ML19325D299

Amend 168 to License DPR-62,changing Tech Specs to Implement Expanded Operating Domain

ML19325D299
Person / Time
Site:	Brunswick
Issue date:	10/12/1989
From:	Adensam E Office of Nuclear Reactor Regulation
To:
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ML19325D300	List: ML19325D299
References
NUDOCS 8910230056
Download: ML19325D299 (30)

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~# UNITED STATES 3 1,,. NUCLE AR REGULATORY COMMISSION } {> wasmNotoN. o. c. gosss t %,.....,e CAROLINA F0WER & LIGHT COMPANY. et al. ) DOCr.Ef NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 2 i AHENDMENT TO FACILITY CPERATING LICENSE i Anenoment No.168 License No. DPR-62 i l The huclear Regulatory Coveission (the Commission; has found that: 1. A. The applicat1&n for amendment by Caro 11ria Pcwer & Light Con.pany (the licensee), dated March 29, 1989; complies with the standards and requirerr.ents of the Atomic Energy Act of 1954, as amended (the Act), and the Ccalission's rules and regulations set forth in 10 CFR j Chapter I; L. The facility will eperate in conformity with the application, the provisions of the Act, ano the rules and regulations of the Connission; C. There is reasonable assurance (i) that the activities authorized by this amenonent can be conducted without endangering the hetith and safety of the public, and (ii) that such activities will be i conducted in compliance with the Coinnission's regulations; O. The issuance of this amendh.ent will r,ct be inimical to the common defense and se'curity or to the health and safety of the l public; and E. The issuance of this anendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendnent; as.o paragraph 2.C.(2) of facility Optrating License No. DPR-62 is hereby amended to read as fo110ws: t l l 8910230056 891012, FDR ADOCK 0300 P ~

r .~ l 2 (2) Technical Specifications I The Technical Specifications contained in Appendices A and 8, as revised through Amendment No.168, are hereby incorporated in the i license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications. I 3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Orignal signed by: Elinor G. Adensam, Director Project Directorate II.1 l Division of Reactor Projects I/II Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical Specifications Date of Issuance: October 12, 1989 l i ?'" .d h[4 N 0FC :LA 4 DRPR:PM: PD21:D OGC )$': D: 2 :DRPR : ) NAME : PJW r y:dt: ..... ;............ :............ :. Jh.o3011{...::.....nsam IDATE:9/d/89

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j ATTACHMENT TO LICENSE AMENDMENT P:0.168 FACILITY OPEPATING LICENSE NO. DPR-62 l DOCKET h0. 50-324 l l Replace the folicking pages of the Appencix A Technical Specifications with the enclosee pages. The revissa 6rtas are indiceteo by marginal lines. Rencvs Pages Irisert Pages i 1 1 [ 1Y !Y X X 1-2 1-2 1-3 1-3 2-4 2-4 i 2-5 2-5 f 2-6 2-6 B 2-7 B27 0 2-8 [t 2-8 3/4 1-17 3/4 1-17 t 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 0/4 2-3 3/4 2-3 3/4 2-4 3/42-4 i 3/4 3-47 3/4 3-47 3/4 3 49 3/4 3-49 3/4 3-50 3/4 3-50 B 3/4 1-4 B 3/4 1-4 P 3/4 1-1 B 3/4 2-1 i B 3/4 2-2 6 3/4 2-2 B 3/4 2-3 8 3/4 2-3 6-25 6-25 6-26 6-26 l 1 1 l I J l 1

i i i l s. INDEX l I ptFINITIONS SECTION 1.0 DEFINITIONS PACE ACTION.......................................................... 1-1 AV E RAC E P LAN AR EX POS UR E......................................... 1 - 1 AVERACE PLANAR LINEAR HEAT CENERATION LATE...................... 1-1 CH ANW E L CAL I S LATI ON............................................. 1 - 1 CHANNEL CHECK................................................... 1-1 CHAKNEL FUNCTIONAL TEST......................................... 1-1 l CORE ALTERATION................................................. 1-2 g CORE OPERATINC LIMITS REP 0RT..................................... 1-2 CRITICAL POWER RATIO............................................ 1-2 i DOSE EQUIVALENT I-131........................................... 1-2 I -AVERACE DISINTEGRATION ENERGY................................ 1-2 i EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME.............. 1-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPOWSE TIME....... 1-3 I I F R EQU EN C Y WOT AT I ON.............................................. 1 - 3 CASEOUS RADWASTE TREATMENT SYSTEM............................... 1-3 l I D ENT I F I E D L E AKAC E.............................................. 1 - 3 l ISOLATION SYSTEM RESPOWSE TIME.................................. 1-4 LIMI TI WC CONTROL ROD PATTERN.................................... 1-4 LOGIC SYSTEM FUNCTIONAL T,EST.................................... 1-4 MAXIMUM AVERACE PLANAR HEAT CENERATION RATE RATIO................ 1-4 MEMBER (5) 0F THE PUBLIC......................................... 1-4 MINIMUN CRITICAL POWER LATIO.................................... 1-4 ODYW OPTION A.................................................... 1-4 O D YW O PT I ON B.................................................... 1 - 4 0FFSITE DOSE CALCULATIONAL MANUAL (ODCM)........................ 1-5 OPERABLE - OPERABILITY.......................................... 1-5 OPERATIONAL CONDIT!0W........................................... 1-5 BRUNSWICK - UNIT 2 I Amendment No. 28.751 168

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INDEX l LIMIT!NC CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS ( SECT!0N PACE 3/4.0 APPLICABILITY.............................................. 3/4 0-1 i 3/4.1 REACTIVITY CONTROL SYSTEMS I 3/4.1.1 SHUTDOWN MARCIN.......................................... 3/4 1-1 3/4.1.2 REACTIVITY AN0KALIES..................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod operability.................................. 3/4 1-3 i Control Rod Maximum Scram Insertion Times................ 3/4 1-5 Constol Rod Average Scram Insertion Times................ 3/4 1-6 [ Four Control Rod Group Scram 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 i 3/4.2.1 AVERACE PLANAR LINEAR HCAT CENERATION RATE............... 3/4 2-1 r 3/4.2.2 M I N I M UM C RI T I C AL POWE R RAT I 0............................. 3/4 2-2 r i l t BRUNSWICK - UNIT 2 IV Amendment No. SS.1$7. 151. 168

t INDEX BASES i SECTION PACE 3/4.0 APPLICABILITY.............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS l 3/4.1.1 SHUTDOWN MARCIN.......................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES,.................................... S 3/4 1-1 3/4.1.3 CONTROL R0DS............................................. B 3/4 1-1 3/4.1.4 CONTROL ROD PROCRAM 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 LATE............... B 3/4 2-1 i 3/4.2.2 MINIMUM CRITICAL POWER RAT!0....t........................ B 3/4 2-2 l 3/4.3 INSTRUMENTATION + 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................ B 3/4 3-1 3/4.3.2 I SOLATION 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 SLOCK INSTRUMENTATION............. B 3/4 3-2 3/4.3.5 MONITORING 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 C00LINC SYSTEM ACTUATION INSTRUMENTATION....................................... B 3/4 3-7 i l l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM..................................... B 3/4 4-1 1/4.4.2 SAFETY / RELIEF VALVES..................................... B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-1 l f i BRUNSWICK - UNIT 2 X Amendment No. 74. J$3 168

l l DEFINITIONS I l CHAWWEL FUNCTIONAL TEST (Continued) l b. Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip i functions. 1 CORE ALTERATION CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core j 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. CORE OPERATIWC LIMITS REPORT The CORE OPERAT!WC LIMITS REPORT is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall 4e 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 operatien within these core operating limits is addressed in individual specifications. l l CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in an assambly which is calculated, by application of an NRC approved CPR correlation, to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. l i DOSE EQUIVALENT I-131 i 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 misture of I-131, 1-132, I-133, I-134, and I-135 actually present. The following is defined equivalent to 1 uCL of I-131 as determined from Table I!! of [ TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites": 1-132, 28 9Cil I-133, 3.7 ucil I-134, 59 ucil I-135, 12 9C1. I -AVERACE DISINTECRATION ENERCY i i shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gaaEha energies per disintegration (in MeV) for isotopes with half lives creater than 15 minutes making up at least 95% of the total non-iodine activity in the coolant. I i t BRUNSWICK - UNIT 2 1-2 Amendment No. H. 7H M1,168

DEFINITIONS EMERCEWCY CORE C00LINC SYSTEM (ECCS) RESPONSE TIME i 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 safety function (i.e., the valves travel tc their required positions, pump ( discharse pressures reach their required values, etc.). Times shall include diesel senerator starting and sequence loading delays where applicable. END-OF-CYCLE RECIRCULATIOW PUMP TRIP SYSTEM RESPOWSE TIME The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to recirculation pump breaker trip from initial movement of the associatedt a. Turbine stop valves, and b. Turbine control valves. FREQUEWCY NOTATIOW The FREQUEWCY WOTATION specified for the performance of Surveillance Requirements shall correspond to the inter'vals defined in Table 1.1. CASCOUS RADWASTE TREATMENT SYSTEM A CASEOUS RADWASTE TREATNENT SYSTEM is any system designed and installed to reduce radioactive gaseous affluents 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 LEAKACE IDENTIFIED LEAKACE shall beI a. Leakage into collection systems, such as pump seal or valve packing leaks, that is capcured and conducted to a sump or collecting tanit, or b. Leakage into the containment atmosphere from sources that ara both specifically located and known either not to interfere with the operation of the leakage detection systems or not be PRESSURE BOUO ARY LEAKACE. BRUNSWICK - UNIT 2 1-3 Amendment No.II III 163

TABLE 2.2.1-1 si E REACTOR PROTECTION SYSTEM INSTRIMENTATION SETPOINTS 5 M ALIDWABLE FUNCTIONAL UNIT AND INSTR 4NENT NIM 8ER TRIP SETPOINT VALUES c I l. Interanediate Range Monitor, Neutron Flum - High *I $ 120 divisions of full scale $ 120 divisions u of full scale 1 2. Average Power Range Monitor Neutron Flum - High, 151(b) < 15Z of RATED THERMAL POWER $ 15% of RATED a. THERMAL POWER High g g d Simulated Thermal Power - $ (0.M W + 641) with a ~ $ (0.66 W + 67%) with Flow b. saaminum $ 113.5% of RATED a ananimum $ 115.5% THERMAL POWER of RATED THERMAL 1 POWER Fixed Neutron Flum - High(d) $ 120% of RATED THERMAL POWER $ 120I of RATED c. THERMAL POWER 3. Reactor Vessel Steam Dorae Pressure - High $ 1945 psig $ 1045 psig 4. Reactor Vessel Water Level - Low, level 1 > +162.5 inches 8I 3 +162.5 inches 8} I I 5. Main Steam Line Isolation Valve - Closure *I < 101 closed I $ 101 closed 6. Main Steams Line Radiation - Nigh 1 3 x full power background $ 3.5 m full power background y -E 7. Drywell l'ressure - High $ 2 psig $ 2 psig ~2 E,E 8. Scram Discharge Volume Water Level - High $ 109 gallons

109 gallons Turbine Stop Valve-Closure (f)

$ 102 closed $ 10% closed gy 9. E

10. Turbine Control Valve Fa

closure,

-g ,t* Control Oi1 Pressure-Low 3 500 psig > 500 psig _.,. ~ - - -..

v e < TAgtE 2.2.1-1 (Continued) ] REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS t NOTES ( (a) The Intermediate Range Monitor scram functions are automatically bypassea I when the reactor mode switch is placed in the Run position and the Average Power Range Monitors are on scale. (b) This Aversge Power Range Monitor scram function is a fixed point and is I increased when the reactor mode switch is placed in the Run position. (c) The Average Power Range Monitor scram function is varied, Figure 2.2.1-1, as a function of the fraction of rated recirculation loop flow (W) in percent. (d) The APRM flow-blased simulated thermal power signal is fed through a time l l constant circuit of approximately 6 seconds. The APRM fixed high neutron r flux signal does not incorporate the time constant, but responds directly ( to instantaneous neutron flux. t (e) The Main Steam Line Isolation Valve-Closure scram function is automatically bypassed when r.he reactor mode switch is in other than the Run position. I (f) These scram functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL POWER as measured by turbine first stage pressure. (g) Vessel water levels refer to REFERENCE LEVEL ZERO. i BRUNSWICK - UNIT 2 2-5 Amendment No. ($, Jf9 168

l l i ) \\ I i i 1 1 120 i i ..... _l / AMlW Plow 84A8 SCRAM / \\ / i Q/ \\ 100 i [ i P / l } j y g I l 'I / l eo f P i 1 1 NoMwAL lxpecten t g 2 F,0w cortmot ue, ..,.... 1 / 1 eo i r / i l / i I i i i l' I u } 'O \\ come manuat powsm uur - 20% avun speso use 2e% 4 / i / i l J i i 30 ^ 1 I L l } . NATURAL 4 c:ncuts.G.r { uNe a o 30 Ao to to too tro come Plow mate m et reise Figure 2.2.1-1. AFRH Flow Bias Scras Relationship to Normal i Operating Conditions i l i i I BRUNSWICK - UNIT 2 2-6 Amendment No. $$,f$, 131. 159,168 l 1

i 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Rosetor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the Reactor Trips are set for each i 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 Ranae Monitor. Neutron Flux - liith 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 subcritical and the IRMs are not yet on scale. Additional conservatism was taken in this analysis by assur.:.ng the IRM channel closest to the rod being withdrawn is bypassed. The l 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 Ranae Monitor For operation 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 minori 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 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 l rise is very slow. Generally, the heat flux is in near equilibtium with the fission rate. In an assumed uniform rod withdrawal approach to the trip l l level, the rate of power rise is.not more than 5% of RATED THERMAL POWEk per BRUNSWICK - UNIT 2 B 2-4 Amendment No.Jpf, J$J 168

2.2 LIMITIWC SAFETY SYSTEM SETTINGS BASES (Continued) 2. Average Power Ranae Monitor (Continued) minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% neutron flux trip remains active until the mode switch is placed in the RUN position. The APRM flux scram trip in RUN mode consists of a flow biased simulated thermal power (STP) scram setpoint and a fixed neutron flux scram setpoint. The APRM flow biased neutron flux signal is passed through a filtering network with a time constant which is representative of the fuel dynamics. This provides a flow referenced signal, e.g., STP, that approximates the average heat flux or thermal power that is developed in the core during transient or steady-state conditions. The APRM flow biased simulated thermal power scram trip setting at full recirculation flow is adjustable up to the nominal trip setpoint of 113.5% of RATED THERMAL POWER. This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100'F feedwater heating event, than would result with the 120% fixed neutron flux scram trip. The lower flow biased scram setpoint therefore decreases the severity, ACPR, of a slow thermal transion't and allows lower operating limits if such a transient is the limiting abnormal operational transient during a, 'certain exposure interval in the fuel cycle. The APRM fixed neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux. This scram setpoint scrams the reactor during fast power increase transients if credit is not taken for a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow biased simulated thermal power scram. l l The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. 3. Reactor Vessel Steam Dosie 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 fluz, 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 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. I l l BRUNSWICK - UNIT 2 B 2-5 Amendment No.7$7,7$7 l 168 t

i 1 e .-s I I LIMITING SAFETY SYSTEM SETTING gASE5 (Continued) 4 Reactor Vessel Water Level-Low, Level #1 The reactor water 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 steam line trip was provided to give protection against rapid depressurirstion 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 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 mode switch in the Run position anticipates the* pressure and flua transients which occur during normal or inadvertent isolation valve closure. 6. Main Steam Line Radiation - Hiah 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, i the Main Steam Line Isolation Valves are closed to limit the release of fission products. The trip setting is high enough above background radiation r l levels to prevent spurious scrams, yet low enough to promptly detect gross i failures in the fuel cladding. l 7. Drywell Pressure-High j 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 casuing spurious trips. 8. Scram Discharme 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 1 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. BRUNSWICK - UNIT 2 B 2-6 Amendment No, 157,168 l

.o LIMITING SAFETY SYSTEM SETTINCS l BASES (Continued) l 9. Turbine Stop Valve-Closure l The turbine stop valve closure trip anticipates the pressure, neutron-flux, and heat flux increases that would result from closure of the stop l valves. With a trip setting of 10% of valve closure from full open, the I resultant increase in heat flux is such that adequate thermal margins are l maintained even during the worst case transient that assumes the turbine i bypass valves remain closed. This scram is bypassed when the turbine steam l l flow is below that corresponding to 30% of RATED THERMAL POWER, as measured by the turbine first-stage presJure. l 10. Turbine Control Valve Fase Closu*e. Control Oil Pressure - Low Low turbine control valve hydraulic pressure will initiate the Select Rod Insert function and the preselected group of control rods will be fully inserted. Select Rod Insert is an operational aid designed to insert a predetermined group of control rods'immediately following either a generator load rejection, loss of turbine control valve hydraulic pressure, or by manual operator action using a switch on the R-T,C board. The assignment of control rods to the Select Rod Insert function is based on the start-up and fuel warranty service associated with each control rod pattern, on RCS considerations, and on a dynamic function of both time and core patterns. Approximately ten percent of the control rods in the reactor will be assigned to the Select Rod Insert function by the operator. This selection will be accomplished by moving the rod scram test switch for those rods from the Normal position to the Select Rod Insert position. Any rod selected for Seiect Rod Insert shall also have other rods in its i notch group selected to ensure that the RSCS criteria of plus-minus one notch position equality is met when the rod pattern is greater than 50% ROD DENSITY and THERMAL POWER < 20% of RATED THERMAL POWER. It is possible that a rod pattern within these limits may occur after the Select Rod Insert function ) operates. l In order to reduce the number of reactor scrams, a 200 millisecond time delay, referenced from the low turbine control valve hydraulic pressure and Select Rod Insert signals, was incorporated to determine turbine bypass valve status via limit switches prior to initiating a reactor scram. If the turbine bypass valves opened in < 200 milliseconds, the reactor scram was bypassed. It was found that during certain reload cycles the MCPR penalties involved with this time delay were more penalizing than the number of scrams saved; therefore, CP6L requested and received NRC approval to set this time at "0" in Amendment No. 14. With the timer set at "0", Select Rod Insert and RPS trip will be initiated simultaneously. 1 1 1 BRUNSWICK - UNIT 2 B 2-7 Amendment No.757, 168

LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) i l 10. Turbine Control Valve Fast Closure, Control Oil Pressure - Low (Continued) The control valve closure time is approximately twice as long as that for tho'stop valves which means that resulting transiento, while similar, are less severe than for stop valve closure. No fuel darnage occurs, and reactor system pressure does not exceed the safety relief valve setpoint. This is an anticipatory scram and results in reactor shutdown before any significant i increase in pressure or neutron flux occurs. This scram is bypassed when I turbine steam flow is below that corresponding to 30 percent of RATED THERMAL POWER, as measured by turbine first-stage pressure. j i l l l l ( i BRUNSWICK - UNIT 2 B 2-8 Amendment No. 757.168

REACTIVITY CONTROL SYSTEMS l i R0D Bl.0CK MONITOR LIMIT!.1C CONDITION FOR OPERATION 3.1.4.3 Soth Rod Block Monitor (RBd) channels shall be OPERABLE. APPLICABILITY OPERATIONAL CONDITION 1 with: s. THERMAL POWER Breater than 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER and the MINIMUM CRITICAL POWER RATIO (MCPR) less than 1.70, or b. THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER and the MCPR less than 1.40. ACTIOWs With one R8N channel inoperable, POWER OPERATION may continue a. provided that eithers 1. The inoperable RBM channel is restored to OPERA 8LE status within 24 hours, or 2. The redundant R8N is demonstrated OPERABLE within 4 hours and at least once per 24 hour's until the inoperable RBM is restored to OPERABLE status within 7 days. Otherwise, trip at least one rod block monitor channel. b. With both RBM channels inoperable, prip at lesst one rod block monitor channel within one hour. SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required 18N channels shall be demonstrated OPERA 8LE by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and durin8 the OPERATIONAL CONDITIONS specified in Table 4.3.4-1. BRUNSWICK - UNIT 2 3/4 1-17 Amendment No. $$,7M,157,7$1,168

1a 0 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE 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 limic values that have been approved for the respectivs 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, adjusted for core flow and core power, for the most limiting lattice (excluding natural uranium) of each type of fuel shown in the 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 Technical Specification 3.2.1 initiate corrective action within 15 minutes and continue corrective action so that APLHCR is within the required limits within 4 hours or reduce THERMAL POWER to 1ses than 25% of RATED THERMAL POWER within the next 4 hours. 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 hours, b. Within 12 hours after completion of a THERNAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR. 1 l l j i BRUNSWICK - UNIT 2 3/4 2-1 Amendment No. S3,773.767.168

.e POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO LIMITING CONDITICN FOR OPERATION 3.2.2.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, core power, and cycle average exposure, shall be equal to or greater than the MCPR limit specified in the CORE OPERATING LIMITS REPORT. The MCPR limits for ODYN OPTION A and ODYN OPTION 3 analyses, used in the above determination, shall be specified in the CGRE OPERATING LIMITS REPORT. i 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, core power, and cycle average exposure, less than the applicable MCPR Limit specified in the CORE OPERATING LIMITS i REPORT, initiate corrective action within 15 minutes and restore 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 hours. SURVEILLANCE REQUIREMENTS 4.2.2.1 MCPR, as a function of core flow, core power, and cycle average exposure, shall be determined to be equal to or greater than the applicable limit determined of Specification 3.2.2.11 a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is c. operating in a LIMITING CONTROL ROD PATTERN for MCPR. l l l BRUNSWICK - UNIT 2 3/4 2-2 Amendment No. 197,773,1#7,753,757,168

POWER DISTRIBUTION LIMITS M4.2.2 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B) LIMITINC CONDITION FOR OPERATION 3.2.2.2 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 less than or equal to the Option 8 scram time limit (tg), where t,y, and t g are determined as follo st N 'i i =i1 ,.here t,,, ,i i=1 i = Surveillance test number, n = Number of surveillance tests performed to date in the cycle 4 (including BOC), th Ng = Number of rods tested in the i surveillance test, and tg = Average scram time to notch 36 for surveillance test i B""* '0' ( n N.) (c), wheret t 1 i=1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including BOC), g = Number of rods tested in the ich surveillance test N Ng = 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), l 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. BRUNSWICK - UNIT 2 3/4 2-3 Amendment No. 197,773,747,753,767,168

i 1 i 1 POWER DISTRIBUTION LIMITS i I LIMITINC CONDITIONS FOR OPERATION (Continued) ACTION: ) Within twelve hours after determining that t,y, is greater than tg, the operating limit MCPRs shall be eithers Adjusted for each fuel type such that the operating limit MCPR a. is the maximum of the non-pressurization transient MCPR operating limit specified in the CORE OPERATING LIMITS REPORT or the adjusted pressurization transient MCPR operating limits, where che adjustment is made by: 1 t t adjusted ' option B ~ T Ption A option B B where: tg = 1.05 seconds, control rod average scram insertion time 1 limit to notch 36 per Specification 3.1.3.3, l I option A " S ecified in the CORE OPERATING LIMITS REPORT, MCPR P option B " S ecified in the CORE OPERATING LIMITS REPORT, or, MCPR P b. The OPTION A MCPR limits specified in the CORE OPERATINC LIMITS REPORT. SURVEILLANCE REQUIREMENTS 4.2.2.2 The values of t and t shall be determined and compared each time ( a scram time test is perfoImed. kherequirementforthefrequencyofscram time testing shall be identical to Specification 4.1.3.2. 1 [ BRUNSWICK - UNIT 2 3/4 2-4 Amendment No. 107,77),797,75), 7$7,168

i INSTRUMENTATION 3/4.3.4 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION t 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 Satpoint column of Table 3.3.4-2. APPLICABILITY: As shown in Table 3.3.4-1. ACTION With a control rod withdrawal block instrumentation channel trip a. setpoint less conservstive 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 OPERA 8LE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value. With' the requirements for the minimum number of OPERABLE channels not b. i satisfied for one trip system, POWER OPERATION may continue provided that.either 1. The inoperable channel (s) is restored to OPERA 8LE status within l 24 hours, or 2. The redundant trip system is demonstrated OPERABLE within 4 hours and at least once per 24 hours until the inoperable j channel is restored to OPERA 8LE status, and the inoperable channel is restored to OPERABLE status within 7 days, or Otherwise, place at least one trip system in the tripped condition within the next hour. i With the requirements for the minimum number of OPERABLE channels not I c. satisfied for both trip systems, place at least one trip system in the tripped condition within one hour. I d. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5. SURVEILLANCE REQUIREMENTS 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 i CONDITIONS and at the f requencies shown in Table 4.3.4-1. i l 1 I BRUNSWICK - UNIT 2 3/4 3-47 Amendment No. 769,757 168 i

j l \\ TABLE 3.3.4-1 (Continued) CONTROL ROD WITHDRAWAL BLCCK INSTRUMENTATION r NOTES (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. t (c) This function is bypassed when the associated IRM channels are on range 8 or higher. (d) A total of 6 IRH instruments must be OPERABLE. t (e) This function is bypassed when the IRM channels are on ' range 1. (f) When (1) THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER and MCPR is less than 1.70, or (2) THERMAL POWER is greate'r than or equal to 90% of RATED THERMAL POWER and MCPR is less than 1.40. (g) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. (h) This signal is contained in the Channel A logic only. i l BRUNSWICK - UNIT 2 3/4 3-49 Amendment No. 7@$,7$$, 7$@,168

e g TABLE 3.3.4-2 Siy COtrTROL ROD WITHDRAWAL BIDCK INSTRUMENTATION SETPOINTS 2 M TRIP FUNCTION TRIP SETFOINT ALIDWABLE VALUE C 1. APRM h Upscale (Flow Biased) $ (0.66W + 581)(*) with a 5 (0.66W + 611)(a) with a a. maximum of $ 108I of maximum of < 110% of u RATED THERiiAL POWER RATED 111ERMAL POWER b. Inopera tive NA NA c. Downscale > 3/125 of full scale > 3/125 of full scale d. Upscale (Fixed) 312%ofRATEDTHERMALPOWER 312%ofRATEDTHERMALPOWER 2. ROD BLOCK MONITOR As speci fied in the CORE As specified in the CORE a. Upscale OPERATING LIMITS REPORT OPERATINC LIMITS REPORT N* b. Inopera t i ve NA NA Y c. Downscale 3 94/125 of full scale NA l E 3. SOURCE RANCE MONITORS a. Detector not full in MA NA 5 5 b. Upscale $ 1 x 10 cps $1x 10 cps c. Inoperative NA NA d. Downscale 3 3 cps 3 3 cps 4. INTERMEDIATE RANCE MONITORS a. Detector not full in NA' NA b. Upscale $ 108/125 of full scale $ 108/125 of full scale c. Inoperative NA NA g d. Downscale 3 3/125 of full scale 3 3/125 of full scale o E 5. SCRAM DISCHARCE VOLUME f, a. Water Level *iigh $ 73 gallons 5 73 gallons ,2: g (a) Where W is the f raction of rated recirculation loop flow in percent. L O

l ^ 1 i REACTIVITY CONTROL SYSTEM l BASES CONTROL ROD PROCRAM CONTROLS (Continued) Use of the Banked Position Withdrawal Sequence (BPWS) ensures that in the event of a control ~ rod drop accident, the peak fuel enthalpy will not be greater than 280 cal /gm (Reference 4). The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted. The analysis of the rod drop accident is presented in Section 15.4.6 of t the FSAR, Updated and the techniques of the analysis are presented in a topical report (Reference 1) and two supplements (References 2 and 3). The RBM is designed to automatically prevent fuel damage in the event.of erroneous rod withdrawal from locations of high power denetty during high power operation. The RBM is only required to be operable when the Limiting condition described in Specification 3.1.4.3 exists. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal l soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods. Further discussion of the RBM system is provided in Reference 5. 3/4.1.5 STAFDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for maintaining the reactor suberitical in the event that insufficient rods are inserted in the core when a scram is called for. The volume and weight percent of poison material in solution is based on being able to bring the reactor to the subcritical condition as the plant cools to ambient condition. The temperature requirement is necessary to keep the sodium pentaborate in solution. Checking the volume and tem houes assures that the solution is available for use.perature once each 24 With redundant pumps and a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time wtth the system inoperable or for longer periods of time with one of the redundant components inoperable. i I 1. C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis for Large BWRs " C. E. Topical Report NEDO-10527, March 1972. 2. C. J. Paone, R. C. Stirn, and R. M. Yound, Supplement 1 to NEDO-10527, July 1972. 3. J. A. Haus, C. J. Paone, and R. C. Stirn, addendum 2 " Exposed Cores" supplement 2 to NEDO-10$27, January 1973. 4. NEDE-24011-P-A, "Ceneral Electric Standard Application for Reactor Fuel," Revision 6, Amendment 12. I 5. NEDC-31654P, " Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant," February 1989. l l r BRUNSWICK - UNIT 2 B 3/4 1-4 Amendment No. 157,168

l l J 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 4 June 1971 considering the postulated effects of fuel pellet densification. 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 CORE OPERATINC LIMITS REPORT for each fuel type and, when required, for the most limiting lattico for i multiple lattice fuel bundle types. Power and flow dependent adjustments are provided in the CORE OPCRATING LIMITS REPORT to assure that the fuel thermal-mechanical design criteria are preserved daring abnormal transients initiated from off-rated conditions. 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. i i Mechanical Design Analysist 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 i 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 j comply with the ECCS limits specified in 10 CFR 50.46. The analysis is l performed for the most limiting break size, break location, and single failure combination for the plant. j The Technical Specification APLHCR limit is the most limiting composite of the fuel mechanical design analysis APLHCR and the ECCS APLHCR limit. i l l l l BRUNSWICK - UNIT 2 B 3/4 2-1 Amendment No.S),J/),7$7 168 l l i

h e O a O POWER DISTRIBUTION LIMITS BASES 3/4.2.2 MINIMUM CRITICAL POWER RATIO l The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.2 are derived from an established fuel cladding integrity Safety Limit MCPR approved by the NRC and an analysis of abnormal cperational 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 below the Safety Limit MCPR at any time during the transient, assuming an r 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 POWER 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 l conditions are given in References 1 and 2 and the CORE GPERATINC LIMITS REPORT. At core thermal power lev 61s less than or equal to 25% RATED THERMAL POVER, the reactor will be operating at minimum recirculation pump speed and the f 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 exce'ss of requirements by a considerable margin. With tnis 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 i calculating MCPR above 25% RATED THERMAL POWER is sufficient since power l distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculcting 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 l operation at a thermal limit. 1 BRUNSWICK - UNIT 2 B 3/4 2-2 Amendment No. 83.747. 767, 168

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• e P0WER DISTRIBUTION LIMITS BASES

\\ 1 References 1. NEDE-24011-P-A, " Central Electric Standard Application for Reactor Fuel", latest approved version. 2. NEDC-31654P, " Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant," February 1989. 1 i BRUNSWICK - UNIT 2 B 3/4 2-3 Amendment No. 191,773,757.151 168

1 I ~ e \\ l 1 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 appilcable reference specification. Inoperable Seismic Honitoring Instrumentation, Specification 3.3.5.1. a. 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 Ef fluents Dose, Specification 3.11.1.2. j. Liquid Radweste Treatment, Specification 3.11.1.3. k. Dose - Noble Cases, Specification 3.11.2.2. l. Dose - 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 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 followings a. The AVERACE PLANAR LINEAR HEAT CENERATION RATES (APLHCR) for Specification 3.2.1 including core flow and core power adjustments. 1 BRUNSWICK - UNIT 2 6-25 Amendment No.770,767 168

-a.,,, w,. a ,e a + a.n s:_.- + -,,, -, e a 4 w,. l s ADMINISTRATIVE CONTROLS CORE OPERATINC LIMITS REPORT (Continued) b. The core flow and core power adjustments for Specification 3.2.2.1 [ The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.2.1 c. and 3.2.2.2. d. The rod block monitor upscale trip setpoint and allowable value for Specification 3.3.4. and shall be documented in the CORE OPERATING LIMITS REPORT. 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. WEDE-24011-P-A, "Ceneral Electric Standard Application for Reactor a. Fuel" (latest approved version). b. The May 18, 1984 and October 22, 1984 NRC Safety Evaluation Reports for the Brunswick Reload Methodologies described int l 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-8," February 1983. + i 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-suchanical limits, core thermal-hydraulic 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, inc?.uding any mid-cycle revisions or supplements shall be provided, upcn issuance for each reload cycle, to the NRC Decument 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 yearst i Records and loss of facility operation covering time interval at each a. power level. BRUNSWICK - UNIT 2 6-26 Amendment No. JJ0,J$J 168 l -}}