analysis is performed in accordance with 10 CFR 50 Appendix K to demonstrate that the permissible planar power ( APLHCR) limits comply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination ~ for the plant.

The Technical Specification APLHCR limit is the most limiting composite of the fuel mechanical design analysis APLHCR and the ECCS APLHCR limit.

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

)

.______-_____-_-___--__-O

r g . ~,2

.i

1 (BSEP-1-104)

g. jf POWER DISTRIBUTION LIMITS.

BASES 3/4.2.2' APRM SETPOINTS The scram setting'and rod block functions.of the.APRM instruments are adjusted.

to' ensure.that: fuel design and safety, limits.areLnot exceeded. The scram settings and. rod block' settings are adjusted in accordance with the

- relationship provided.in Specification 3.2.2. This adjustment may be accomplished by increasing the APRM gain and thus reducing,the' slope and intercept point _of the flow referenced APRM high flux scram curve by the reciprocal of the APRM gain-change.

3/4.2.3 MINIMUM CRITICAL' POWER RATIO The required operating limit MCPR's at steady state operating conditions as specified in Specification 3.2.3 are derived from'an established fuel' cladding integrity Safety Limit.MCPR approved by the NRC and an analysis of abnormal operational Ltransients. For any abnormal' operating transient analysis evaluation with the initial condition of the reactor being at the st'eadyl state-operating limit, it is required that the resulting MCPR does not decrease i below the: Safety Limit MCPR at any time during the transient, assuming l instrument trip setting as given in Specification 2.2.1.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting-transients have been analyzed to. determine which result in the largest reduction in CRITICAL POWE.R RATIO (CPR).

Details on how evaluations are performed, on the methods used, and how the MCPR limit is adjusted for operation at less than rated power and flow conditions are given in Reference 1 and the CORE OPERATING LIMITS REPORT.

At core THERMAL POWER levels less than or equal to 25% RATED THERMAL POWER, the reactor will be operating at a minimum recirculation pump spe'ed and the moderator void content will be very small. For all designatedbcontrol rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, an MCPR evaluation will be made at 25% THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR q evaluation below this power level will be shown to be unnecessary. The daily '

requirement'for calculating MCPR above 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for ,

calculating MCPR when a limiting control rod pattern is approached ensures '

that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.

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

w----______ - --

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ie , ; s . 4 '

i -

wll I^. . '

%- %. ;s,..![ , '(BSEP-A-104).

i. '

L POWER' DISTRIBUTION LIMITS li; ' n- BASES

,;i f;;

i r 1

- Referenc.,,es '

t,"c -

z 1.. NEDE-24011-P-A,;" General Electric Standard Application:for.. Reactor

'n Fuel," latest approved version.

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I BRUNSWICK - UNIT 1 8 3/4 2-3 Amendment No.

L. _ . _ . _ _.____________________o

(BSEP-1-104) t

-ADMINISTRATIVE CONTROLS' SPECIAL REPORTS 6.9.2 :Special reports shall be. submitted to the Regional Administrator of the

Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.. '

Ja. Inoperable' Seismic Monitoring Instrumentation, Specification'3.3.5.1.

b.- . Seismic event analysis, Specification 4.3.5'.1.2.

c.. Accident Monitoring Instrumentation, Specification 3.3.5.3.

I

d. Fire detection instrumentation, Specification 3.3.5.7.

e. Reactor coolant specific activity analysis, Specification-3.4.5.

f. ECCS actuation,' Specifications 3.5.3.1 and 3.5.3.2.

g. Fire suppression systems, Specifications 3.7.7.1, 3.7.7.2, 3.7.7.3,

.and.3.7.7.5.

h. Fire barrier penetration, Specification 3.7.8.

-i. Liquid. Effluents Dose, Specification 3.11.1.2.

j.- Liquid'Radwaste Treatment, Specification 3.11.1.3.

k. Dose - Noble Cases, Specification 3.11.2.2.

1. Dose 3 Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form, Specification 3.11.2.3.

m. ' Caseous Radwaste Treatment, Specification 3.11.2.4.

n. Ventilation Exhaust Treatmeng, Specification 3.11.2.5.

o. Total Dose, Specification 3.11.4.

p. Monitoring Program, Specification 3.12.1.b.

q. Primary Containment Structural Integrity, Specification 4.6.1.4.2 CORE OPERATING LIMITS REPORT 6.9.3.1 Core operating limits shall be established prior to each reload

~ cycle, or prior to any remaining portion of a reload cycle, for the following:

a. The AVERACE PLANAR LINEAR HEAT CENERATION RATES (APLHCR) for ,

Specification 3.2.1. l 1

BRUNSWICK - UNIT 1 6-25 Amendment No.

(BSEP-1-104)

ADMINISTRATIVE CONTROLS I

CORE OPERATINC LIMITS REPORT (Continued)

b. .The Kg' core flow adjustment factor for Specification 3.2.3.1

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.3.1 and 3.2.3.2.

and shall be cocumented in the CORE OPERATING LIMITS REPOT.

6.9.3.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those

- described in the following documents.

a. NEDE-24011-P-A, "Ceneral Electric Standard Application for Reactor Fuel" (latest approved version).

b. The May 18, 1984 and October 22, 1984 NRC Safety Evaluation Reports for the Brunswick Reload Methodologies described in:

1.. Topical Report.NF-1583,01, "A Description and Validation of Steady-State Analysis Meihods for Boiling Water Reactors,"

February 1983.

2. ' Topical Report NF-1583.02, " Methods of RECORD," February 1983.

4, 3. Topical Report NF-1583.03, " Methods of PRESTO-B," February 1983.

4. Topical Report NF-1383.04, " Verification of CP&L Reference BWR Thermal-Hydraulic Methods Using the FIBWR Code," May 1983.

6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-mechanical limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met. ;

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

6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-N45.2.9-1974.

6.10.1 The following records shall be retained for at least five years:

a. Records and logs of facility operation covering time interval st each power level.

BRUNSWICK - UNIT 1 6-26 Amendment No.

)

e 6 (BSEP-1-104)

\

ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued)

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c. All REPORTABLE EVENTS.

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

e. Records of changes made to Operating Procedures.

f. Records of radioactive shipments.

g. Records of sealed source and fission detector leak tests and results.

h. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a. Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of facility radiation and contamination surveys. ,

i

d. Records or radiation exposure for all individuals entering radiation control areas.

c. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those facility components identified in Table 5.7.1-1.

g. Records of reactor tests and experiments.

h. Records of training and qualification for current members of the plant staff.

i. Records of inservice inspections performed pursuant to these Technical Specifications.

j. Records of Quality Assurance activities required by the QA Manual.

BRUNSWICK - UNIT 1 6-27 Amendment No.

- _ _ _ _ _ _ . (

e e .;

q' (BSEP-1-104)

ADMINISTRATIVE CONTROLS RECORDS RETENTION (Continued)

k. Records of reviews performed for changes made to. procedures or equipment or reviews of tests and experiments pursuant to-10 CFR 50.59.

l.- Records of the service lives of all hydraulic and mechanical snubbers referenced in Section 3.7.5 including the data at which the service-Life commences and associated installation ~and maintenance-records,

m. Records of analyses required by the radiological environmental monitoring program.

n. Records of (1) meetings of'the PNSC, (2) meetings of the. previous off-site' review organization, the Company Nuclear Safety Committee (CNSC), (3) the independent reviews performed by the Corporate Nuclear Safety Section, and (4) the independent reviews performed by.

the Corporate Quality Assurance Audit Program, Performance Evaluation Unit.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation prote'etion shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation-exposure.

6.12 HICH RADIATION AREA 6.12.1 In lieu of the " Control Device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radia' tion area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously. posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the !

radiation dose rate in the area. ]

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.: Entry into such areas with this monitoring device may be made after.the dose rate levels in the area have been established and personnel have been made knowledgeable of them.

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

BRUNSWICK - UNIT 1 6-28 Amendment No.

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I (BSEP-1-104) !

I l- !

I ADMINISTRATIVE CONTROLS _

1 l

j HICH RADIATION AREA (Continued) .

c. An individual qualified in radiation protection procedures who is I i

equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit. .

)

6.12.2 The requirements of 6.12.1 above shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Operations Shift Foreman on duty and/or the Radiation Control Supervisor.

6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.13.1 The OFFSITE DOSE CALCULATION MANUAL (ODCH) shall be approved by the Corrmission prior to implementation.

6.13.2 Licensee initiated changes to the ODCM:

a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

1. Sufficiently detailed information to totally support rationale without benefit of additional or supplernental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

2. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and,

3. Documentation of the fact that the change has been reviewed and found acceptable by the PNSC.

b. Shall become effective upon review and acceptance by the PNSC.

6.14 PROCESS CONTROL PROGRAM (PCP) 6.14.1 The PROCESS CONTROL PROCRAM (PCP) shall be approved by the Commission prior to implementation. I 6.14.2 Licensee initiated changes to the PCP:

l BRUNSWICK - UNIT 1 6-29 Amendment No. l 1

(BSEP-1-104)

ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (PCP) (Continued) ,

a. Shall be suhmitted to the Commission in the Semiannual Radioactive -

Effluent Release Report for the period in which the change (s) was l made. This submittal shall contain: 1

1. Sufficiently detailed information to totally support the {

rationale for the change without benefit of additional or supplemental information;

2. A determination.that the change did not reduce the overall conformance of the solidification waste product to existing criteria for solid wastes; and

3. Documentation of the fact that the change has been reviewed and found acceptable by the PNSC. j

b. Shall become effective upon review and acceptance by the PNSC.

6.15 MAJOR CHANGES TO LIQUID, CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS 7I 6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Semiannual Radioactive Effluen: Release Report for the period in which the evaluation was reviewed by the PNSC. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59; 2 .- Suf ficient detailed information to totally suppor- the reason l

for the change without benefit of additional or supplemental

! information;

3. A detailed description of the equipment, components, and processes involved and the incerfaces with other plant systems;

4. An evaluation of the change that shows the predicted release of radioactive materials in liquid and gaseous effluents and/or quantity of sclid waste that differ from those previously predicted in the license application and amendments thereto;

5. An evaluation of the change that shows the expected maximum exposure to an individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; 1

l l 7/ Licensees may choose to submit the information called for in this Specification as part of the annual FSAR update.

BRUNSWICK - UNIT 1 6-30 Amendment No.

'?,:

r y. .

>- v ' )

(BSEp-1-104) :

I i

l ADMINISTRATIVE CONTROLS i 4 .. c ;

MAJOR CHANCES'TO LIQUID,' CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS

^(Continued)- .I j

~

6.- A comparison of the. predicted releases of radioactive materials',-

in liquid and gaseous effluents and in solid waste, to-the

?!

actual. releases for the period prior to'when the? changes are to 1 be made; ,

7.. -An estimate of the exposure to plant' operating personnel as a i' result of the change; and 1

8. Documentation of the fact that the change was reviewed.and found ;

acceptable to'the PNSC. j

b. Shall become effective upon review and acceptance by the-PNSC.-

1 BRUNSWICK - UNIT 1 6-31 Amendment No.

v= .

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y 1...

ENCLOSURE 6 1

[-. BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2-NRC. DOCKETS 50-325 & 50-324i OPERATING LICENSES DPR-71 & DPR-62' SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT'-

- ELIMINATION.OF CYCLE SPECIFIC ~ PARAMETER' LIMITS

-. TECHNICAL SPECIFICATION PAGES - UNIT'2 .

p i

1:

.~'

L 1-h n

e + - = - -

I

e hI ' .T.Ccef%

, <3 lE'. L (BSEP-2-105)L ..

1 a H_

i INDEX-q-

s . . .

. DEFINITIONS-

, D1 ,

SECTION'

n -

En '1.0 DEFINITIONS PACE ACTION'...........................................~............... 1-1 1

- AVERACE P LANAR EX PO S URE . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . 1-1

'AVERACE' PLANAR LINEAR HEAT CENERATION RATE .......................:1-1 CHANNEL-CALIBRATION'...;......................................... 1-1 q C H A NN E L C H E C K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 CHANNEL FUNCTIONAL' TEST .......................................... 1-1

-CORE ALTERATION'................................'.................. 1-2

. 4 CORE MAXIMUM AVERACE ' PLANAR LINEAR HEAT GENERATION RATE RATIO. . .. 1-2 CORE OPERATING LIMITS REPORT..................................... 1-2 CRITICAL POWER RATIO ............................................. 1-2 DOSE EQUIVALENT I-131 ........................................... 1-2.

-E -AVERACE DISINTEGRATION ENERGY ................................. 1-2

'EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ............... 1-3 END-OF-CYCLE. RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME ........ 1-3 FRACTION OF RATED THERMAL P0WER.................................. 1-3 l FREQUENCY N0TATION .............................................. 1-3 CAS EOUS RADWASTE TREATMENT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

~

IDENTIFIED LEAKACE .............................................. 1-3 ISOLATION SYSTEM RESPONSE TIME .................................. 1-4 LIMITING CONTROL ROD PATTERN ................................ ... 1-4 i

LOGIC SYSTEM FUNCTIONAL TEST .................................... 1-4' MAXIMUM AVERACE PLANAR HEAT CENERATION RATE RATIO................ 1-4 l MEMBER (S) OF THE PUBLIC ......................................... 1-4 MINIMUM CRITICAL POWER RATIO ..................................... 1-4 ODYN OPTION A.................................................... 1-4 J 0DYN OPTION B..................................................... 1-4 0FFSITE DOSE CALCULATIONAL MANUAL (ODCM) . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 OPERABLE - OPERABILITY .......................................... 1-5 OPERATIONAL CONDITION ........................................... 1-5 1

l j

BRUNSWICK - UNIT 2 I Amendment No.

1

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~~fD % ,j aan s.

o r <

(BSEP-2-105) 1 1

INDEX l l

1 1

,1 DEFINITIONS 1 SECTION 3 t

1.0 ' DEFINITIONS (Continued) PAGE.

PHYSICSETESTS .................................................... 1-5

.PRESSUkE. BOUNDARY LEAKAGE ........................................-1-5 PRIKARY CONTAINMENT INTEGRITY ................................... 1-5'

.. PROCESS' CONTROL PROGRAM'(PCP) .................................... 1-6

'PURCE - PURGING ................................................. 1-6 RATED THERMAL POWER .............................'................ 1-6 REACTOR PROTECTION SYSTEM RESPONSE TIME ......................... 1-6 j REFERENCE LEVEL ZERO ............................................. 1-6 REPORTABLE EVENT.................................................. 1 ROD DENSITY ..................................................... 1-7 SECONDARY CONTAINMENT INTECRITY ................................. 1-7 SHUTDOWN MARCIN ..................................................'l-7

. SITE BOUNDARY ................................................... 1-7 SOLIDIFICATION .................................................. 1-7 SOURCE CHECK .................................................... 1-7 SPIRAL RELOAD ................................................... 1-8 SPIRAL UNLOAD ................................................... 1 STACCERED TEST BASIS ............................................ 1-8 THERMAL POWER ................................................... 1-8 UNIDENTIFIED LEAKAGE ............................................ 1-8

. UN R E S TR I CT E D AR E A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

. VENTI LATION ' EXHAUST TREATMENT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-9 VENTING'......................................................... 1-9 FREQUENCY NOTATION, TABLE 1.1 ................................... 1-10 OPERATIONAL CONDITIONS , TABLE 1. 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-11 i

i

. BRUNSWICK - UNIT 2 II Amendment No.

, ~

d I

._4. ,[ g ,

(BSEP-2-105)-

LINDEX

' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

.SECTION PACE 2.1 SAFETY LIMITS t

Thermal' Power (Low Pressure or Low Flow)........................... 2-1 Thermal Power.(High. Pressure and-High F1ow)....................... 2-1.

Reactor Coolant System Pressure................................... 2-1 Reactor Vessel Water Leve1........................................ 2-2.

2.2 LIMITING SAFETY SYSTEM SETTINGS- ,

Reactor Protection System Instrumentation Setpoints............... 2-3 BASES 2.1 SAFETY LIMITS Thermal Power (Low Pressure or Low Flow)........................... B 2-1 1

Thermal Power (High Pressure and High F1ow)........................ B 2-2 Reactor Coolant System Pressure.................................... B 2-3 Reactor Vessel Water Level......................................... B 2-3 2.2- Limiting Safety System Settings

. Reactor Protection System Instrumentation Setpoints................ B 2-4

)

. BRUNSWICK - UNIT 2 III Amendment No.

. +

(BSEP-2-105) l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE l

3/4.0 APPLICABILITY.............................................. 3/4 0-1

'3/4.1 REACTIVITY CONTROL SYSTEMS L 3/4.1.1 SHUTDOWN MARCIN.......................................... 3/4 1-1 l

L 3/4.1.2 REACTIVITY AN0MALIES..................................... 3/4 1-2 3/4.1.3 CONTROL RODS .

Control Rod Operability.................................. 3/4 1-3 Control Rod Maximum Scram Insertion Times................ 3/4 1-5 Control Rod Average Scram Insertion Times................ 3/4 1-6 Four Control Rod Group Insertion Times................... .3/4 1-7 Control Rod Scram Accumulators........................... 3/4 1-8 Control Rod Drive Coupling..........'..................... 3/4 1-9 Control Rod Position Indication.'......................... 3/4 1-11 Control Rod Drive Housing Support........................ 3/4 1-13 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer...................................... 3/4 1-14 Rod Sequence Control System.............................. 3/4 1-15 Rod Block Monitor........... ............................. 3/4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................ 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE............... 3/4 2-1 3/4.2.2 APRM SETP0INTS........................................... 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO............................. 3/4 2-3 1

fl l

BRUNSWICK - UNIT 2 IV Amendment No. !

1

.e. ,

l (BSEP-2-105)

E o

INDEX

. : BASES 1

'SECTION PACE 3/4.0 APPLICABILITY.............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 . SHUTDOWN MARCIN.................................... B 3/4 1-l' l 3/4.1.2 REACTIVITY AN0MALIES.'..............................- B 3/4 1-1 3/4.1.3 ' CONTROL R0DS....................................... B 3/4.1-l' 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS....................... B 3/4'1-3 3/4.1.5 ' STANDBY-LIQUID CONTROL SYSTEM...................... B'3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR' LINEAR HEAT CENERATION' RATE......... B 3/4 2 3/4.2.2- APRM SETP0INTS..................................... B 3/4 2-1?

3/4.2.3 MINIMUM CRITICAL POWER RATI0.......................B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1' iREACTOR PROTECTION SYSTEM INSTRUMENTATION.......... B 3/4 3-1

~3/4.3.2' ISOLATION ACTUATION INSTRUMENTATION................ B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.................................. B 3/4 3-2 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION. . . . . . . B 3/4 3-2 3/4.3.5' HONITORING INSTRUMENTATION......................... B 3/4 3-2 3/4.3.6 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION... .............................. B 3/4 3-6 3/4.3.7 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION........................ B 3/4 3-7 3/4.4. REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM............................... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES............................... B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE..................... B 3/4 4-1 1

i BRUNSWICK - UNIT 2 X Amendment No.

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"O INDEX.

1'

. ADMINISTRATIVE-CONTROLS ~

SECTION '

- PACE-6.5.4 CORPORATE NUCLEAR' SAFETY.SECTION Function................................................

. 6-13 Organization.............................................- 6-13 Review.'................................................. '6-14 Records................................................. 6 6.5.5 .CORPORATEQUALITY' ASSURANCE AUDIT PROGRAM Function.s.......................'....................... 6-16 Audits...............................................'... ,6-16

.;. Records.................................................. 6-17 Authority........................'....................... 6 Personne1............................................... 6-17 6.5.6 OUTSIDE AGENCY INSPECTION AND-AUDIT-PROGRAM............. 6-18

'6.6 REPORTABLE EVENT ACTION.................................... 6-18 6.7 SAFETY LIMIT VIOLATION..................................... 6-18 .

6.8 PROCEDURES-AND PR0 CRAMS.................................... 6-19 6.9 REPORTING REQUIREMENTS R$utine Reports ........................................ 6-20 Startup Reports......................................... 6-20 Annual Reports.......................................... 6-21 Personnel Exposure and Monitoring Report................ 6-21 Annual Radiological Environmental Operating Report...... 6-22 Semiannual Radioactive Effluent Release Report.......... 6-23 Monthly Operating Reports............................... 6-24 Special Reports......................................... 6 Core Operating Limits Report............................ 6-25 BRUNSWICK - UNIT 2 XV Amendment No.

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.(BSEP-2-105)-

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.10 RECORD RETENTION..........................................- 6-26 6.11 RADIATION PROTECTION PR0 CRAM.............................. 6 6.12 HICH RADIATION AREA.............~.......................... 6-28 6.13 0FFS ITE DOSE CALCULATION MANUAL (0DCM) . . . . . . . . . . . . . . . . . . . . ' 6-29

' 6.14' PROCESS CONTROL PROGRAM (PCP)............................. 6 ' 6.15 MAJOR CHANCES TO LIQUID, CASEOUS, AND L SOLID WASTE TREATMENT SYSTEMS.......................... 6-30

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BRUNSWICK - UNIT 2 XVI Amendment No.

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-(BSEP-2-105) l 1.0 DEFINITIONS 'f j

The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type f and are applicable throughout these Technical. Specifications. i ACTION ACTIONS are those additional requirements specified as corollary' statements to '

each specification and shall be part of the specifications.

AVERACE PLANAR EXPOSURE The AVERACE PLANAR EXPOSURE shall be applicab' le to a specific planar height and is equal to the sum of the exposure of all of the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at that height.

l AVERACE PLANAR LINEAR HEAT CENERATION RATE The AVERACE PLANAR LINEAR HEAT CENERATION RATE (APLHCR) shall be applicable to a specific planar height and is equal to the' sum of the heat generation rate per unit length of fuel rod for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at that height.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment as necessary of the channel-output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be 4e qualitative assessment of channel behavior during operation by observation, %is determination shall include, where possible, comparison of the channel 4.ation and/or status with other indication and/or status derived from 1.idependent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel -

as close to the primary sensor as practicable to verify l OPERABILITY including alarm and/or trip functions.

BRUNSWICK - UNIT 2 1-1 Amendment No.

'i. r (BSEP-2-105)

DEFINITIONS CHANNEL FUNCTIONAL TEST (Continued)

b. Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions.

CORE. ALTERATION CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative location.

I CORE MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO '

The CORE MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO (CMAPRAT) shall be the largest value of the MAPRAT that exists in the core.

CORE OPERATING LIMITS REPORT The CORE OPERATINC LIMITS REPORT is the unit-specific document that provides core operating limits for the current reload ' cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.3.1, 6.9.3.2, 6.9.3.3, and 6.9.3.4. Plant operation within these core operating limits is addressed in individual specifications.

CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is-calculated, by application of an NRC approved correlation, to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be concentration of I-131, pCi/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The following is defined equivalent to 1 pCi of I-131 as determined from Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites": I-132, 28 pCi; I-133, 3.7 uCi; I-134, 59 uCi; I-135, 12 pCi.

E -AVERACE DISINTEGRATION ENERGY E shall be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes with half lives greater than 15 minutes making up at least 95% of the total non-iodine activity in the coolant.

BRUNSWICK - UNIT 2 1-2 Amendment No.

g i' r; (BSEP-2-105)- )

~ DEFINITIONS .

I EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE: TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the-channel sensor until the ECCS equipment is capable of performing lits' j safety function (i.e., the valves travel to their required positions, pump' 1

discharge pressures reach their required-values, etc.). Times shall' include diesel generator' starting and sequence loading delays where applicable.'. ;

'END-OF-CYCLE RECIRCULATION PUMP TRIP' SYSTEM RESPONSE TIME .

)

1 J

'The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM' RESPONSE TIME shall be that

. time interval to recirculation pump breaker trip from initial movement of the associated:

W Turbine stop valves, and a.

i

b. ' Turbine control valves.

FRACTION OF~ RATED THERMAL' POWER i

.The FRACTION OF RATED-THERMAL POWER (FRTP) shall be the measured THERMAL POWER

' divided by the RATED THERMAL POWER.

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

CASEOUS RADWASTE TREATMENT SYSTEM A CASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive' gaseous effluents by. collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or

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

LEAKAGE.

{

BRUNSWICK - UNIT 2 1-3 Amendment No.

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(BSEP-2-105) p 1

DEFINITIONS l

ISOLATION SYSTEM RESPONSE TIME -

The ISOLATION SYSTEM RESPONSE TIME shall,be that time interval from when the monitored' parameter exceeds its isolation actuation setpoint'at the channel sensor until the isolation valves travel to their required positions. Times

'shall include diesel generator starting and sequence loading delays where applicable.

LIMITING CONTROL ROD PATTERN A LIMITING CONTROL ROD. PATTERN shall be a pattern which results in the core

,- being on a' limiting value for APLHCR or MCPR.

LOGIC SYSTEM FUNCTIONAL TEST A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all relays and contacts of a

' logic circuit, from sensor output to activated device, to ensure that components are OPERABLE.

MAXIMUM AVERACE PLANAR LINEAR HEAT GENERATION RATE RATIO The MAXIMUM'AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO (MAPRAT) for a !

-bundle shall be the maximum ratio of the APLHCR at a specific height in the bundle' divided by the exposure dependent APLHCR limit for that specific height.

MEMBER (S) 0F THE PUBLIC-MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the

. utility, it.s contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category d6es not include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

ODYN OPTION A -

ODYN OPTION A shall be analyses which refer to minimum critical power ratio limits which are determined using a transient analysis plus an analysis uncertainty penalty.

ODYN OPTION B ODYN OPTION B shall be analyses which refer to minimum critical power ratio limits determined using a transient analysis which includes a requirement for 20% scram insertion times to reduce the analysis uncertainty penalty.

BRUNSWICK - UNIT 2 1-4 Amendment No.

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(B5EP-2-105)

DEFINITIONS SPIRAL RELOAD -

A SPIRAL RELOAD is the reverse of a SPIRAL UNLOAD. Except for fuel bundles around each of the four SRMs, the fuel in the interior of the core, symmetric to the SRMs, is loaded first. Up to four fuel bundles may be loaded around each of the four SRMs.

SPIRAL UNLOAD A SPIRAL UNLOAD is a core unload performed by first removing the fuel from'the outermost control cells (four bundles surrounding a control blade). Unioading continues in a spiral fashion by removing fuel from the-outermost periphery to the interior of the core, symmetric about the SRMs, except for fuel bundles around each of the four SRMs.- Up to four fuel bundles may be left around each SRM to maintain adequate count rate.

STACCERED TEST BASIS

-A STACCERED TEST BASIS shall consist of:

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

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

THERMAL POWER THERMAL POWER shall be the total reactor core heat tranefer rate to the reactor coolant.

UNIDENTIFIED LEAKACE UNIDENTIFIED LEAKACE shall be all leakage which is not IDENTIFIED LEAKACE.

I BRUNSWICK - UNIT 2 1-8 Amendment No. l

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$ ': g (BSEP-2-105)

DEFINITIONS-UNRESTRICTED AREA '

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

VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive mat 2 rial 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 .

particulate 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 Feature (ESP) atmospheric cleanup systems are not' considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or.ges is not provided or required. Vent, used in system names, does not imply a VENTING process.

BRUNSWICK - UNIT 2 1-9 Amendment No.

(BSEP-2-105) 2.1 SAFETY LIMITS BASES 2.0 The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that l the MINIMUM CRITICAL POWER RATIO (MCPR) is no less than the Safety Limit CPR of Specification 2.1.2. The Safety Limit.MCPR represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fwei cladding is one of the physical barriers which separate 1 the radioactive trater ials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER (Low Pressure or Low Flow)

The use of the NRC approved CPfl correlation is not valid for all l critical power calculations at pressures below 800 psia or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at lowpowerandflowswigtalwaysbegreaterthan4.5 psi. Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. 3Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 800 psia is conservative.

BRUNSWICK - UNIT 2 B 2-1 Amendment No.

s. e (BSEP-2-105)

SAFETY LIMITS BASES (Continued)-

2.1.2 THERMAL POWER (High Pressure and High Flos)

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in' fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate; boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power, result in an

, uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety. Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using an approved critical power correlation. Details of the fuel cladding integrity safety limit calculation are given in Reference 1.

Uncertainties used in the determination of the fuel cladding integrity safety limit and the bases of these uncertainties are presented in Reference 1.

The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution in Brunswick Unit 2 during any fuel cycle could not be as severe as the distribution used in the analysis. The pressure safety limits are arbitrarily selected to be the lowest transient overpressure allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.

Reference

1. NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," latest approved revision.

1 BRUNSWICK - UNIT 2 B 2-2 Amendment No.

t t _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1

(BSEP-2-105)

's ,

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION'SETPOINTS The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the Reactor Trips are set for each parameter. The. Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

1. Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5-decade, 10-range instrument. The trip setpoint'of 120 divisions is active in each of the 10 ranges. Thus, as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. Range 10 allows the IRM instruments to remain on scale at higher power levels to provide for additional overlap and also permits calibration at these higher powers.

The most significant source of reactivity change during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed in Section 7.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just suberitical and the IRMs are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed. The results of this analysis show that the reactor is shut down and peak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above the Safety Limit MCPR of Specification 2.1.2. Based on this analysis, the IRM provides

~

protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Average Power Range Monitor For operatioa at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides an adequate thermal margin between the setpoint and the Safety Limits. This margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor; cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained by the RSCS and RWM. Of all BRUNSWICK - UNIT 2 B 2-4 Amendment No.

(BSEP-2-105) 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

2. , Average Power Range Monitor (Continued) the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally, the heat flux is in near-equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THEMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% APRM trip remains active until the mode switch is placed in the Run position.

The APRM flow-biased trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and, therefore, the monitors respond directly and quickly to changes due to transient operation; i.e., the thermal power of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer. Analyses demonstrate that with only the 120% trip setting, none of the abnormal operational transients analyzed violates the fuel safety limit and there is substantial margin from fuel damage.

Therefore, the use of the flow-referenced trip setpoint, with the 120% fixed setpoint as backup, provides adequate margins of safety.

The APRM trip setpoist was selected to provide an adequate margin for the Safety Limits and yet allows an operating margin that reduces the possibility of unnecessary shutdowns. The flow referenced trip setpoint must j be adjusted by the specified formula in Specification 3.2.2 in order to '

maintain these margins.

3. Reactor Vessel Steam Dome Pressure-High High Pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids, thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase by decreasing heat generation. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the i

BRUNSWICK - UNIT 2 B 2-5 Amendment No.

cd .

^(BSEP-2-105)

REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR 4

LIMITING CONDITION FOR OPERATION-3.1.4.3 Both' Red Block Monitor (RBM) channels shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL' POWER is greater than or equal to 30% of RATED THERMAL POWER.

ACTION:

a. With one RBM channel inoperable, POWER OPERATION may continue i provided that either:

1. The inoperable RBM channel is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. The redundant RBH is demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable RBM is restored to OPERABLE status within 7 days, or

3. THERMAL' POWER is limited such that MCPR will remain above the Safety Limit MCPR of Specification 2.1.2, assuming a single error that results in complete withdrawal of any single control rod that is capable of withdrawal.

Otherwise, trip at least one rod block monitor channel..

b. With both RBH channels inoperable, trip at least one rod block monitor channel within one hour.

SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.

1 BRUNSWICK - UNIT 2 3/4 1-17 Amendment No.

. a (BSEP-2-105) 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT GENERATION RATE.

LIMITING CONDITION FOR OPERATION-3.2.1 During power operation, the AVERACE PLANAR LINEAR HEAT GENERATION RATE (APLHCR) for each type of fuel as a function of axial location and AVERACE PLANAR EXPOSURE shall not exceed limits based on applicabic APLHCR limit values that have been approved for the respective fuel and lattice type and determined by the approved methodology described in CESTAR-II. When hand calculations are required, the APLHGR for each type of fuel as a function of AVERACE PLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) of each type of fuel shown in the i applicable figures in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHCR exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and continue corrective action so that APLHCR is within the required limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWFR within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS i i

4.2.1 All APLHCRs shall be verified to be equal to or less than the limits specified in Specification 3.2.1:

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and i

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHCR.

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

l

, (BSEP-2-105) 1 POWER-DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The flow-biased APRM scram trip setpoint (S) and rod block trip set point (SRB) shall be established according to the following relationship:

S $ (0.66W + 54%) T SRB $ (0.66W + 42%) T where: S and S RB are in percent of RATED THERMAL POWER.

W = Loop recirculation flow in percent of rated flow, T = FRACTION OF RATED THERMAL POWER (FRTP) divided by CORE MAXIMUM AVERACE LINEAR HEAT CENERATION RATE RATIO (CMAPRAT)

(T is applied only if less than 1.0.).

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.-

ACTION:

With S or SRB exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and S RB are within the required limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to Iess than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1 SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and CMAPRAT shall'be determined, the value of T calculated, and the flow-biased APRM scram trip and rod block trip setpoints verified to be within the above limits or adjusted, as required:

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with CHAPRAT greater than or equal to FRTP.

I BRUNSWICK - UNIT 2 3/4 2-2 Amendment No.

. . i (BSEP-2-105) r n

POWER DISTRIBUTION LIMITS

-l 3/4.2.3 MINIMUM' CRITICAL POWER RATIO M LIMITINC CONDITION FOR OPERATION 3.2.3.1 The MINIMUM CRITICAL' POWER RATIO (MCPR),'as a function of core flow and cycle' average exposure, shall be equal to or greater than the MCPR limit times the K 'specified g in the CORE OPERATING LIMITS REPORT. The MCPR limits i for ODYN OPTION A and ODYN OPTION B analyses, used in the above determination, shall be specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is~ greater than or equal to 25% RATED THERMAL POWER ACTION:

With MCPR, as a function of core flow and cycle average exposure, less than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within the applicable limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.3.1 MCPR, as a function of core flow and cycle average exposure, shall be determined to be equal to or greater than the applicable limit determined of Specification 3.2.3.1:

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

b. Within l'2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating in a LIMITING CONTROL ROD PATTERN for MCPR.

BRUNSWICK - UNIT 2 3/4 2-3 Amendment No.

. . .. ,; 4:

s -(BSEP-2-105)'

L POWER DISTRIBUTION LIMITS

{

4- '3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)

/,

LIMITINC'COND )N FOR OPERATION .I j

l 3.2.3.2 For the OPTION B MCPR limits provided in the CORE OPERATINC' LIMITS '!

REPORT'to be used, the cycle average 20% (notch 36) scram time (t,y,) shall be l less than or equal to the Option B scram time limit (tg), where t,y,.and T B are determined as follogs:

[ iI , where r

ave

. I"1 n N.

b i=1 i = Surveillance test number, na Number of surveillatice tests performed to date in the cycle (including BOC), ,

Ni = Number of rods tested in the i'h surveillance test,_and t i.= Average scram time to notch 36 for surveillance test i t g =u+(.65( v n N. ' " *#*

i u I"1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including BOC),

th Ni = Number of rods tested in the i surveillance test Ng = Number of rods tested at BOC, ,

p = 0.813 seconds '

(mean value for statistical scram time distribution from d de-energization of. scram pilot valve solenoid to pickup on notch 36),

o = 0.018 seconds (standard deviation of the above statistical distribution) l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER.

l I

I

(-

BRUNSWICK - UNIT 2 3/4 2-4 Amendment No.

-: l . .(BSEP-2-105) ,

j 4

s

' POWER DISTRIBUTION LIMITS -

h LIMITINC CONDITIONS ~FOR OPERATION (Continued)

]

3. ACTION:. >

Within twelve hours after determining that t,y,'is greatersthan TB, the

. operating limit'MCPRs shall be either: '(

a. ' Adjusted for.each fuel type such that'the operating limit MCPR is the. maximum of the non-pressurization transient MCPR operating limit specified in the CORE OPERATINC. LIMITS REPORT-or the adjusted pressurization transient MCPR dperating limits, s 0 where,the adjustment is made by:

T T MCPR = MCPR (MCPR - MCPR i ad j.usted option B.+ T -T Ptton A option B)' .

A B where: r 1.05 seconds,' control rod average scram insertion time -j A = limit to notch 36 per Specification 3.1.3.3, MCPRoption A

Specified in the CORE OPERATING LIMITS REPORT, MCPR option B

S ecified P in the' CORE OPERATINC LIMITS REPORT, or,

b. The OPTION A MCPR limits specified in the CORE OPERATING LIMITS REPORT.

SURVEII;I;ANCE - REQUIREMENTS 4.2.3.2 The. values of T and t shall be determined and compared each time a scram time test is perfoImed. Yherequirementforthefrequencyofscram time testing shall.be' identical to Sp' 'fication 4.1.3.2.

kl,:

1 i

1 I

lp, .

,4 BRUNSWICK'- UNIT 2 3/4 2-5 Amendment No.

a a (BSEP-2-105)

INSTRUMENTATION t 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION i 3.3.4 l The control rod withdrawal block instrumentation shown in Table 3.3.4-1 i shall be OPERABLE with their trip setpoints set consistent with the values !

shown in the Trip setpoint column of Table 3.3.4-2.

3 APPLICABILITY: As shown in Taste 3.3.4-1.

ACTION:

i

a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable i Values column of Table 3.3.4-2, declare the channel inoperable until I the channel is restored to OPERABLE status with its Trip Setpoint !

adjusted consistent with the Trip Setpoint value. I

b. With the requirements for the minimum number of OPERABLE channels not satisfied for one trip system, POWER OPERATION may continue provided i that either:

1. The inoperable channel (s) is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. The redundant trip system is demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable channel is restored to OPERABLE status, and the inoperable channel is restored to OPERABLE status within 7 days, or

3. For the Rod Block Monitor only, THERMAL POWER is limited such that the MCPR will remain above the Safety Limit MCPR of Specification 2.1.2, assuming a single error that results in complete withdrawal of any single control rod that is capable of l withdrawal.

4. Otherwise, place at least one trip system in the tripped !

condition within the next hour.

c. With the requirements for the minimum number of OPERABLE channels not satisfied for both trip systems, place at least one trip system in the tripped condition within one hour.

d. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5.

1

,S, SURVEILLANCE REQUIREMENTS l y

4.3.4 Each of the above required control rod withdrawal block instrumentation channels shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK, CHANNEL CALIBRATION, and a CHANNEL FUNCTIONAL TEST during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.4-1.

i l

BRUNSWICK - UNIT 2 3/4 3-39 Amendment No. Il

. o (BSEP-2-105)

REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) potential effects of the rod ejection accident are limited. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem; therefore, with a centrol rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period

} which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

l Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the non-fully-inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be : inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shut down for investigation und resolution of the problem.

The control rod system is analyzed to bring the reactor subcritical at a rate fast enough to prevent the MPCR from becoming less than the Safety Limit i MCPR of Specification 2.1.2 during the limiting power transient analyzed in '

Section 14.3 of the FSAR. This analysis shows that the negative reactivity rates resul' ting from the scram with the average response of all the drives as i given in the specifications, provide the required protection and MCPR remains greater than the Safety Limit MCPR of Specification 2.1.2 The occurrence of scram times longer than those specified should be viewed as an indication of a systemic problem with the rod drives and, the-efore, the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion i

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

- - - - _ _ _ - _ _ _ _ - _ \

z

' 1 r ,

(BSEP-2-105) i J

l i

c. POWER DISTRIBUTION LIMITS-BASES

~

\

The specifications of this section assure that the peak cladding temperature l following the postulated design basis loss-of-coolant accident will not exceed !

the 2200*F: limit specified in the Final Acceptance Criteria (FAC) issued in j June 1971 considering the postulated effects.of fuel pellet densification.

3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERiTION RATE

)

The limiti'ng values for 6PLHCR when confermance to.the operating limit is performed by hand calculation are provided in the CORE OPERATING LIMITS REPORT for each fuel typt and, when required, for the most. limiting lattice for multiple lattice foel bundle types.

This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the fuel design analysis limits

~

specified in NEDE-24011-P-A (Reference 1) will not be exceeded.

Mechanical Design Analysis: NRC approved methods (specified in Reference 1) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 1. No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the fuel design analysis APLHCR limit.

LOCA Analysist A LOCA analysis is performed in accordance with 10 CFR 50 Appendix K to demonstrate that the permissible planar power (APLHCR) Limits comply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the'most limiting break size, break location, and single failure combination for the plant.

The Technical Specification APLHCR limit is the most limiting composite of the fuel mechanical design analysis APLHCR and the ECCS APLHCR limit.

I 1

i BRUNSWICK - UNIT 2 B 3/4 2-1 Amendment No.

L

'~

=-.

(BSEP-2-105)

-POWER DISTRIBUTION LIMITS BASES' 3/4.2.2 APRM SETPOINTS The scram' setting and rod block functions of the APRM instruments are adjusted to ensure that fuel design and safety limits are not exceeded. The scram settings and rod block settings are adjusted in accordance with the relationship provided in Specification 3.2.2. This adjustment may be accomplished by increasing the APRM gain and thus reducing the' slope and intercept point of the flow referenced APRM high flux scram curve by the reciprocal of the APRM gain change.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from an established fuel cladding H integrity Safety Limit MCPR approved by the NRC and an analysis of abnormal operational' transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state ,

operating limit, it is required that the resulting MCPR does not decrease j below the Safety Limit MCPR at any time during the transient, assuming an instrument trip setting as given in Specification 2.2.1. l To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

Details on how evaluations are performed, on the methods used, and how the i MCPR limit is adjusted'for operation at less than rated power and flow conditions are given in Reference 1 and the CORE OPERATINC LIMITS REPORT.

At core thermal power levels less than or equal to 25% RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating H:.PR above 25% RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.

_- _ _ _ _ _ _ J

(BSEP-2-105)

POWER DISTRIBUTION LIMITS BASES

References:

1. NEDE-24011-P-A, "Ceneral Electric Standard Application for Reactor Fuel", latest approved version.

BRUNSWICK - UNIT 2 B 3/4 2-3 Amendment No. I l

l' (BSEP-2-105) 1 I ADMINISTRATIVE CONTROLS l

~

SPECIAL REPORTS 6.9.2 .Special. reports shall be submitted to the-Regional Administrator of the Regional Office within the time period specified for each report. These. j reports shall be submitted covering the activities identified below pursuant j to the' requirements of the applicable reference specification. '

i

a. Inoperable Seismic Monitoring Instrumentation, specification 3.3.5.1. !

b.- ' Seismic event analysis, Specification 4.3.5.1.2.

c. Accident Monitoring Instrumentation, Specification 3.3.5.3.

d. Fire detection instrumentation, Specification 3.3.5.7.

e. Reactor coolant specific activity analysis, Specification 3.4.5.

f. ECCS actuation, Specifications 3.5.3.1 and 3.5.3.2.

g. Fire suppression systems, Specifications 3.7.7.1, 3.7.7.2, 3.7.7.3, and 3.7.7.5.

h. Fire barrier penetration, Specification 3.7.8.

i. Liquid Effluents Dose, Specification 3.11.1.2.

j. Liquid Radwaste' Treatment, Specification .3.11.1.3.

k. Dose - Noble Cases, Specification 3.11.2.2.

1. Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Pa'rticulate Form, Specification 3.11.2.3. i

m. Caseous Radwaste Treatment, Specification 3.11.2.4. l

n. Ventilation Exhaust Treatment, Specification 3.11.2.5.

o. Total Dose, Specification 3.11.4.

p. Monitoring Program, Specification 3.12.1.b.

q. Primary Containment Structural Integrity, Specification 4.6.1.4.2 CORE OPERATINC LIMITS REPORT 6.9.3.1 Core operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a. The AVERACE PLANAR LINEAR HEAT CENERATION RATES (APLHCR) for Specification 3.2.1.

BRUNSWICK - UNIT 2 6-25 Amendment No.

s a (BSEP-2-105)

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

b. The Kg core flow adjustment factor for Specification 3.2.3.1

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.3.1 and 3.2.3.2.

i and shall be documented in the CORE OPERATING LIMITS REPOT.

6.9.3.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

a. NEDE-24011-P-A, "Ceneral Electric Standard Application for Reactor j Fuel" (latest approved version).

b. The May 18, 1984 and October 22, 1984 NRC Safety Evaluation Reports !

for the Brunswick Reload Methodologies described in:

1. Topical Report NF-1583.01, "A Description and Validation of Steady-State Analysis Methods for Boiling Water Reactors,"

February 1983.

2. Topical Report NF-1583.02, " Methods of RECORD," February 1983.

3. Topical Report NF-1583.03, " Methods of PRESTO-B," February 1983.

{

4. Topical Report NF-1583.04, " Verification of CP&L Reference BWR Thermal-Hydraulic Methods Using the FIBWR Code," May 1983.

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

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

6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-N45.2.9-1974.

6.10.1 The following records shall be retained for at least five years:

a. Records and logo of facility operation covering time interval at each power level.

l BRUNSWICK - UNIT 2 6-26 Amendment No.

1.4 . , :.

.(BSEP-2-105)

T.

-ADMINISTRATIVE CONTROLS RECORD RETENTION.(Continued)

b. : Records and logs of principal maintenance activities, inspections, repair and replacement of prin'ipal c items of equipment related'to nuclearisafety..

c.~ All' REPORTABLE EVENTS.

d. Records'of surveillance activities, inspections,.and calibrations required by these Technical Specifications.

e. Records of changes made to Operating Procedures. _.

f. Records of radioactive shipments.

g .' . Records of sealed source and fission detector leak tests and results.

h. Records of annual physical inventory of all sealed' source material of-record.

6.10.2 The.following records'shall be retained for the duration of the' Facility Operating License: j

a. Records'and drawing changes reflecting facility design. modifications "made..tol systems and equipment described in the Final Safety Analysis Report.

b. Records'of new.and irradiated fuel inventory, fuel transfers and assembly.burnup histories.

l 4 c. ' Records of facility radiation and contamination surveys. I d.- Records or radiation exposure for all individ'uals entering radiation control areas,

e. Records of gaseous.and liquid radioactive material released to the environs'.

i

f. Records of transient.or operational cycles for those facility l

l. . components identified in Table 5.7.1-1.

l l g. Records of reactor tests and experiments.

l h .- Records of training and qualification for current members of the plant staff. l

i. Records of inservice inspections performed pursuant to these Technic.al' Specifications. I

j. Records of Quality Assurance activities required by the QA Manual.

BRUNSWICK.- UNIT 2 6-27 Amendment No.

___1________-____-- _

4 e (BSEP-2-105) i ADMINISTRATIVE CONTROLS RECORDS RETENTION.(Continued)

k. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

1. Records of the service lives of all hydraulic and mechanical snubbers referenced in Section 3.7.5 including the data at which the service life commences and associated installation and maintenance records.

m. Records of analyses required by the radiological environmental monitoring program.

n. Records of (1) meetings of the PNSC, (2) meetings of the previous off-site review organization, the Company Nuclear Safety Committee j (CNSC), (3) the independent reviews performed by the Corporate Nuclear Safety Section, and (4) the independent reviews performed by l the Corporate Quality Assurance Audit Program, Performance Evaluation !

Unit.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HICH RADIATION AREA 6.12.1 In lieu of the " Control Device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity o'f radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high; radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device ,

may be made after the dose rate levels in the area have been j established and personnel have been made knowledgeable of them. 4 1

I

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance .

of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation i areas.

BRUNSWICK - UNIT 2 6-28 Amendment No. l

4- ,

(BSEP-2-105)

ADMINISTRATIVE CONTROLS HICH RADIATION AREA (Continued) c.- An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.

6.12.2 The requirements of 6.12.1 above shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Operations Shift Foreman on duty and/or the Radiation Control Supervisor.

6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.13.1 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be approved by the Commission prior to implementation.

6.13.2 Licensee initiated changes to the ODCM:

a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

1. Sufficiently detailed information to totally support rationale without benefit of additional or supplemental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

2. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and,

3. Documentation of the fact that the change has been reviewed and found acceptable ey the PNSC.

b. Shall become effective upon review and acceptance by the PNSC.

6.14 PROCESS CONTROL PROGRAM (PCP) 6.14.1 The PROCESS CONTROL PROGRAM (PCP) shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the PCP:

BRUNSWICK - UNIT 2 6-29 Amendment No. l

~d (BSEP-2-105)

ADMINISTRATIVE CONTROLS PROCESS CONTROL PROCRAM (PCP) (Continued) .

a. Shall be submitted to the Commission in the Semiannual Radioactive-Effluent Release Report for the period in which the' change (s) was made. This submittal shall contain:

1. Sufficiently detailed'information to totally support.the rationale for the change without benefit of additional.or supplemental information;

2. A determination that the change did not reduce the overall conformance of the solidification waste product to existing criteria for solid wastes; and

3. Documentation of the fact that the change has been reviewed and found' acceptable by the PNSC.

b. Shall become effective upon review and acceptance by the PNSC.

6.15 MAJOR CHANGES TO LIQUID, CASE 0VS, AND SOLID WASTE TREATMENT SYSTEMS 7/

6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous,-and solid):

a. -Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PNSC. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59;

2. Sufficient detailed information to totally support the reason for the change without benefit of additional or. supplemental information;

' 3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

4. An evaluation of the change that shows the predicted release of radioactive miterials in liquid and geseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

5. An evaluation of the change that shows the expected maximum exposure to an individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; 7/ Licensees may choose to submit the information called for in this Specification as part of the annual FSAR update.

BRUNSWICK - UNIT 2 6-30 Amendment No.

o. .,

(BSEP-2-105)

ADMINISTRATIVE CONTROLS MAJOR CHANCES TO LIQUID, CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS (Continued)

6. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

7. An estimate of the exposure to plant operating personnel as a result of the change; and

8. Documentation of the fact that the change was reviewed and found acceptable to the.PNSC.

b. Shall become effective upon review and acceptance by the PNSC.

s

)

I 1

BRUNSWICK - UNIT 2 6-31 Amendment No.

7 v .

'~

"1, I.$ i0[ [

q

q +

z..

' ..ENCIDSURE 7

}.:

~

BRUNSWICK ~ STEAM ELECTRIC PLANT,' UNITS 1 AND 2 NRC-DOCKETS 50-325 & 50-324~

OPERATING LICENSES DPR-71'& PPR '

' SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT

- e. ELIMINATION OF CYCLE-SPECIFIC PARAMETER LIMITS CORE OPERATING LIMITS' REPORT UNIT 1. CYCLE 7 i

_ _ . _ _ . . _ . _ _ _ _ _ _ _ _ . . _ . - _ _ _ _ _ _ . _ . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ ________

ML20235Q840

Text

References:

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