Text

Proposed Revisions to Tech Specs to Allow Unit Operation Up to 50% of Rated Thermal Power W/One Recirculation Loop Out of Svc
	ML20053E375
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Site:	Brunswick
Issue date:	06/03/1982
From:	; CAROLINA POWER & LIGHT CO.
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2.0 SAFETY IIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Iow Flow) 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam done pressure less than 800 psia or core flow less than 10% of rated flow.

APPLIC/BILITY: CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POKER 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 S3UTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

THERMAL POWER (High Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 for two recirculation loop operation and less than 1.08 for one recirculation loop operation with the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow.

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.07 for two recirculation loop operation or less than 1.08 for one recirculation loop operation and the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

! 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.

APPLICABILITY: 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 ;< 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

8206070843 820603 i PDR ADOCK 05000324 l P PDR BRUNSWICK - UNIT 1 2-1 Amendment No.

J

N TABLE 2.2.1-1 5

h REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS R

, ALLOWABLE e FUNCTIONAL UNIT AND INSTRUMENT NUMBER TRIP SETPOINT VALUES z

5 1. Intermediate Range Monitor, Neutron Flux - High(I) <120 divisions of full scale <120 divisions

~ (C51-IRM-K601A,B,C,D,E,F,G,H) of full scale

2. Average Power Range Monitor (C51-APRM-CH.A,B,C,D,E,7)

a. Neutron Flux - High 15%(2) 115% of RATED DIERMAL POWER 115% of RATED THERMAL POWER

b. Flow Biased Neutron Flux - High(3)(4)(8) <(0.66 W + 54%) - 0.664W <(0.66 W + 54%) - 0.66AW

c. Fixed Neutron Flux - High(4) 1120% of RATED THERMAL POWER 1120% of RATED THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High (B21-PTM-N023A-1,B-1,C-1,D-1) <1045 psig <1045 psig

4. Reactor Vessel Water Level - Low, Leve1#1(7) _>+162.5 inches _>+162.5 inches (B21-LTM-N017A-1,B-1,C-1,D-1)

5. Main Steam Line Isolation Valve - Closure (5) ~

<10% closed ^

<10% closed (B21-F022A,B,C,D; B21-F028 A,B,C,D)

6. Main Steam Line Radiation - High <3 x full power background 13.5 x full power (D12-RM-K603A,B,C,D) background N -

8 .

B n

l .

TABLE 2.2.1-1 (Continued)

~

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TABLE NOTATION (1) The Intermediate Range Monitor scram functions are automaticall bypassed when the reactor modo switch is placed in the Run position and the Average Power Range Monitors are on scale.

(2) This Average Power Range Monitor scram function is a fixed point and is increased when the reactor mode switch is placed in the Run position.

(3) The Average Power Range Monitor scram function is varied, Figure 2.2.1-1, as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.

(4) The APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.

(5) The Main Steam Line Isolation Valve-Closure scram function is automatically bypassed when the reactor mode switch is in other than the Run position.

(6) These scram functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL POWER.

(7) Vessel water levels refer to REFERENCE LEVEL ZERO.

(8) W = Loop recirculation flow rate in percent of rated core flow.

i AW = Difference between two loop and single loop effective recirculation l drive flow rate at the same core flow. During single loop operation, the reduction in trip setting (-0.66AW) is accomplished by correcting the flow input of the flow biased scram to prevent the original (two loop) relationship between APRM scram setpoint and recirculation drive flow or by adjusting the APRM flux trip setting.

AW = 0 for two loop operation.

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BRUNSWICK - UNIT 1 2-6 Amendment No.

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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'ench type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the following limits:

a. During two recirculation loop operation, the limits are shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, or 3.2.1-6.

b. During single recirculation loop operation, the limits shown in the above figures shall be multiplied by the following reduction factors:

Fuel Type Reduction Factor 8X8 0.85 8X8R 0.85 P8X8R 0.85 APPLICABILITY: CONDITION 1, when THERMAL POWER )_ 25% of RATED THERMAL POWER.

ACTION: With an APLHGR exceeding the limits of Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, or 3.2.1-6, initiate corrective action within 15 minutes and continue corrective action so that APLHGR is within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGR's shall be verified to be equal to or less than the applicable limit determined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, or 3.2.1-6:

i

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

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

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

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i 1

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

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l - . - - - - -. - - -

l l

l 1

3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The flow biased APRM scram trip setpoint (S) and rod block trip set point (SRB) shall be established according to the following relationship:

S j[ (0.66W + 54% - 0.66AW) T

~

S *

RB db where S and S RB are in percent of RATED THERMAL POWER.

W = Loop recirculation flow in percent of rated flow, AW = Difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting (- 0.666W) is accomplished by correcting the flow input of the flow biased scram to prevent the original (two loop) relationship between APRM scram setpoint and recirculation drive flow or by adjusting the APRM flux trip setting.

AW = 0 for two loop operation.

T= Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T,< 1.0), and Design TPF for: 8x8 fuel - 2.43 8x8R fuel = 2.39 P8x8R fuel = 2.39 APPLICABILITY: CONDITION 1, when THERMAL POWER 1 25% of RATED THERMAL POWER.

ACTION:

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

SURVEILL WCE 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:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , l b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and i

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

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

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO -

LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of co,re flow, shall be equal to or greater than MCPR x the Kg shown in Figure 3.2.3-1 where MCPR values are:

BOC3* to EOC3** EOC3-2000 MWD /t

-2000 MWD /t to EOC3 8x8 fuel 1.24*** 1.30***

8x8R fuel 1.24*** 1.30***

P8x8R fuel 1.30*** 1.32***

APPLICABILITY: CONDITION 1, when THERMAL POWER )_ 25% of RATED THERMAL POWER ACTION:

With MCPR, as a function of core flow, less than the applicable limit determined f rom Figure 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER TO LESS TRAN 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, as a function of core flow, shall be determined to be equal to or greater than the applicable limit determined from Figure 3.2.3-1:

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

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

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

Beginning of Cycle 3.

- End of Cycle 3.

- - For single recirculation loop operation, these limits shall be increased by 0.01.

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

E

$ TABLE 3.3.4-2

?!

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS h

TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE E

C 1. APRM (C51-APRM-CH.A,B,C,D,E,F)

a. Upscale (Flow Biased) < (0.66W + 42% - 0.66AW) T* < (0.66W + 42% - 0.66AW) T*

MTPF MTPF

b. Inoperative NA NA

c. Downscale > 3/125 of full scale > 3/125 of full scale ,

d. Upscale (Fixed) {12%ofRATEDTHERMALPOWER [12%ofRATEDTHERMALPOWER

2. ROD BLOCK MONITOR (C51-RBM-CH.A,B)

a. Upscale < (0.66W + 41% - 0.66AW) T* < (0.66W + 41% - 0.66AW) T*

MTPF MTPF M

b. Inoperative NA NA

c. Downscale > 3/125 of full scale > 3/125 of full scale

3. SOURCE RANGE MONITORS (C51-SRM-K600A,B,C,D)

a. Detector not full in NA NA

b. Upscale < 1 x 105 cps < 1 x 105 cps

c. Inoperative NA NA d Downscale > 3 cps > 3 cps

4. INTERMEDIATE RANGE MONITORS (C51-IRM-K601A,B,C,D,E,F,G,H)

a. Detector not full in NA NA g b. Upscale < 108/125 of full scale < 108/125 of full scale g c. Inoperative NA NA g d. Downscale > 3/125 of full scale > 3/125 of full scale E

T=2.43 for 8x8 fuel T=2.39 for 8x8R fuel T=2.39 for P8x8R fuel

_ . -. . -- - _ . = . _ = . . _ - . -

3/4.4 REACTOR COOLANT SYSTEM

~

3/4.4.1 RECIRCULATION SYSTEM -

RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant recirculation loops shall be in operation with the cross-tie valve closed, the pump discharge valves OPERABLE and the pump discharge bypass valves OPERABLE or closed, except as permitted under single loop operation.

APPLICABILITY: CONDITION 1* and 2*.

ACTION:

With both recirculation loops not in operation, operation may continue; restore one loop to operation within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With only a single recirculation loop in service, plant operation may continue

, with the implementation of the requirements in Specifications 2.1.2, 2.2.1, 3.2.1, 3.2.2, 3.2.3.1, 3.3.4, and 3.4.1.2. These requirements shall be in

', effect within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the removal of one recirculation loop from service, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4 SURVEILLANCE REQUIREMENTS 4.4.1.1 Each pump discharge valve and bypass valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full j travel during each COLD SHUTDOWN which exceeds 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if not performed in the previous 31 days.

t 4.4.1.2 Each pump discharge bypass valve, if not OPERABLE, shall be verified to be closed at least once per 31 days.

See Special Test Exception 3.10.4.

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

, REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 'All jet pumps associated with an OPERABLE recirculation lo'op shall be OPERABLE.

APPLICABILITY: CONDITIONS 1 and 2 ACTION: With less than 10 jet pumps OPERABLE in an OPERABLE recirculation loop, be in at least hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.1.2 Each of the above required jet pumps shall be demonstrated OPERABLE t prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that all of the following conditions do not occur

simultaneously. .

a. The recirculation pump flow differs by more than 10% from the established speed-flow characteristics,

b. The indicated total core flow differs by more than 10% from the core flow value derived from established power-core flow relationships, and

c. The diffuser-to-lower plenum differential pressure reading on any individual jet pump varies from the mean of all jet pump differential pressures, in that loop, by more than 10%.

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

SPECIAL TEST EXCEPTION RECIRCULATION LOOPS .

LIMITING CONDITION FOR OPERATION ;.

3.10.4 The requirement of Specification 3.4.1.1 that one or two recirculation loops be in operation may be suspended for ap to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during the performance of start-up and PHYSICS TESTS.

APPLICABILITY: CONDITIONS 1 and 2.

, ACTION:

With the above specified time limit exceeded, deenergize the scram solenoid valves. _

SURVEILLANCE REQUIREMENTS 4.10.4 The time during which the above specified requirement hcs been suspended shall be verified to be less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at least once per hour during start-up and PHYSICS TESTS.

l BRUNSWICK - UNIT 1 3/4 10-4 Amendment No.

SAFETY LIMITS BASES (Continued) 2.1.2 THERMAL POWER (High Pressure and Iligh Flow)

The fuel cladding integrity Safety Limit is set such that no fuei damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power, result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transit. ion considering the power distribution within the core and all uncertainites.

The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB , which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L),

(GEXL), correlation.

The CEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation. The conditions are:

l Pressure: 800 to 1400 psia l

Mass Flux: 0.1 to 1.25 X 106 lb/hr-ft 2 l

l Inlet Subcooling: 0 to 100 Btu /lb l

Local Peaking: 1.61 at a corner rod to

! 1.47 at an interior rod l The analyses used for Safety Limit MCPR calculation are used for both one and

! two loop recirculation loop' operation. For single loop operation, these analysesarereviewedandfoundtohaveadditionaluncytaintieswhichwill l require the MCPR safety limit to be increased by 0.01 i

Reference 1

" General Electric BWR Thermal Analysis Basis (CETAB) Lata, Correlation and Design Application," NED0-10958 and NEDE-10958.

2

" Brunswick Steam Electric Plant Units 1 and 2 Single-Loop Operation,"

NEDO-24344.

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

3/4.4 REACTOR COOLANT SYSTEM BASES .

3/4.4.1 RECIRCULATION SYSTEM Operation with one recirculation loop in service is permitted. In such instances, th2 designated adjustments for APRM rod block and scram setpoints, RBM setpoints, HCPR fuel cladding integrity safety limits, MCPR operating limits, and MAPLIIGR limits are required. j

! An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design basis accident by increasing the blowdown area and eliminating the capability of reflooding the core. Thus, the requirement for shutdown of the facility with a jet pump inop( <ble.

In order to prevent urEae stress on the vessel nozzles and bottom head region, the recirculation .oop temperatures should be within 50*F of each other prior to startup of an idle loop.

Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference were greater than 145'F.

3/4.4.2 SAFETY / RELIEF VALVES The reactor coolant system safety valve function of the safety-relief valves operate to prevent the system from being pressurized above the Safety Limit of 1325 psig. The systen is designed to meet the requirements of the ASME Boiler and Pressure Vessel Code Section III for the pressure vessel and ANSI B31.1, 1967, code for the reactor coolant system piping.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 374.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are i provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection System."

b BRW4SWICK - UNIT 1 B 3/4 4-1 Amendment No.

1 l

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Low Flow) 2.1.1 THERMAL ."0WER 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 exceeding 25% of RATED THERMAL 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

THERMAL POWER (High Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 for

+

two recirculation loop operation and less than 1.08 for one recirculation loop operation with the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow.

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.07 for two recirculation loop operation or less than 1.08 for one recirculation loop operation and the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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.

l l APPLICABILITY: CONDITIONS 1, 2, 3, and 4.

l 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 ;( 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l BRUNSWICK - UNIT 2 2-1 Amendment No.

l _ _

en

@ TABLE 2.2.1-1 Si

$ REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS O

ALLOWABLE

'@ FUNCTIONAL UNIT AND INSTRUMENT NUMBER TRIP SETPOINT VALUES ,

s n 1. Intermediate Range Monitor, Neutron Flux - High(I) < 120 divisions of full scale < 120 divisions (C51-IRM-K601A,B,C,D,E,P,G,H) of full scale

2. Average Power Range Monitor (C51-APRM-CH.A,B,C,D,E,F)

a. Neutron Flux - High, 15%(2) i 15% of RATED THERMAL POWER 1 15% of RATED THERMAL POWER

b. Flow-Biased Neutron Flux - High(3)(4)(8) 1 (0.66 W + 54%) - 0.66aW < (0.66 W + 54%) - 0.66aW w c. Fixed Neutron Flux - High(4) < 120% of RATED THERMAL POWER 1 120% of RATED E THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High 1 1045 psig i 1045 psig (B21-PS-N023A,B,C,D)

4. Reactor Vessel Water Level - Low, Level' #I(7) _>+ 162.5 inches _

>+ 162.5 inches (B21-LIS-N017A,B,C,D)

5. Main Steam Line Isolation Valve - Closure (5) ~

< 10% closed < 10% closed (B21-F022A,B,C,D; B21-F028A,B,C,D)

6. Main Steam Line Radiation - High 1 3 x full power background 1 3.5 x full power (D12-RM-K603A,B,C,D) background E. 7. Drywell Pressure - High 1 2 psig < 2 psig

R (C72-PS-N002A,B,C,D) -

5 .

2 8. Scram Discharge Volume Water Level - High 1 109 gallons i 109 gallons 0

(C12-LSH-N013A,B,C,D)

TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TABLE NOTATION (1) The Intermediate Range Monitor scram functions are automaticall'y bypassed when the reactor mode switch is placed in the Run position and the Average .

Power Range Monitors are on scale.

(2) This Average Power Range Monitor scram function is a fixed point and is increased when the reactor mode switch is placed in the Run position.

(3) The Average Power Range Monitor scram function is varied, Figure 2.2.1-1, as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.

(4) Tha APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly

to instantaneous neutron flux. T

,j' (5) The Main Steam Line Isolation valve-closure scram function is automatically bypassed when the reactor mode switch is in other than the ,

Run position. <

(6) These scram functions are bypassed when THERMAL POWER is less than 30% of RATED THERMML POWER. / .'!

(7) Vessel water levels refer to REFERENCE LEVEL ZERO. .

t (8) W = Loop recirculation flow rate in percent of rated core flow.

AW = Difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting (-0.66AW) is accomplished by correcting the -

flow input of the flow biased scram to prevent the original (two loop) relationship between APRM scram setpoint and recirculation drive flow or by adjusting the APRM flux trip settin".

AW = 0 for two loop operation.

t BRUNSWICK - UNIT 2 2-6 Amendment No. - [i s

- ,. 4

3' 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 4

LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGR's) f6r each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the-following limits:

a. During two recirculation loop operation, the limits are 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, 3.2.1-7, 3.2.1-8, or 3.2.1-9.

b. During single recirculation loop operation, the limits shown in the above figures shall be multiplied by the following reduction factors:

Fuel Type Reduction Factor

, . 7x7 0.84 n' .' 8x8 0.85 8x8R 0.84 P8x8R 0.84 APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.

I

-* ACTION: With an APLHGR exceeding the limits of Figuras 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, 3.2.1-7, 3.2.1-8, or 3.2.1-9, initiate corrective action within 15 minutes and continue corrective action so that APLHCR is within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCEREQUIRENENTS

, 4.2.1 All APLHGR's shall be verified to be equal to or less than the l applicable limit determined from 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, 3.2.1-7, 3.2.1-8 or 3.2.1-9: ,

l I l <

l

a. At least once pert 242 hours0.0028 days 0.0672 hours 4.001323e-4 weeks 9.2081e-5 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter' completion of a THERMAL POWER increase of at least 15% of RATED TIIERMAL POWER, and j t ,. f c. Initially and at 'least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is l r operating with a LIMITING CONTROL R0D PATTERN for APLHGR.

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l l i' l

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

r

PCWER 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 setpoint (SRB) shall be established according to the following relationships:

S j( (0.66W + 54% -0.66AW) T SRB f_ (0.66W + 42% -0.66AW) T I where: S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow in percent of rated flow, AW = Difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting (-0.66AW) is accomplished by correcting the flow input of the flow biased scram to prevent the original (two loop) relationship between APRM scram setpoint and recirculation drive flow or by adjusting the APRM flux trip setting.

AW = 0 for two loop operation.

T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T f_1.0), and Design TPF for: P8 X 8R fuel = 2.39 8 X 8R fuel = 2.39 7X7 fuel = 2.60 8X8 fuel = 2.43 APPLICABILITY: CONDITION 1, when THERMAL POWER )_ 25% of RATED THERMAL POWER.

ACTION:

With S or SRB exceeding the allowable value, initiate corrective action within

. 15 minutes and continue corrective action so that S and SRB are within the l required limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of i

RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l 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:

l

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

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

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

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BRUNSWICK - UNIT 2 3/4 2-10 Amendment No.

l l

t

POWER DISTRIBUTION LIMITS

~

3/4.2.3 MINIMUM CRITICAL POWER RATIO

_ LIMITING CONDITION FOR OPERATION a.2.3.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shall be equal to or greater than the MCPR limit times the Kg shown in Figure 3.2.3-1, provided that the end-of-cycle recirculation pump trip system is OPERABLE per Specification 3.3.6.2, with:

a. If ODYN OPTION A analyses are in effect, the MCPR limits are listed below:

1. MCPR for 7x7 fuel = 1.21* **

2. MCPR for 8x8 fuel = 1.29**

3. MCPR for 8x8R fuel = 1.26** and,

4. MCPR for P8x8R fuel = 1.27**

b. If ODYN OPTION B analyses are in effect (refer to Specification 3.2.3.2), the MCPR limits are listed below:

1. MCPR for 7x7 fuel = 1.20* **

2. MCPR for 8x8 fuel = 1.29**

3. MCPR for 8x8R fuel = 1.21** and,

4. MCPR for P8x8R fuel = 1.22**

APPLICABILITY: CONDITION 1 when THERMAL POWER 125% of RATED THERMAL POWER ACTION:

a. With the end-of-cycle recirculation pump trip system INOPERABLE per Specification 3.3.6.2, operation may continue and the provisions of Specification 3.0.4 are not applicable with the following MCPR limit

. adjustments:

1

1. Beginning-of-cycle (BOC) to end-of-cycle (E0C) minus 2000 MRD/t, j within one hour determine that MCPR, as a function of core flow, is l equal to or greater than the MCPR limit times the Kg shown in Figure t

3.2.3-1 with:

a. If ODYN OPTION A analyses are in effect, the MCPR limits are listed below:

1. MCPR for 7x7 fuel = 1.20* **

2. MCPR for 8x8 fuel = 1.29**

3. MCPR for 8x8R fuel = 1.25** and,

, 4. MCPR for P8x8R fuel = 1.26** '

! b. If ODYN OPTION B analyses are in effect (refer to Specification l 3.2.3.2), the MCPR limits are listed below:

l

1. MCPR for 7x7 fuel = 1.20* **

2. MCPR for 8x8 fuel = 1.29**

l 3. MCPR for 8x8R fuel = 1.22** and, l 4. MCPR for P8x8R fuel = 1.25**

BRUNSWICK - UNIT 2 3/4 2-11 Amendment No.

I

(

I

2. EOC minus 2000 MWD /t to EOC, within one hour determine that MCPR, as a function of core flow, is equal to or greater than the MCPR. limit ,

times the Kg shown in Figure 3.2.3-1 with:

a. If ODYN OPTION A analyses are in effect, the MCPR limits are listed below:

1. MCPR for 7x7 fuel = 1.29* ** -

1. 2. MCPR for 8x8 fuel = 1.36**

3. MCPR for 8x8R fuel = 1.37** and,

4. MCPR for P8x8R fuel = 1.40**

b. If ODYN OPTION B analyses are in effect (refer to Specification 3.2.3.2), the MCPR limits are listed below:

1. MCPR for 7x7 fuel = 1.20* **

2. MCPR for 8x8 fuel = 1.29,**

3. MCPR for 8x8R fuel = 1.25** and,

4. MCPR for P8x8R fuel = 1.28**

b. With MCPR, as a function of core flow, less than the applicable limit determined from Figure 3.2.3-1 initiate corrective action within 15 minutes and restore MCPR to within the applicable limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.3.1 MCPR, as a function of core flow, shall be determined to be equal to or greater than the applicable limit determined from Figure 3.2.3-1:

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

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

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

s

For 7x7 fuel assemblies, the Kg factor is based on the 112% flow curve of l Figure 3.2.3-1 rather than the actual setpoint of 102.5%.

- For single recirculation loop operation, these limits shall be increased by 0.01.

BRUNSWICK - UNIT 2 3/4 2-11A Amendment No.

$ TABLE 3.3.4-2 e

y CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS s

M TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE B

@ 1. APRM (CSI-APRM-CH.A,B,C,D,E,F) 3 m a. Upscale (Flow Biased) < (0.66 W + 42% - 0.66AW) T* < (0.66 W + 42% - 0.66AW) T*

b. Inoperative NA MTPF NA MTPF

c. Downscale > 3/125 of full scale -> 3/125 of full scale

d. Upscale (Fixed) -< 12% of RATED THERMAL POWER < 12% of RATED THERMAL POWER

2. ROD BLOCK MONITOR (C51-RBM-CH.A,B)

a. Upscale < (0.66W + 39% - 0.66AW) T* < (0.66 W + 39% - 0.66AW) T*

b. Inoperative NA MTPF NA MTPF w c. Downscale > 3/125 of full scale > 3/125 of full scale 3

w 3. SOURCE RANGE MONITORS (C51-SRM-K600A,B,C D) b N

a. Detector not full in NA NA

b. Upscale < 1 x 105 cps < 1 x 10 5 cp,

c. Inoperative NA NA

d. Downscale > 3 eps > 3 cps 4 INTERMEDIATE RANGE MONITORS (C51-IRM-K601A,B,C,D,E F,G,H) .

a. Detector not full in NA NA

b. Upscale < 108/125 of full scale < 108/125 of full scale

c. Inoperative NA NA

d. Downscale > 3/125 of full scale > 3/125 of full scale E.

5 .

F

T=2.60 for 7 x 7 fuel.

T=2.43 for 8 x 8 fuel.

T=2.39 for 8 x 8R fuel.

T=2.39 for P8 x 8R fuel.

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM -

RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION -

3.4.1.1 Two reactor coolant recirculation loops shall be in operation with the cross-tie valve closed, the pump discharge valves OPERABLE and the pump discharge bypass valves OPERABLE or closed, except as permitted under single loop operation.

APPLICABILITY: CONDITIONS 1* and 2*.

ACTION:

With both recirculation loops not in operation, operation may continue; restore one loop to operation within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With only a single recirculation loop in service, plant operation may continue with the implementation of the requirements in Specifications 2.1.2, 2.2.1, 3.2.1, 3.2.2, 3.2.3.1, 3.3.4, and 3.4.1.2. These requirements shall be in effect within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the removal of one recirculation loop from service, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

t SURVEILLANCE REQUIREMENTS I

4.4.1.1 Each pump discharge valve and bypass valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each COLD SHUTDOWN which exceeds 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if not performed in the previous 31 days.

4.4.1.2 Each pump discharge bypass valve, if not OPERABLE, shall be verified to be closed at least once per 31 days.

l l

See Special Test Exception 3.10.4.

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

f

REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps associated with an OPERABLE recirculation loop shall be l OPERABLE.

APPLICABILITY: CONDITIONS 1 and 2 ACTION: With less than 10 jet pumps OPERABLE in an OPERABLE recirculation -

loop, be in at least hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.1.2 Each of the above required jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and .at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that all of the following conditions do not occur simultaneously. .

a. The recirculation pump flow diffeto by more than 10% from the established speed-flow characteristics,

b. The indicated total core flow differs by more than 10% from the core flow value derived from established power-core flow relationships, and

c. The diffuser-to-lower plenum differential pressure reading on any i individual jet pump varies from the mean of all jet pump differential pressures, in that loop, by more than 10%. .

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

SPtCIAL TEST EXCEPTION RECIRCULATION LOOPS .- ,.

LIMITING CONDITION FOR OPERATION 3.10.4 The requirement of Specification 3.4.1.1 that one or two recirculation loops be in operation may be suspended for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during the performance of start-up and PHYSICS TESTS.

APPLICAPILITY: CONDITIONS 1 and 2.

ACTION:

With the above specified time limit exceeded, deenergize the scram solenoid valves.

SURVEIIIANCE REQUIRE %"NTS 4.10.4 Tha time during which the above specified requirement has been suspended shall be verified to be less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at least once per hour during start-up and PHYSICS TESTS.

BRUNSWICK ~ UNIT 2 3/4 10-4 Amendment No.

SAFETY LIMITS BASES (Continued) 2.1.2 THERMAI POWER Uligh Pressure and High Flow) ,

The fuel cladding integrit'y Safe'ty Limit is set such that no fuel' damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power, result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB 1 , which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L),

(GEXL), correlation.

The GEXL correlation is valid over the range of conditions used in the l tests of the data used to develop the correlation. These conditions are:

Pressure: 800 to 1400 psia Mass Flux: 0.1 o 1.25 x 106 lb/hr-f t 2 Inlet Subcooling: 0 to 100 Btu /lb Local Peaking: 1.61 at a corner rod to 1.47 at an interior rod The analyses used for Safety Limit MCPR calculation are used for both one and two loop recirculation loop operation. For single loop operation, these analyses are reviewed and found to have additional uncertainties which will

! require the MCPR safety limit to be increased by 0.01.2 l

Reference l

1

" General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application," NEDO-10958 and NEDE-10958.

2

" Brunswick Steam Electric Plant Units 1 and 2 Single-Loop Operation,"

NEDO-24344 BRUNSWICK - UNIT 2 B 2-2 Amendment No.

l

T 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one recirculation loop in service is permitted. In such ins tances , the designated adjustments for APRM rod block and scram setpoints, RBM setpoints, MCPR fuel cladding integrity safety limits, MCPR operating limits, and MAPLHGR limits are required. ,

An inoperable jet pump is not, in itself, a suf'ficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design basis accident by increasing the blowdown area and eliminating the capability of reflooding the core. Thus, the requirement for shutdown of the.

facility with a jet pump inoperable.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures should be within 50*F of each other prior to startup of an idle loop.

Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference were greater than 145'F.

3/4.4.2 SAFETY / RELIEF VALVES The reactor coolant system safety valve function of the safety-relief valves operate to prevent the system from being pressurized above the Safety Limit of 1325 psig. The system is designed to meet the requirements of' the ASME Boiler and Pressure Vessel Code Section III for the pressure vessel and ANSI B31.1, 1967, Code for the reactor coolant system piping.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE

_3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of i Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary leakage Detection System."

BRUNSWICK - UNIT 2 B 3/4 4-1 Amendment No, e

ML20053E375

Text

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