Text

Amends 24 & 48 to Licenses DPR-71 & DPR-62,respectively, Providing Tech Specs for Protective Instrumentation Associated W/Atws Recirculation Pump Trip
	ML19289F454
Person / Time
Site:	Brunswick
Issue date:	05/02/1979
From:	Ippolito T; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19289F455	List: ML19289F454; ML19289F463; ML19289F467;
References
	NUDOCS 7906070391
	Download: ML19289F454 (68)
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{{#Wiki_filter:. 6 o# 4,$ UNITED STATES j NUCLEAR REGULATORY COMMISSION ^ 3" $I h WASHINGTON, D. C. 20555 _'e i s;.v / CAROLINA POWER & LIGHT COMPANY DOCKET N0. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 24 License No. DPR-71 1. The Nuclear Regulatory Conmission (the Conmission) has found that: A. The facility will operate in conformity with the provisions of the Atomic Energy Act of 1954, as amended (the Act), and the~ rules and regulations of the Commission; B. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; C. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and D. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requiremer.ts have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-71 is hereby amended to read as follows: 2.C(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 24, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 2232 027 7906070893

. 3. This license amendment is effective as of the date of issuance. FOR lHE NUCLEAR REGULATORY COMMISSION l'() ,e T omas A ppolito, Chief Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: May 2, 1979 2232 028

ATTACHMENT TO LICENSE AMENDMENT NO. 24 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following pages of the Technical Specifications contained in Appendix A of the above indicated license with +he attached pages. The changed area of the revised page is reflected by a marginal line. Remove Insert V V VI* VI* IX* IX* X X 3/4 3-61* 3/4 3-61

3/4 3-62 3/4 3-63 3/4 3-64 3/4 3-65 3 3/4 3-3*

B 3/4 3-3 B 3/4 3-4 B 3/4 3-4

0verleaf pages supplied for convenience 2232 029

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLAN PAGE SECTION_ /4.2 POWER DISTRIBUTION LIMITS (Continued) 3 3/4 2-9 LINEAR HEAT GENERATION RATE............................. 3/4.3 INSTRUMENTATION 3/4 3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............... 3/4.3.1 3/4 3-9 ISOLAT ION ACTU ATION INSTRUMENTATION..................... 3/4.3.2 3/4 3-30 EMERGENCY CORE COOLING SYSTEM ACTUATION INST 3/4.3.3 3/4 3-39 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION......... 3/4.3.4 3/4.3.5 MONITORING INSTRUMENTATION 3/4 3-44 Seismic Monitoring Instrumentation...................... 3/4 3-47 Remote Shutdown Monitoring Instrumenta tion.............. 3/4 3-50 Post-accident Monitoring Instrumentation................ 3/4 3-53 Source Range Monitors................................... 3/4 3-54 Chlorine Detection System............................... 3/4 3-55 Chloride Intrusion Monitors............................. 3/4 3-59 Fire Detection Instrumentation.......................... 3/43-62l ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMEN 3/4.3.6 3/4.4 REACTOR COOLANT SYSTEM 2232 030 3/4.'4'.1 l RECIRCULATION SYSTEM t 3/4 4-1 s Re c i r cul a t i o n lo o p s..................................... 3/4 4-2 Jet Pumps............................................... 3/4 4-3 Idle Recirculation loop Startup......................... 3/4 4-4 SAFETY / RELIEF VALVES.................................... 3/4.4.2 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3 3/4 4-5' Leakage Detection Systems............................... 3/4 4-6 Ope ra ti onal Le aka g e..................................... Amendment No. 23, 24 BRUNSWICK - UrdT 1 V, r-

a INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENT PAGE SECTION 3/4.4 REACTOR COOLANT SYSTEM (Continued) 3/4 4-7 3/4.4.4 CHEMISTRY............................................ 3/4 4-10 3/4.4.5 S P E C I F I C ACT IV IT Y.................................... 3/4.4.6 PRESSURE / TEMPERATURE LIMITS 3/4 4-13 Reactor Coolant System............................... 3/4 4-18 Reactor Steam Dome................................... 3/4.4.7 MAIN STENi LINE ISOLATION VALVES..................... 3/4 4-19 3/4 4-20 3/4.4.8 ST RU CTU RAL I NT EG R I T Y................................. 3/4.5 EMERGENCY CORE COOLING SYSTEMS ~ 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM............... 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM.................... 3/4 5-3 3/4.5.3 LOW PRESSURE COOLING SYSTEMS 3/4 5-4 Core Spray System.................................... Low Pressure Coolant Injection System................ 3/4 5-7 3/4 5-9 3/4.5(4j SUPPRESSION P00L..................................... 3/4.6 CONTAINMENT SYSTEMS 2232 03) 3/4.6.1 PRIMARY CONTAINMENT 3/4 6-1 Primary Containment Integrity........................ 3/4 6-2 Primary Containment Leakage.......................... 3/4 6-4 Primary Containment Air Lock................ 3/4 6-6 Primary Containment Structural Integrity............. 3/4 6-7 Primary Containment Internal Pressure................ 3/4 6-8 Primary Containment Average Air Temperature.......... BRU.;SWICK - UNIT 1 VI

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS (. Continued) 3/4.9.3 CONTROL R0D P0SITION............................... 3/4 9-5 3/4.9.4 DECAY TIME......................................... 3/4 9-6 3/4.9.5 COMMUNICATIONS.................................... 3/49-7 3/4.9.6 CRANE AND HOIST OPERABILITY........................ 3/4 9-8 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE P00L............... 3/4 9-9 3/4.9.8 WATER LEVEL-REACTOR VESSEL......................... 3/4 9-10 3/4.9.9 WATER LEVEL -SPENT FUEL STORAGE P00L................ 3/4 9-11 3/4.9.10 CONTROL R0D REMOVAL Single Control Rod Remova1......................... 3/4 9-12 Mul tipl e Co ntrol Rod Removal....................... 3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INT EGRITY...................... 3/4 10-1 3/4.10.2 R0D SEQUENCE CONTROL SYSTEM........................ 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS..................... 3/4 10-3 3/4.10.4 RECIRCULATION L00PS................................ 3/4 10-4 3/4.10.5 R EACTO R V ES SEL DRAI N I N G............................ 3/4 10-5 l 2232 032 BRU.NSWICK - UNIT 1 IX Amendment No. 22 .- 3 j:

INDEX BASES PAGE SECTION B 3/4 0-1 3/4.0 APPLICABILITY...................................... 3/4.1 REACTIVITY CONTROL SYSTEMS B 3/4 1-1 3/4.1.1 SHUTDOWN MARGIN............................ B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES....................... B 3/4 1-1 3/4.1.3 c ')NT RO L R0D S............................... B 3/4 1-3 3/4.1.4 CONTROL R0D PROGRAM CONTROLS............... 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.............. B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATING RATL B 3/4 2-1 B 3/4 2-0 3/4.2.2 APRM SETP0lNTS............................ 3/4.2.3 MINIMUM CRITICAL POWER RATIO............... B 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE................ B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.. B 3/4 3 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION........ B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION B 3/4 3-2 INSTRUMENTATION......................... 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK B 3/4 3-2 INSTRUMENTATION......................... B 3/4 3-2 3/4.3.5 MONITORING INSTRUMENTATION................. 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM B 3/4 3-4 l INSTRUMENTATION.......................... 3'/ 4 '. 4 REACTOR COOLANT SYSTEM B 3/4 4-1 3/4.4.1 RECIRCULATION SYSTEn....................... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES....................... 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE............. B 3/4 4-1 Amendment No. 24 BRUNSWICK - UN'IT 1 X, 2232 033

TABLE 3.3.5.7-1 (Continued) INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE FLAME HEAT SM0KE 4. Service Water Building Zone 1 4' 0 0 6 Zone 2 20 0 0 5 5. A0G Building Zone 1 20' 1 0 0 Zone 2 20' 1 0 0 Zone 3 20' 1 5 1 Zone 4 37' - 49' 1 6 0 2232 034 . i: ? BRUNSWICK - UNIT 1 3/4 3-61 Amendment No.23

INSTRUMENTATION 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3. 3. 6.1 The Anticipated Transient Without Scram recirculation pump trip (ATWS-RPT) system instrumentation trip systems shown in Table 3.3.6.1-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6.1-2. APPLICABILITY: CONDITION 1. ACTION: With an ATWS recirculation pump trip system instrumentation a. trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6.1-2, declare the trip system inoperable until the trip system is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. With the regirements for the minimum number of OPERABLE trip b. sfstems per operating pump not satisfied for one Trip Fdnhion, restore the inoperable trip system to OPERABLE status within 14 days or be in at least STARTUP within the next 8 hoc.s. SURVEILLANCE REQUIREMENTS Each ATw$ recirculation pump trip system instrumentation trip 4.3.6.1.1 system shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.6.1.1-1. LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic opera-4.3.6.1.2 tion of all channels shall be perforced at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system. 2232 035 BRUNSWICK - UNIT 1 3/4 3-62 Amendment No. 24

TABLE 3.3.6.1-1 Eg ATMS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION Ea 3 MINIMUM NUMBER OPERABLE TRIP TRIP FUNCTION AND INSTRUMENT NUMBER SYSTEMS PER OPERATING PUMP E 11.t. I p Reactor Vessel Water Level - " Low Low, Level 2 (B21-LIS-N024 A, B; B21-LIS-N025 A, B) 2. Reactor Vessel Pressure-Low 1 (B21-PS-N045 A, B, C, D) t,' O iT a N a N M u a N Ei co u O

TABLE 3.3.6.1-2 E ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS y{ E TRIP ALLOWABLE n $3 TRIP FbHCTION AND INSTRCiENT NUMBER SETPOINT VALUE > -38 inches < > -38 inches s-El 1. Reactor Vessel. Water Level - low low, Level 2 s (B21-LIS-N024 A,'B; B21-LIS-N025 A, B) 2. Reactor Vessel Pressure-Low 21120 psig 2.1120 psig (821-PS-N045 A, B, C, D) t' ? rs) 3g Ps) o (#J Ps) S CC) 2 (sw P sa I

TABLE 4.3.6.1-1 Ey ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 CHANNEL CHANNEL FUNCTIONAL CHANNEL n [ TRIP FUNCTION AND INSTRUMENT NUMBER CHECK TEST C..!.IB, aTION .Z 1. Reactor Vessel Water Level - S M R 2 Low Low, level 2 ~ (B21-LIS-N024 A, B; B21-LIS-N025 A, B) 2. Reactor Vessel Pressure - Low NA M (B21-PS-N045 A, B, C,.D) t' 8; E N 8 N ~ w N g i

a u

i g.

INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.5.2 REMOTE SHUTDOWN MONITORING INSTRUMENTATION Tne OPERABILITY of the remote shutdown monitoring instrumentation ensures that sufficient capability is available to pemit shutdown and maintenance of HOT SHUTDOWN of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of CFR 50. 3/4.3.5.3 POST-ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the post-accident monitoring instrumentation ensures that sufficient information is available on selected plant perameters to monitor and assess important variables following an accident. This capability is consistent with the recormlendations of Regulatory Guide h 97 " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975. 3/4.3.5.4 SOURCE RANGE MONITORS The source range monitors provide the operator with infonnation on the status of the neutron level in the core at very low power levels during startup. At these power levels reactivity additions should not be made without this flux level information available to the operator. When the intemediate range monitors are on scale adaquate information is available without the SRM's and they can be retracted. 3/4.3.5.5 CHLORINE DETECTION SYSTEf1 The OPERABILITY of the chlorine detection systems ensures that an accidential chlorine release will be detected prompt 13 and the necessary protective actions will be automatically initiated to provide protection for control room personel. Upon detection of a high concentration of chlorine the control room emergency ventilation system will automatically isolate the control room and initiate operation in the recirculation mode to provide _the required protection. The detection systems required by this specifications are consistent with the recommendations of Regulatory Guide 1.95 " Protection of NJClear Power Plant Control Room Operators against an accidental Chlorine Release. BRUNSWICK-UNIT 1 B 3/4 3-3

INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 2/4.3.5.6 CHLORIDE INTRUSTION MONITORS The chloride intrusion monitors provide adequate warni.ig of any leakage in the condenser or hotwell so that actions can be taken to mitigate the consequences of such intrusion in the reactor coolant system. With only a minimum number of instruments available increased sampling frequency provides adequate information for the same purpose. 3/ 4. 3. 5.'i CIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection instrumentation is inoperable, increasing the frequency of fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION The ATWS recirculation pump trip system has been added at the suggestion of ACRS as a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events given in General Electric Company Topical Report NED0-10349, dated March,1971. 2232 040 ( ,U Sk BRUNSWICK-UNIT 1 B 3/4 3-4 Amendment No. 24

UNITED STATE! [ NUCLEAR REGULATORY COMM!sSION 3 f 1 WASHINGTON, D. C. 2055s d([ CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 48 License No. DPR-62 1. The Nuclear Regulatory Conmission (the Comission) has found that: A. The applications for amendments by Car ) lina Power & Light Company (the licensee) dated February 2, 1979, as supplemented March.16, 21 and 27, April 13 and 27, and May 1,1979, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public and E. The' issuance of.this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows: 2232 041

. 2.C(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through knendment No. 48, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specificacions. 3. This license amendment is effective as of the date of issuance. FOR THE NUCLEAR REGULATORY COM'ilSSION i r h T omas ppolito, Chief Operati g Reactors Branch #3 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: May 2,1979 g32 042

ATTACHMENT TO LICENSE AMENDMENT NO. 48 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following pages of t1e Technical Specifications contained in Appendix A of the above indicated license with the attached pages. The changed area of the page is reflected by a marginal line. Remove Insert _ III* III* IV IV V V VI* VI* IX* IX* X X 2-1 2-1 2-2* 2-2* B 2-1 B 2-1 3 2-2* B 2-2* B 2-9 B 2-9 B 2-10* B 2-10* 3/4 1-17 3/4 1-17 3/4 1-18* 3/4 1-18* 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-11 3/4 2-11 3/4 2-12* 3/4 2-12* 3/4 2-13 3/4 2-13 3/4 3-39 3/4 3-39 3/4 3-40* 3/4 3-40* 3/4 3-41* 3/4 3-41* 3/4 3-42 3/4 3-42 3/4 3-43 3/4 3-43 3/4 3,44* 3/4 3-44* 2232 043

. Remove Insert 3/4 3-61* 3/4 3-61* 3/4 3-62 3/4 3-63 3/4 3-64 3/4 3-65 B 3/4 1-l* B 3/4 1-l* B 3/4 1-2 B 3/4 1-2 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4 2-3 B 3/4 2-4* B 3/4 2-4* B 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 B 3/4 3-3* B 3/4 3-3* B 3/4 3-4 8 3/4 3-4 5'-l 5-1 5-2* 5-2*

0verleaf pages supplied for convenience 2232 044

_INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Thermal Power (Low Pressure or Low Flow).................. 2-1 Thermal Power (High Pressure and High Flow)............... 2-1 Reactor Coolant System Pressure........................... 2-1 Reactor Vessel Water Level................................ 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 Thermal Power (High Pressure and High Flow)............... B 2-2 Reactor Coolant System Pressure........................... B 2-8 Reactor Vessel Water Level................................ B 2-8 2.2 Limiting Safety System Settings Reactor Protection System Instrumentation Setpoints....... B 2'9 ~ 2232 045 (h s

e' BRUNSWICK - UNIT 2 III

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION L _E, 1 3/4.0 A P PL I CA B IL I T Y............................................. 3/4 -1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN........................................ 3/41-1 3/4.1.2 R EA CT I V IT Y AN 0 MAL I E S................................... 3/4 1-2 3/4.1.3 CONTROL RODS Con t rol Ro d Op e ra bil i ty................................ 3/4 1-3 Control Rod Max imum Scram Ins erti on Times.............. 3/4 1-5 Control Rod Average Scram Inserti on Times...........,.. 3/41-6 Four Control Rod Group In sertion Times................. 3/41-7 Control Rod Scram Accumulators......................... 3/4 1 8 Con trol Rod Dr iv e Cou pl i ng............................ 3/41-9 Control Rod Posi ti on Indi cati on........................ 3/4 1-11 Con trol Rod Drive Housing Support...................... 3/4 1-13 3/4.1.4 CONTROL R00 PROGRAM CONTROLS Rod Wo rt h t11 n imi ze r.................................... 3/4 1-14 Ro d S e q u e n c e C o n tr ol Sy s t em............................ 3/4 1-15 Ro d B1 o ck M on i tor.......,,,............................ 3/4 1-17 3/4.1.5 STAND BY LIQU ID CONTROL SY STEM.......................... 3/4 1-3/4.2 POWER DISTRIBUTION LIMITS AVERAGE PLAT;AR L INE AR HEAT GENERATION RATE............ 3/42-1 A P RM SE T P 0 I t:T S........................................ 3/42-9 s dpg MI N IlfJM CR ITI C AL POWER R AT I O....................... 3 2 i ...,.. 2 3 2 0 4 6.... 3/4 2-12 i LINEAR HEAT GENERATION RATE... BRUNSWICK - UNIT 2 Iy Amendment No. 48

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.............. 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................... 3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION. 3/4 3 3/4.3.4 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION...........3/4 3-39 3/4.3.5 MONITORING INSTRUMENTATION Sei smic Moni toring Ins crumentation..................... 3/4 3-44 Remote Shutdown Monitoring Instrumentation............. 3/4 3-47 Post-accident Moni toring Instrumentation............... 3/4 3-50 Sou rc e Ra ng e Mo ni tors.................................. 3 /4 3-53 Chl ori ne De tecti on System.............................. 3/4 3-54 Chl oride Intrusion Moni tors............................ 3/4 3-55 Fire Detection Instrumentation......................... 3/4 3-59 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION.... 3/4 3-62 l 3/4.3.6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.................................... 3/4 4-1 Jet Pumps.............................................. 3/4 4-2 Idle Recirculction Loop Startup........................ 3/4 4-3 3/4.4.2 SAFETY / RELIEF VALVES................................... 3/4 4-4 s,'., 2232 047 .. ~ BRUNSWICK - UNIT 2 V Amendment No. 48 + y?I p-

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................ 3/4 4-5 Operational Leakage.................................. 3/4 4-6 3/4.4.4 C H E M I ST RY............................................ 3/4 4-7 3/4.4.5 SPECIFIC ACTIVITY.................................... 3/4 4-10 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System............ 3/4 4-13 Reactor Steam Dome................................... 3/4 4-18 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES..................... 3/4 4-19 3/4.4.8 STRUCTURAL INTEGRITY................................. 3/4 4-20 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM............... 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM.................... 3/4 5-3 3/4.5.3 LOW PRESSURE COOLING SYSTEMS 3/4 S-4 Core Spray System.................................... Low Pressure Coolant Injection System................ 3/4 5-7 3/4.5.4 SUPPRESSION P00L..................................... 3/4 5-9 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integri ty........................ 3/4 6-1

Primary Containment Leakage..........................

3/4 6-2 yhg Primary Containment Air Loc k......................... 3/4 6-4 BRUNSWICK - UNIT 2 VI 2232 04A

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.9 REFUELINGOPERATIONS(Continued) 3/4.9.2 INSTRUMENTATION...................................... 3/49-3 3/4:9.3 CONTROL R0D P0SITION................................. 3/4 9-5 3/4.9.4 DECAY TIME........................................... 3/4 9-6 3/4.9.5 COMMUNICATIONS....................................... 3/4 9-7 3/4.9.6 CRANE AND HOIST OPERABILITY.......................... 3/4 9-8 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE P00L................. 3/4 9-9 3/4.9.8 WATER LEVEL-REACTOR VESSEL........................... 3/4 9-10 3/4.9.9 WATER LEVEL-SPENT FUEL STORAGE P00L.................. 3/4 9-11 3/4.9.10 CONTROL R0D REMOVAL Single Control Rod Removal........................... 3/4 9-12 Mul tipl e Control Rod Removal......................... 3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY....................... 3/4 10-1 3/4,,10.2 ROD SEQUENCE CONTROL SYSTEM.......................... 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS....................... 3/4 10-3 3/4.10.4 RECIRCULATION L00PS..........................,....... 3/4 10-4 2232 049

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) ( L, t. U NIT'2 IX BRUNSWICK U W-

INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY.......................................... B 3/4 0 l 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN................................ 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 CONTROL R0D 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 AVERf.GE PLANAR LINEAR HEAT GENERATING RATE..... B 3/4 2-1 3/4.2.2 APRM SETP0lNTS................................. B 3/4 2-3 3/4.2.3 MINIMUM CRITICAL POWER RATI0................... B 3/4 2-3 3/4.2.4 LINEAR HEAT GENEPATION RATE.................... E 3/4 2-5 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 R0D WITHDRAWAL BLOCK INSTRUMENTATION... B 3/4 3-2 3/4.3.5 MONITORING INSTRUMENTATION..................... B 3/4 3-2 i3/443.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION.............................. B 3/4 3-4 3/4.4 REACTOR COOLANT SYSTEM 2232 050 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 2 X Amendment No. 48

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or low Flow) 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 800 psia or core flow less than 10% of rated flow. APPLICABILITY: CONDITIONS 1 and 2. ACTION: With THERMAL POWER excet-ding 25% of RATED THERhAL POWER and the reactor vessel steam dome pressure less than 800 psia or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours. THERMAL POWER (High Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with the reactor vessel steam dome pressure greater than 800 psia l and core flow greater than 10% of rated flow. APPLICABILIT Y : CONDITIONS 1 and 2. ACTION: With MCPR less than 1.07 and tjie reactor vessel steam dome pressure l greater than 800 psia and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours. REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig. APPLI'CABILITY : CONDITIONS 1, 2, 3 and 4. ACTION: With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure 1 1325 psig within 2 hours. 2232 051 BRUNSWICK - UNIT 2 2 -1 Amendment No. 48

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS-(Continued) REACTOR VESSEL WATER LEVEL 2.1. 4 The reactor vessel water level shall be above the top of the active irradiated fuel. APPLICABILITY: CONDITIONS 3, 4 and 5 ACTION: With the reactor water level at or below the top of the active irra-diated fuel, manually initiate the low pressure ECCS to restore the reac+.or vessel water level, af ter der essurizing the reactor vessel, if required. 2232 052 b 4 BRUNSWICK - UNIT 2 2-2

2.1 SAFETY LIMITS BCES 2.0 The fuel cladding, reactor pressure vessel and primary sys. piping are the principr.i barriers to the release of radioactive mat-Safety limits are established to protect the erials to the environs. integrity of these barriers during normal plant operations and antici-The fuel cladding integrity limit is set such that no pated transients. Be-fuel damage is calculated to occur if the limit is not violated. cause fuel dPage is not directly observeble, a step-back approach is used to establish a Safety Limit such that the MINIMUM CRITICAL POWER RATIO (MCPR) is no less than 1.07 MCPR > 1.07 represents a consern-l tive margin relative to the conditions required to maintain fuel clLding The fuel cladding is one of the physical barriers which integrity. separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perfor-ations 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 ree:ter operation significantly above design conditions and the Limiting Safety System Settings. While fission product migra-tion from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold, bey,ond 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 These con-would produce onset of transition boiling, MCPR of 10. ditions 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 GEXL :orrelation is not vs. lid 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 This is done by establishing a limiting condition on core by other means. THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at lor power and fh ws will always be greater than 4.5 psi. Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure 3 drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 3 lbs/hr. Full scale AfLAS test data taken at presseres from 28 x 10 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is a;, proximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL Thus, a THERMAL POWER limit of 25". of RATED THERMAL POWER for POWER. reactor pressure below 800 psia is conservative, 2232 053 s knendment No. 48 BRUNSWICK - DNIT 2 B 2-1

SAFETY LIMITS BASES (Continued) 2.1.2 THERMAL POWER (High Pressure and Hioh 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 dur-ing reactor operat',.: the thermal and hydraulic conditions resultir.g in a departure from i.ucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Altnough it is recognized that a depart: ire 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 integ-rity 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 the General Electric Thermal Analysis Basis, GETAB,1 which is a statistical model that combines all of the uncertainties in operating parameters and the pro-cedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L), (GEXL), correlation. The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation. These conditions are: Pressure: 800 to 1400 psia 6 Mass Flux: 0.1 to 1.25 10 lb/hr-ft Inlet Subcooling: 0 to 100 Btu /lb Local Peaking: 1.61 at a corner rod to 1.47 at an interior rod Reference 1 " General Electric BWR Thermal Analysis Basis (GETAB) Data, Correla-y ) - (tion and Design application," NED0-10958 and NED0-10958. 2232 054 BRUNSWICK - UNIT 2 B 2-2

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 set-point 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 rcmain 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 are 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 subtritical 'and the IRM's 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 shutdown and peak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above 1.07. Based on this analysis, the l IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protaction for the APRM. 2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram set;ing of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. This margin accom-modates 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 2232 055 ? A:;1 , ff) Amendment No.48 ' BfkUNSWICK (UNIT 2 8 2-g

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) 2. Average Power Range Monitor (Continued) the possible sources of reactivity input, uniform control rod withdrawal Because the is the most probable cause of significant power increase. 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, la an assumed uniform rod 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 placeo in the Run oosition. 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 thdrmal 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 operationa! transients analyzed violates the fuel safety limit and there is substan- .tial 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 the Safety Limits and yet allows operating margin that reduces the possi-bility of unnecessary shutdowns. The flow referenced trip setpoint must be adjusted by the specified formula in c;h to maintain these margins, 3. Reactor Vessel Steam Dome Pressure-High High Presnre in the nuclear systcm could cause a rupture to the nuclear system process barrier resulting in the release of fission A pressure increase while operating, will also tend to increase

products, 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 The setting provides for a wide 'nargin to the maximum d.spurioustrips. allow 5ble design pressure and takes into account the locatio i 2232 056 3RUNSWICK - UNIT 2 B 2-10

REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION Both Rod Block Monitor (RBM) channels shall be OPERABLE. 3.1.4. 3 APPLICABILITY : CONDITION 1, when THERMAL POWER is greater than the preset power level of the RWM and RSCS ACTION : With one RBM channel inoperable, POWER OPERATION may continue a. provided that either: 1, The inoperable RBM channel is restored to OPERABLE status within 24 hours, or 2. The redundant RBM is demonstrated OPERABLE within 4 hours ~ and at least once per 24 hours 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 1.07. assuming a single error that results. in complete. l withdrawal of.any single control rod that is capable of Withdrawal. Otherwise, trip at least one rod block monitor channel. With both RBM channels inoperable, trip at least one rod block b. monitor channel within one hour. SURVEILLANCE REQUIREMENTS Each of the above required RBM channels shall be demonstrated 4.1. 4. 3 OPERABLE by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALI-BRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Tab'le 4 !3.4-1. 2232 057 ro: Bk NSllICK - UNIT 2 3/4 1-17 Amendment No. 4:

REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE with: An OPERABLE flow path from the storage. tank to the reactor a. core, containing two pumps and and two inline explosive injection valves, b. The contained solution volume-concentration within the limits of Figure 3.1.5-1, and The solution temperature above the limit of Figure 3.1.5-2. c. APPLICABILITY: CONDITIONS 1, 2, and 5. ACTION: a. In CONDITION 1 or 2: 1. With one pump and/or one explosive valve inoperable, restore the inoperable pump and/or explosive valve to OPERABLE. status within 7 days or be in at least HOT SHUTOOWN within the next 12 hours. 2. With the standby liquid control system inoperable, restore the system to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours. 6. In CONDITION 5: 1. With one pump and/or one explosive valve inoperable, restare the inoperable pump and/or explosive valve to OPERABLE status within 31 days or suspend all operations involving CORE ALTERATIONS or positive reactivity changes. 2. With the standby liquid control system inoperable, sus-pend all operations involving CORE ALTERATIONS or posi-tive reactivity changes and fully insert all insertable control rods within one hour. 3. The provisions of Specification 3.0.3 are not applicable. e'LtCI 2232 058 BRUNSWICK - UNIT 2 3/4 1-18 Amendment No. 46

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGR's) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1 5, 3. 2.1 -6 or 3,2.1 -7. APPLIC ABILITY _: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER. ACTION. With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6 or 3.2.1-7, initiate corrective action within l 15 minutes and continue corrective action so that APLHGR is within the limit within 4 hours or reduce THERMAL POWER to less than 25' of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGR's shall be verified to be equal to.or less than the applicable limit determined from Figure 3.2.1-1,'3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1 -5, 3.2.1-6 or 3.2.1 -7 : a. At least once per 24 hours, b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and Initially and at least once per 12 hours when the reactor is c. operating with a LIMITING CONTROL R0D PATTERN for AFLHGR 2232 059 Ud i

RUilSWICK - UNIT 2 3/42-1 Amendment No, 48 n.

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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 foilowing relationships: S 1 (0.66W + 54%) T S 1 (0.66W + 42%) T RB where: S and S are in percent of RATED THERMAL POWER, DB W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T 1 1.0),and Design TPF for: 8 x 8R fuel = 2.48 7x7 fuel = 2.60 8x8 fuel = 2.45 APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER. ~ ACTION: With S or S exceeding the allowable value, initiate ~ corrective ac' ion within 15 m.qtlutes and continue corrective action so that S and S are I RB within the required limits within 4 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.2 The MTPF for each class of fuel shall be determined, the value of T calculated, and the flow biased APRM trip setpoint adjusted, as required: 2232 067 a At least once per 24 hours, ni. 00i.: b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and Initially and at least once per 12 hours when the reactor is c. operating with a LIMITING CONTROL R0D PATTERN for MTPF. BRUNSWICK , UNIT 2 3/4 2-9 Amendment No. 48

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core 3.2.3 shown in fIrw. shall t,e equal to or greater than MCPR times the Kf Figure 3.2.3-1, for Beginning-of-cycle (B0C) to end-of-cycle (EOC) minus a. 2000 MWD /t, with: 1. MCPR for 7x7 fuel = 1.20, 2. MCPR for 8x8 fuel e 1.21, 3. MCPR for 8x8R fuel = 1.25. b. EOC minus 2000 MWD /t to E0C, with: 1. MCPR for 7x7 fuel = 1.21, 2. MCPR for 8x8 and 8x8R fuel = 1.27. APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% RATED TMERMAL POWER ACTION _: With MCPR less than the applicable limit determined from Fio ue 3.2.3-1, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable limit within 4 hours or reduce THERMAL POWER to less than 25% of RATED THERMA within the next 4 hours. SURVEILLANCE REQUIREMENTS ' dO \\ c ' E 4.2.3" HCPR'shall be determined to be equal to or greater than the applicable limit determined from Figure 3.2.3-1: 2232 068 At least once per 24 hours, a. Whenever THERMAL POWER has been increased by at least 15% b. of RATED THERMAL POWER and steady state operating c onditiore have been established, and Initially and at least once per 12 hours when the reactor is c. operating with a LIMITING CONTROL R0D PATTERN for MCPR. I Amendment No. 48 BRUNSWICK - UNIT 2 3/4 2-10

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POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 All LINEAR HEAT GENERATION RATES (LHGR's) shall not exceed: For 7 X 7 fuel assemblies, as a function of core height for a. any fuel rod in an assembly, the maximum allowable LHGR shown in Figure 3.2.4-1. b. For 8 X 8 and 8 X 8R fuel assemblies,13.4 kw/ft. APPLICABILITY : CONDITION 1. when THERMAL POWER > 25% of RATED THERMAL POWER. ACTION: With the LHGR of any fuel rod exceeding the above limits, initiate corrective action within 15 minutes and continue corrective action so that the LHGR is within the limit within 4 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours.. SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than the appli-cable above limit: a. At least once per 24 hours, b. When THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and Initially and at least once per 12 ' hours when the reactor is c. operating on a LIMITING CONTROL R0D PATTERN for LHGR. (Ig{ pq t 223:2 070 BRUNSWICK - UNIT 2 3/4 2-12 Amendment No. 4/,48

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LHGR (kw/ft) LIMITIt4G VALUE FOR LHGR 2232 071 D D D @ b 7 x 7 FUEL mo n wS ol Hb n FIGURE 3.2.4-1 Amendment fio.,27,48 BRUf;5 WICK - UtilT 2 , 3/4 2-13

INSTRUMENTATION 3/4.3.4 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4 The control rod withdrawal block instrumentation shown i:. Table 3.3.4-1 shall be OPERABLE with their trip setpoints set ccW-tent with the values shown in the Trip Setpoint column of Table 3.3.4. APPLICABILITY: As shown in Table 3.3.4-1. ACTION: a. With a control rod withdrawal block instrumentation channel tr setpoint less conservative than the value shown in the Allowable. Values column of Table 3.3.4-2, declare the channel incperable 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 hours, or 2. The redundant trip system is demonstrated OPERABLE within 4 hours and at least once per 24 hours 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 1.07 assuming a single l error that results in complete withdrawal of any single control rod that is capable of withdrawal. 4. Otherwise, place at least one trip system in the tripped condition within the next hour. With the requirements for the minimum number of OPERABLE c. channels not satisfied for both trip systems, place at least one trip s/ stem in the tripped condition within one hour, d. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5. SURVEILLANCE RE0VIREMENTS 4.3.4 Each of the above required control rod withdrawal block instrumen-tation; channels shall be demonstrated OPERABLE by t:ie 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. BRUNSWICK-UNIT 2 3/4 3-39 Amendment No, 48

v, TABLE 3.3.4-1 /o CONTROL' ROD WITHDRAWAL BLOCK INSTRUMENTATION E .n h -) MINIMUM NUMBER OF AFPLICABLE E OPERABLECHANNEL{,) OPERATIONAL TRIP FUNCTION AND INSTRUMENT NUMBER PER TRIP SYSTEM CONDITIONS g k 1. APRM (C51-APRM-CH.A,B.C,0,E,F) N a. Upscale (Flow Biased) 2 1 b. Inoperative 2 1,2,5 c. Downscale 2 1 d. Upscale (Fixed) 2 2, 5 2. R0D BLOCK MONITOR (CSI-RBM-CH.A,B) a. Upscale 1 1* b. Inoperative 1 1* c. Downscale 1 1* 3. SOURCE RANGE MONITORS (C51-SRM-K600A,B,C,D) Detectogotfullin(b) 1 2, 5 a. 1 2, 5 b. Upscale ) 1

2. 5 Inoperati Downscale{g) c.

1 2, 5 d. g INTERMEDIATE RANGE MONITORS (d) (C51-IRM-K601A,B,C,D,E,F,G,H) u 4. N Detector not full in(e) 2 2, 5 a. O b. Upscale .2 2, 5 c. Inoperabl 2 2, 5 W d. Downscale *) 2 2 i I

TABLE 3.3.4-1 (Continued) CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION NOTE When THERMAL POWER exceeds the preset power level of the RM4 and RSCS. a. The minimum number of OPERABLE CHANNELS may be reduced by one for up to 2 hours in one of the trip systems for maintenance and/or testing except for Rod Block Monitor function. b. This function is bypassed if detector is reading > 100 cps or the IRM channels are on range 3 or higher. c. This function is bypassed when the associated Imi channels are on range 8 or higher. d. A total of 6 IRM instruments must be OPERABLE. e, This function is bypassed when the IRM channels are on range 1. 1 2232 074 6 t. .\\ [. BRUNSillCK-UNIT 2 3/4 3-41

TABLE 3.3.4-2 CONTROL R0D WITH0RAWAL BLOCK INSTRUMENTATION SETPOINTS m v, 5 TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE p Q ./ f. 1. APRM (CSI-APRM-CH.A,B,C,D,E,F) c = a. Upscale (Flow Biased) ,< (0.66 W + 42%) T* < (0.66 W + 42%) T* U b. Inoperati'e 'NA MTPF NA MIPF ~ c. Downscale > 3/125 of full scale > 3/125 of full scale d. Upscale (Fixed) ' 7 12% of RATED THERMAL POWER 7 12% of RATED THERMAL POWER 2. ROD BLOCK MONITOR (C51-RE:1-CH. A,B) < (0.66 W + 39%) RTPT l a. Upscale < (0.66W + 39 ) T* T* TTA b. Inoperative NA MTPF c. Downscale > 3/125 of full scale > 3/125 of full scale R 3. SOURCE RANr3 "C:41 TORS (C51-SRM-K600A,B,C,D) a w a. Detector not full in NA NA 5 5 b b. Upscale < 1 x 10 cps < 1 x 10 cps c. Inoperative NA ITA d. Downscale > 3 cps > 3 cps N N 4. INTERMEDIATERANGEMONITORS(C51-IRM-K601A,B,C,D,E,F,G,H) u a. Detector not full in NA NA N ya b. Upscale < 108/125 of full scale < 108/125 of full scale w c. Inoperative NA NA g & U1 d. Downscale > 3/125 of full scale > 3/125 of full scale c, = , T=2.60 for 7 x 7 fuel. T=2.45 for 8 x 8 fuel. T=2.48 for 8 x 8R fuel.

TABLE 4.3.4-1 + CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS N CHANNEL OPERATIONAL ^, CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH y -TRIP FUNCTION AND INSTRUMENT NUldBER CHECK TEST CALIBRATION (3) SURVEILLANCE REQUIRED 7 l. APRM (C51-APRM-CH.A,B,C,0,E F) a. Upscale (Flow Biased) NA S/U(c),M R(b) ) [ b. Inoperative NA S/U ,Q NA 1, 2, 5 c. Downscale NA M NA 1 S/U(c),Q S/U R 2, 5 l d. Upscale (Fixed) NA 2. ROD BLOCK MONITOR (C51-RBM-Cil.A,B) S/U((c) M a. Upscale NA R 1* S/U c),Q NA 1* b. Inoperative NA c. Downscale NA S/U( ),,M R 1* 3. SOURCE RANGE MONITORS (C51-SRM-K600A,B.C,D) a. Detector not full in NA S/U ,W NA 2, 5 mg b. Upscale NA S/U ,W NA 2, 5 c. Inoperative NA S/U(c,W NA 2, 5 S/U ,W NA 2, 5 g d. Downscale NA 4. INTERf1EDIATE RANGE MONITORS (C51-IRM-K601A,B,C,D,E,F,G,H) a. Detector not full in NA S/U(c) 9(d) NA 2 NA W NA 5 S/U(c) g(d) b. Upscale NA NA 2 m NA W NA 5 N S/U(c) g(d) U c. Inoperative NA NA 2 kN NA W NA 5 c. Downscale NA S/U(c) g(d) NA 2 h NA H NA 5 g 2 a. CHANNEL CALIBRATIONS are electronic. b. This calibration shall consist of the adjustment of the APRM flow biased setpoint to conform to a calibrated flow signal. w?' c. Within 24 hours prior to startup, if not performed within the previous 7 days. d. When changing from CONDITION 1 to CONDITION 2, perform the required surveill within 12 hours after entering CONDITION 2. When THERMAL POWER is greater than the preset power level of the R1"' RSCS.

INSTRUMENTATION 3/4.3.5 MONITORING INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION The seismic monitoring instrumentation shown in Table 3.3.5.1-1 3.3.5.1 shall be OPERABLE. APPLICABILITY: At all times. ACTION: With one or more seismic monitoring instruments inoperable for a. more than 31 days, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission, within the next 14 days outlining the cause of the malfunction and the plans for restoring the instruments to OPERABLE status, b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS Each of the above required seismic monitoring instruments shall 4.3.5.1.1 be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.5.1-1. Each of the above required seismic monitoring instruments actu-4.3.5.1.2 ated during a seismic event shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 5 days following the 9ata shall be retrieved from actuated instruments and seismic event. In analyzed to determine the magnitude of the vibratory ground motion. lieu of any other report required by Specificatior. 6.9.1, a Special Report shall be prepared and submitted to the Cumission pursuant to Specification 6.9.2 within 14 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety. 2232 077 M 09 ti, BRUNSWICK-UNIT 2 3/4 .44 a

TABLE 3.3.5.7-1 (Continued) MINIMUM INSTRUMENTS OPERABLE INSTRUMENT LOCATION FLAME HEAT SM0KE 4. Service Water Building Zone 1 4' 0 0 6 Zone 2 20 0 0 5 5. A0G Building Zone 1 20' 1 0 0 Zone 2 20' 1 0 0 Zone 3 20' 1 5 1 Zone 4 37' - 49' 1 6 0 2232 078 5 ? \\ iN ' Y.,. w BRUNSWICK - UNIT 2 3/4 3-61 Amendment No. 47

I INSTRUMENTATION 3/4.3.6 AT., RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.1 The Anticipated Transient Without Scram recirculation pump trip (ATWS-RPT) system instrumentation trip systems shown in Table 3.3.6.1-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6.1-2. APPLICABILITY: CONDITION 1. ACTION: With an ATWS recirculation pump trip system instrumentation a. trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6.1-2, declare the trip system inoperable until the trip system is restored to f,'OPERABLEistatus with its trip setpoint adjusted consistent " with -the-Trip Setpoint value, b. With the requirements for the minimum number of OPERABLE trip systems per operating pump not satisfied for one Trip Function, restore the inoperable trip system to OPERABLE status within 14 days or be in at least STARTUP within the next 8 hours. SURVEILLANCE REQUIREMENTS 4.3.6.1.1 E.ch ATWS recirculation pump trip system instrumentation trip system shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNLu FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.6.1.1-1. 4.3.6.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic opera-tion of all channels chall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system. 2232 079 48 BRUNSWICK - UNIT 2 3/4 3-62 Amendment No.

TABLE 3.3.6.1-1 w E ATWS RECIRCULATION PUMP TP.IP SYSTEM INSTRUMENTATION 55 C MINIMUM NUMBER OPERABLE TRIP TRIP FUNCTION AND INSTRUMENT NUMBER SYSTEMS PER OPERATING PUMP h, ~,' I-1. Reactor Vessel Water Level - 1 Low Low, level 2 N-(B21-LIS-N024 A, B; B21-LIS-N025 A, B) 2. Reactor Vessel Pressure-Low I (B21-PS-N045 A, B, C, D) R u Y S 2T a R N N 5 W N Oco O

~2 TABLE 3.3.6.1-2 m CD

n h

C ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS 5 ./ p ~ e TRIP ALLOWABLE TRIP FUNCTION AND INSTRUMENT NUMBER SETPOINT VALUE c 2 d 1. Reactor Vessel, Water Level - > -38 inches > -38 inches ~ Low low, level 2 (B21-LIS-N024 A, B; B21-LIS-N025 A, B)

7..

Reactor Vessel Pressure-Low '->1120 psig ->l120 psig (B21-PS-N045 A, B, C, D) t ? N N U N F o* O n 5

TABLE 4.3.6.1-1 m ATWS RECIRCULATION PUMP TRIP SYSlEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS s 9 CHANNEL CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION AND INSTRUMENT NUMBER CHECK TEST CALIBRATION c_5 H 1. Reactor Vessel Water Level - S M R Low Low, Level 2 (B21-LIS-N024 A, B; 321-LIS-N025 A, B) 2. Reactor Vessel Pressure - Low NA M R (B21-PS-N045 A, B, C, D) m V' E N n N E LN 2 N =" CD 5 CD N

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration cf SHUTDOWN MARGIN will be performed in the cold xenon-free condition and shall show the core to be subcritical by at least R + 0.38% AK. The value of R in units of %AK is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R must be positive or zero and must be determined for each fuel loading cycle. Satisfaction of this limitation can be best demonstrated at the time of fuel loading but the margin must be determined any time a control rod is incapable of insertion. This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determinvJ anytime a control red i,s incapable of insertion. 3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns. Since the comparisons are easily done, frequent checks are not an imposition on nomal operations. A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated. A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients. 3/4.1.3 CONTROL RODS 2232 083 The specifications of this section ensure that 1) the minimum SHUTDOWN MARGIN is maintained, 2) the control rod insertion times are i consistent with those used in the accident analysis, and 3) the BRUNSWICK - V'!IT 2 B 3/4 1-1

\\ REACTIVITY C0ffTROL SYSTEMS BASES CONTROL RODS (Continued) potential effects of the rod ejection accident are limited. The ACTION statements pennit variations from the basic :cquirements but at the same time impose more restrictive criteria for continued operation. A limita-tion on inoperable rod' is set such that the resultant effect on total rod worth and scram shape will be kept to a minicum. lhe requirements for the various scram time measurements ensure that any indication of

ystematic problems with rod drives will be investigated on a timely b~a s i s.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod innovable becpse of excessive f riction 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 larga numb'er of inoperable control rods. Control rods that are inoperable for othe.r reasons are permitted to be taken out of service provided that those in the non-fully-inserted position are consistent wi+h: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 occurrance of eight inoperable rods could be indicative of a generic problem and' the reactor must be shutdown for investigation and resolution of the probl em. The control rod system is analyzed to bring the reactor subcritical at a rate f ast enough to-prevent the MCPR from becoming less than 1.07 during the limiting power transient analyzed in Section 14.3 of the FSAR, This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MPCR remains greater than 1.07 The occurrence of scram times longer than those specified ( should b'e viewed as aa indication of a systemic problem with the rod drives and therefere the surveillance interval is reduced in order to prevent operation of the reactor for long periocs of time with a potentially serious problem. Control rods with inoperable accumulators are declared inoperable hddbpecification 3.1.3.1 then applies. This prevents c pattern of inoperable accumulators that would result in less reactivity insertion 2232 OM BRUNSWICK - UNIT 2 B 3/4 1-2 Amendment No. 48

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature 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 June 1971 considering the postulated effects of fuel pellet densification. 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding tempeature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K. The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat genera-tion rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within a assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap The Technical Speci-conductance and rod-to-rod local peaking factor. f,1 cation APHGR is this LHGR of the highest powered rod divided by its The limiting value for APLHGR is shown in local peaking factor. Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6 and 3. 2.1 -7. The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6 and 3.2.1-7 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Differences in this analysis compared to previous analyses Reference 1. performed with Reference 1 are: (1) The analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6 and 3.2.1-7 ; (2) Fission l product decay is computed assuming an energy release rate of 200 MEV/ Fissio (3) Pool boiling is assumed af ter nucleate boiling is lost during the flow stagnation period; (4) The effects of core spray entrainraent and counter-current flow limitation as described in Reference 2, are included in the reflooding calculations. 223{085 .) ' A list of the significant plant input parameters to t e Foss-of-2 ll' coolant accident analysis is presented in Bases Table B 3.2.1-1. BRUNSWICK - UNIT 2 B 3/4 2-1 /cendment No. 48 s...

Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE l LOSS-0F-COOLANT ACCIDENT ANALYSIS FOR BRUNSWICK - UNIT 2 Plant Parameters _; Core Thermal Power 2531 Mst which corresponds to 105% of rated steam flow l 6 Vessel Steam Output 10.96 x 10 lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure 1055 psia Recirculation Line Break Area for large Breaks a. Discharge 2.4 ft2 (DBA); 1.9 ft2 (80% DBA) b. Suction 4.2 ft Number of Drilled Bundles 520 Fuel Parameters: PEAK TECHNICAL INITIAL ~ SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING. POWER,, . FUEL TYPES GE0 METRY (kw/ft) FACTOR RATIO l Reload Core 8x8 13.4 1.4 1.20 7x7 18.5 1.5 1.20 A more detailed list of input to each model and its source is presented in Section II of Reference 1. This power level meets the Appendix K requirement of 102%. To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e. 1.2 divided by 1.02) for a bundle with an initial (?f.10 M 51sPR of 1.20. 2232 086 BRUNSWICK - UNIT 2 B 3/4 2-2 Amendment No. 48

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity safety limits of Specification 2.1 were based on a TOTAL PEAKING FACTOR of 2.60 for 7 x 7 fuel, 2,45 for 8 x 8 fuel and 2.48 for 8 x BR fuel. The scram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not beco1,u less than 1.0 in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the fomula in this specification when the combination of THERMAL POWER and peak flux indicates a TOTAL PEAKING FACTOR greater than 2.60 for 7 x 7 fuel, 2,45 for 8 x 8 fuel and 2,48 for 8 x SR fuel, The method used to determine the design TPF shall be consistent with the method used to determine the MTPF. 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 the established fuel cladding integrity Safety Lgt MCPR of 1.07, and an analysis cf For any abnormal operating tran. abnormal operational transients sient 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 below the Safety Limit MCPR at any time during the transient assuming instrtment 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 detemine which result in the largest reduction in CRITICAL POWER RATIO (CPR), The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient which detemines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass. This transient yields the largest a MCPR. When added to the Safety Limit MCPR of 1.07 the required minimum operating limit MCPR of Specification l 3.2.3 is obtained. Prior to the analysis of abnomal operational tran-sients an initial fuel bundle MCPR was determined. This parameter is based on the bundle flow calculated by a GE multi-channel steady gteand flow distribution model as described in Section 4.4 of NED0-20360 on core parameters shown in Reference 3, response to items 2 and 9. 2232 087 o 0 (BRUNSWICK - UNIT 2 B 3/4 2-3 Amendment No. 48 s

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) The evaluation of a given transient begins with the system initial parameters shown in Attachment 5 of Reference 6 that are input to a GE-core dynamic behavior transient computer program described in NED0-10802(5). Also, the void reactivity coefficients that were input to the transient calculational procedure are based on a new method of calcula-tion termed NEV which provides a better agreement between the calculated and plant instrument power distributions. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic SCAT code described in hED0-20566(1). The principal result of this evaluation is the reduction in MCPR caused by the transient. factor is to define operating limits at' other than The purpose of the Kf rated flow conditions. At less than 100% flow the required MCPR is the product of the operating limit MCPR and the K, factor. Specifically, the K factor provides the required thermal margin to protect against a flow increase transient. The most limiting transient initiated from f less than rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure. For operation in the automatic flow control mode, the K, factors assure that the operating limit MCPR of Specification 3.2.3 will not be vio-lated should the most limiting transient occur at less than rated flow. factors assure that the Safety In the manual flow control mode, the Kf Limit MCPR will not be violated should the most limiting transient occur at less than rated flow. The K factor values shown in Figure 3.2.3-1 were developed generically f The K which are applicable to all BWR/2, BWR/3, and BWR/4 reactors thermal power at rated core flow. factors were calculated such For +.he manual flow control mode, the Kf that the maximum flow state (as limited by the pump scoop tube set point) and the corresponding core power (along the rated flow control line), the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different core flows. The ratio of the I g o,MCPR calculated at a given point of core flow, divided by the opera limit MCPR determines the K. f 2232 088 BRUNSWICK - UNIT 2 8 3/4 2-4

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) For operation in the automatic ficw control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at rated power and flow. The K factors shown in Figure 3.2.3-1 are conservative for the General Electric Plant operation with 8 x 8 and 8 x 8R fuel assemblies because l f the operating limit MCPR's of Specification 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of The k curves are conservative for 7 x 7 fuel whenever the operating K it MCPR is greater than 1.23 as documented in Appendix C of NEDE I i fm. A correction to the K curves is, therefore, necessary when-240ll-P-A. r ever the MCPR for the 7 x 7 fuel is equal to or less than 1.23 in order to This ensure that the fuel cladding integrity Safety Limit is not violated. correction is made by using a scoop tube set point of 102.5%. The MCPR for 7 x 7 fuel is then the product of the value given in Specification 3.2.3 and the K curve based on 112% as shown in Figure 3.2.3-1. Whenever the MCPR for the 7 x 7 fuel is greater than 1.23, this correction.is not g applied. At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void For all designated control rod patterns content will be very small. which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considera-ble margin. Jith this low void content, any inadvertent core flow increase woul'd' only' place operation in a more conservative mode relative During initial start-up testing of the plant, a MCPR evaluation to MCPR. will be made at 25% thermal power level with mimimum recirculation pump The MCPR margin will thus be demonstrated such that future MCPR speed. The evaluation below this power level will be shown to be unnecessary. daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have The requirement for not been significant power or control rod changes. 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. 3.2.4 LINEAR HEAT GENERATION RATE The LHGR specification assures that the linear heat generation rate in any rod is less than the design linear heat generation even if fuel The power spike penalty specified pellet densification is postulated.islased on the analysis present report NEDM-10735 Supplement 6, and assumes a linearly increasing variation in axial gaps between core bottom and top, and assures with a 95% confidence that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking. BRUNSWICK - ; NIT 2 B 3/4 2-5 Amendment No. 48

POWER DISTRIBUTION LIMITS BASES General Electric Company Analytical Model for less-of-Coolant 1. Analysis in Accordance with 10 CFR 50, Appendix K, NE00-20566, January,1976. General Electric Refill Reflood Calculation (Sapplement to SAFE 2. Code Description) transmitted to USAEC by letter, G. L, Gyorey to V. Stello, Jr., dated December 20, 1974. Letter from J. A. Jones, Carolina Power and Light Company to 3. B. C. Rusche, NRC transmitting Amendment 31 to the Bninswick Uni'.1 Docket No. 50-325, dated November 26, 1975. General Electric BWR Generic Reload Application for 8 x 8 Fuel, 4. NEDO-20360, Revision 1, November 1974. S. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NED010802). 6. Letter from J. A. Jones, Carolina Power and Light Company, to B. C. Rusche, NRC dated May 7,1976. 2232 090' P O S ?. (, s BRUNSWICK - UNIT 2 B 3/4 2 -6 Amendment No. 48

INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Cont mued) 3/4.3.5.2 REMOTE SHUTDOWN MONI~0 RING INSTRUMENTATION The OPERABILITY of the remote shutdown monitoring instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of CFR 50, 3/4.3.5.3 POST-ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the post-accident monitoring instrumentation ensures that sufficient information is available on selected plant pa-rameters to monitor and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97 " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975. 3/4.3.5.4 SOURCE RANGE MONITORS The source range monitors provide the operator with information on the status of the neutron level in the core at very low power levels At these power levels reactivity additions should not during startup. be made without this flux level information available to the operator. When the intermediate range monitors are on scale adequate information is available without the SRM's and they can be retracted. 3/4.3.5.5 CHLORINE DETECTION SYSTEM The OPERABILITY of the chlorine detection systems ensures that an accidental chlorine release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room personnel. Upon detection of a high concentration of chlorine the control room emergency ventilation system will automatically isolat'e'the control room and initiate operation in the recirculation The detection systems required mode to provide the required protection. by this specification are consistent with the recommendations of Regulatory Guide 1.95 " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release." 2232 091 BRUNSWICK - U, NIT 2 B 3/4 3-3

INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Contir.ued) 3/4.3.5.6 CHLORIDE INTRUSION MONITORS The chloride intrusion monitors provide adequate warning of any leakage in the condenser or hotwell so that actions can be taken to mitigate the consequences of such intrusion in the reactor coolant system. With only a minimum number of instruments available increased sampling frequency provides adequate information for the same purpose. 3/4.3.5.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fir s in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection instrumentation is inoperable, incr3asing the frequency of fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION The ATWS recirculation pump trip system has been added at the suggestion of ACRS as a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within an envelope of study events given in General Electric Company Topical Report NECO-10349, dated March,1971. 2232 092 M' BRUNSWICK - UNIT 2 B 3/4 3-4 Amendmen No. 48

5.0 DESIGN FEATURES I 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1, based on the information given in Section 2.2 of the FSAR. 5.2 CONTAINMENT CONFIGURATION-5.2.1 The PRIMARY CONTAINMENT is a steel lined reinforced concrete structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a con-crete steel lined pressure vessel in the shape of a torus. The primary containment has a minimum free air volume of (288,000) cubic feet. DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained fer: a. Maximum internal pressure 62 psig. b. Maximum internal temperature: drywell 300 F. suppression chamber 200*F. c. Maximum external pressure 2 psig. 5.3 REACTOR CORE., 2232 093 FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 560 fuel assemblies with each 7 x 7 fuel assembly containing 49 fuel rods, each 8 x 8 fuel assembly containing 63 fuel rods; and each 8 x SR fuel assembly containing 62 fuel rods. All fuel rods shall be clad with Zircaloy 2. The nominal active fuel length of each fuel rod shall be 144 inches for 7 x 7 fuel assemblies, 146 inches for 8 x 8 fuel assemblies, and 150 inches for 8 x 8R fuel assemblies. Each fuel rod shall contain a maximum total weight of 4430 grams of UO " 2 BRUNSWICK-UtbT2 5-1 ', Amendment No. 48

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~ ' ~'s!b J' n \\. r V. ' "is, o :ntut c:.s::.:. \\ ! y, \\ G. q# ,tfu ( M+/.' 9 M/ wr. s ~, / ,s - /[- ~ / j#" \\ N /' / Y ,/ ,s ~~ /s /ws ! :l tI / au, I i;: -e ._re. [y' / (<CR / ,o3's : ~. =:a G ~..{ Mtc ;S c t C *: - l,as$/ I 5L 2232 094 f0 I D EXCLUSION AREA u Figure 5.1.1-1 ERUi!SWICK - UNIT 2 5-2}}

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