ML20062H067

Requests Revision to Tech Specs Per 10CFR50.90 & Re Cycle 3 reload.NEDO-24182 & Supporting Documentation Encl

ML20062H067
Person / Time
Site:	Brunswick
Issue date:	04/13/1979
From:	Utley E CAROLINA POWER & LIGHT CO.
To:	Ippolito T Office of Nuclear Reactor Regulation
Shared Package
ML20062H070	List: ML20062H067 ML20062H073
References
GD-79-1010, NUDOCS 7904170225
Download: ML20062H067 (23)

Text

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Carolina Power & Light Company April 13, 1979 j

FILE: NG-3514(B)

SERIAL: GD-79-1010 Office of Nuclear Reactor Regulation ATTENTION:

Mr. T. A. Ippolito, Chief Operating Reactors Branch No. 3 l

United States Nuclear Regulatory Commission i

Washington, D. C.

20555 BRUNSWICK STEAM ELEC IC PLANT, UNIT NO. 2 DOCKET N. 50-324 LICENSE

. DPR.

l FUEL CYCLE NO. 3 - RE AD LICENSING

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Dear Mr. Ippolito:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90 and Part 2.101, Carolina Power & Light Company hereby requests revisions to the Technical Specifications for its Brunswick Steam Electric Plant, Unit No. 2.

These revisions are ne,cessary to complete the second refueling of Unit No. 2 and begin Cycle 3,operatiot., and we are submitting them to the Staff as a supplement to our March 21, 1979, Reload Licensing Submittal. Changes to the Technical Specifications are indicated by a vertical line in the right-hand margins of the affected pages which are attached.

A revised safety limit MCPR of 1.07 has been used by General Electric (GE) in their NED0-24179 Revision 1 document and is, therefore, incorporated into our revised Technical Specification pages. The new graphs of MAPLHGR vs.

Planar Average Exposure were prepared using tabulated data from the NEro-24179 Revision 1 document.

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Forty (40) copies of GE's NEDO-24182, " Supplemental Reload Licensing Submittal for BSEP Unit 2 Reload 2," are attached to this letter. This is the non-RPT (Recirculation Pump Trip feature) analysis for Unit 2 Cycle. 3.

This analysis has been used to generate fallback operating MCPR limits if RPT becomes inoperable during Cycle 3.

These fallback values have been incorporated-into the attached technical specification changes, i

The Unit No. 2 refueling outage began on March 2,1979, and criticality is presently scheduled for May 1,1979. Due to the overlapping of outages for Units 1 and 2 and the scheduled Spring, 1979 outage for H. B. Robinson Unit 2, we wish to stress the importance that the Staf f review be completed and the appro-priate license amendments be issued in a time period that allows Unit No. 2 to return to power in accord with this schedule.

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,Mr. T. A. Ippolito, Chief April 13, 1979 l

i In accordance with 10CFR170.12(c), we have determined that this request constitutes a Class III amendment because it involves a single tech-nical issue. Our check for $4,000 is enclosed as payment for this amendment feG.

3 If your staff has any questions concerning the attached information, we will be glad to discuss them either by telephone or at a meeting with repre-sentatives of your staff.

1 Yours very truly,

[e E. E. Utley Senior Vice President Power Supply (m

a JAM /tl Enclosures Sworn to and subscribed before me this 13th day of April,1979 T d N,w1/tdAh Notary Public g giessests, My Cocnission Expires:

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Low Flow) 1

?.l.1 THEP. MAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 800 psia or core flow less than 105 of rated flow.

2 APPLF.A8ILITY: CONDITIONS 1 and 2.

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ACTION:

Y 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 witnin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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THEPMAL POWER (Hioh Pressure and High Flow) 2.l.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not b'e less than 1.07 with the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10" of rated flow.

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d APPLICABILITY: CONDITIONS 1 and 2.

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With MCPP less than 1.07 and the reactor vessel steam dome pressure l

i greater than 800 psia and core flow gre+ater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

't 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.

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

ij ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least ROT SHUTDOWN with reactor coolant system pressure i 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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l TABLE 2.2.1-1

. E REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2

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FUNCTIONAL UNIT AND INSTRUMENT NUMBER TRIP SETPOINT VALUES E

1 Intermediate Range Monitor, Neutron Flux - HighIII U

(C51-IRM-K601 A,B,C,0,E.F,G H)

~< 120 divisions of full scale

~< 120 divisions of full scale na 2

Average Power Range Monitor (C 51 -A PRM-C.H. A,B,C,0, E. F)

Neutron Flux - High, 15%(2) a.

1 15% of RATED THERMAL POWER

< 15% of RATED

_ THERMAL POWER b.

Flow Blased Neutron Flux - High(3)(4)

< (0.66 W + 54%)

1 (0.66 W + 54%)

tI Fixed Neutron Flux - High(4I

< 120% of RATED THERMAL POWER

< 120% of RATED i1 c.

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_ THERMAL POWER 3

Reactor Ves,sel Steam Dome Pressure - High 1 1045 psig

< 1045 psig (B21-PS-N023A,B,C.D)

1. I 4

Reactor Vessel Water Level - Low, Level

-> 12.5 inches above instru-12.5 inches above (821-LIS-N017A,B,C,0) ment zero

- instrument zero 5.

Main Steam Line Isolation. Valve - Closure (5)

(821-F022 A,B,C,0; B21-F028 A,B,C,D)

~< 10% closed

'- 10% closed t

6.

Main Steam Lt.ne Radiation - High

~< 3 x full power background

-- background 3.5 x full power (012-RM-K603 A.B.C.D) 7 Drywell Pressur.e - High

< 2 psig

~ 2 psig (C72-PS-N002 A,B,C D) l_

8 Scram Discharge. Volume Water Level - High

-< 109 gallons 109 gallons (c12-LSH-N013 A,B,C,D)

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2.0 The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive mat '

erials to the environs. Safety limits are established to protect the integrity of these barriers during normal plant operations and antici-pated transients. The fuel cladding integrity limit is set such that no fuel damage is calculated to occur if the limit is not violated. Be-cause fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MINIMUM CRITI, CAL POWER

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RATIO (MCPR) is no less than 1.07.

MCPR > 1.07 represents a conserva-

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tive margin relative to the conditions required to maintain fuel cladding

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integri ty.

The fuel cladding is one of the physical barriers which a

separate the radioactive materials from the environs.

The integrity of this cladding barrier is related to its relative freedom from perfor-

• d ations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this scurce is incrementally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions a'nd the Limiting Safety System Settings. While fission product migra-

%d tion from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a a

threshold, beyond which still greater thermal stresses may cause gross a

5 rather than incremental cladding deterioration.

Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which 3(d would produce onset of transition boiling, MCPR of 1.0.

These con-3 ditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER (Low Pressure or low Flow) 4 a...

The use of the GEXL correlation is not valid for all critical power "N

calculations at pressures below 800 psia or core flows less than 10% of rated flow. Therefore the fuel cladding integrity limit is established i

j by other means. This is done by establishing a limiting condition on core i

i THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will alwa s be greater than 4.5 psi. Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure I

drop is nearly independent of bundle power and has a value of 3'.5 psi.

,,j Thus, the bundle flow with a 4.5 psi driving head will be greater than

,i 28 x 103 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at s

this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 800 ps-ia is conservative.

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2.2 LIMITING SAFETY SYSTEM SETTINGS 1

BASES 3

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t 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the Reactor Trips are set

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for each parameter.

The Trip Setpoints have been selected to ensure that the raactor core and reactor coolant system are prevented from exceeding their safety limits.

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Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor J

trip systems. The IRM is a 5 decade 10 range instrument.

The trip set-point of 120 divisions is active in each of the 10 ranges.

Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. Range 10 allows the IRM instruments to 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 are due to control rod withdrawal.

In order to ensure that the j

IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR. The most severe case

, ----( s) and the IRM's are not yet on scale. Additional conservatism was taken involves an initial condition in which the reactor is just subcritical in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above 1.07.

Based on this analysis, the IRM provides protection against local control rod errors, and continuous 4

withdrawal of control rods in sequence and provides backup protection for the APRM.

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2.

Average Power Range Monitor t

For operation at low pressure and low flow during STARTUP, the APRM i

scram setting of 15% of RATED THERMAL POWER provides adequate thermal i

margin between the setpoint and the Safety Limits.

This margin sceom-

,.j modates the anticipated maneuvers associated with power plant startup.

P Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup, is not much colder than that already in the system, temperature coefficients are small and control rod patterns are constrained by the RSCS and RWM. Of all 1

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q 3.1.4.3 Both Rod Block Monitor (RBM) channels shall be OPERABLE.

i APPLICABILITY: CONDITION 1, when THERMAL POWER is greater than the l

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preset power level of the RWM and RSCS.

t ACTION:

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With one RBM channel inoperable, POWER OPERATION may continue i

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1.

The inoperable RBM channel is restored to OPERABLE. status l

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within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or i

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The redundant RBM is demonstrated OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

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and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable RBM is restored to OPERABLE status, and the inoperable RBM is l

restored to OPERABLE status within 7 days, or j

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THERMAL. POWER is limited such that MCPR will remain above

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,5j 1.07 assuming a single error that results in complete

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withdrawal of any single control rod that is capable of

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Otherwise, trip at least one rod bl'ack monitor channel.

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

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SURVEILLANCE REQUIREMENTS

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4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance ~of a CHANNEL FUNCTIONAL TEST and CHANNEL CALI-i ij BRATION at the frequencies and during the OPERATIONAL CONDITIONS specified IJ in Table 4.3.4-1.

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REACTIVITY CONTROL SYSTEMS

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BASES CONTROL RODS (Continued) i potential effects of the rod ejection accident are limited.

The ACTION statements permit variations from the basic requirements but at.the same i

time impose more restrictive criteria for continued operation. A limita-tion on inoperable rods is set such that the resultant effect on total

-i rod worth and scram shape will be kept to a minimum.

The requirements

25% of RATED THERMAL POWER.

ACTION:

With an AP GGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, or 3.2.1-7, initiate corrective action within 15 minutes and continue corrective action so that APLHGR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THEREU. POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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SURVEILLANCE REQUIREMENTS

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4.2.1 All APLHGR's shall be verified to be equal to or less than the

.J applicable limit determined from Figure 3. 2.1 -1,

3. 2.1 -2, 3.2.1 -3, 3.2.1-4, 3.2.1-5, 3.2.1-6, or 3.2.1-7:

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At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and

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Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

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LIMITING CONDITION FOR OPERATION i

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 relationships:

S 1 (0.66W + 54%) T S

1 (0.66W + 42%) T RB 1

where:

S and S are in percent of RATED THERMAL POWER, RB j

W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T 1 1.0), and Design TPF for:

7 x 7 fuel = 2.60 8 x 8 fuel = 2.45 8 x 8R fuel = 2.48 APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL V

POWER.

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With S or S exceeding the allowable value, initiate corrective action RB within 15 minutes and continue corrective action so that S and S are within the required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWEk to I

less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

i

)

1 SURVEILLANCE REQUIREMENTS t

i Q

4.2.2~ The MTPF for each class of fuel shall be determined, the value

==

j of T calculated, and the flow biased APRM trip setpoint adjusted, as

>j required:

a.

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

>i b.

Whenever THERMAL POWER has been increased by.at leas't 15% of

' 1 RATED THERMAL POWER and steady state operating conditions have been established, and em c.

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

i BRUNSWICK - UNIT 2 3/4 2-7

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i POWER DISTRIBUTION LIMITS l

3/4.2.3 MINIMUM CRITICAL POWER RATIO i

LIMITING CONDITION FOR OPERATION 4

3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shall be equal to or greater than MCPR x the Kg shown in Figure 3.2.3-1 where t

,-S, MCPR = 1.20 for 7x7 fuel' I

j MCPR = 1.20 for 8x8 fuel MCPR = 1.26 for 8x8R fuel

?

---ad APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% RATED THERMAL POWER ACTION:

With MCPR less than the applicable limit determined from 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER 1

within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REOUIREMENTS

~

k'/

4 4.2.3 MCPR 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, d

b.

Whenever THERMAL POWER has been increased by at least 15%

of RATED THERMAL POWER and steady state operating conditions c,

have been established, and c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R0D PATTERN for MCPR.

.~

• For 7x7 fuel assemblies, the K factor is based on the 112% flow

'IN g

curve of Figure 3.2.3-1 rather than the actual setpoint of 102.5%.

rmen l

BRUNSWICK - UNIT 2 3/42-8

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' '. aW 3/4.2 p0WER DISTRIBUTION LIMITS t

BASES 4

i The specifications of this section assure that the peak cladding.

c temperature following the postulated design basis loss-of-coolant ac'cident 4

will not exceed the 2200*F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effects of fuel pellet densification.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE I

i 1

./

This specification assures that the peak cladding tempeature t

following the postulated design basis loss-of-coolant accident will q

not exceed the limit specified in 10 CFR 50, Appendix K.

l The peak cladding temperature (PCT) following a postulated loss-of-coolant accident 1s primarily a function of the average heat genera-tion rate of all the rods of a fuel assembly at any axial location and i,s dependent only secondarily on the rod to rod power distribution within a assembly. The peak clad temperature is calculated assuming a V

LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used

'j in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Spect-Fication ApHGR is this LHGR of the highest powered rod divided by its i

.)

local peaking factor. The limiting value for APLHGR is shown in

(

V l

Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1 6, and 3.2.1-7.

The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, and 3.2.1-7 is [

based on a loas-of-coolant. ace m e analysis The analysis was performed y*e using General Electric (GE) cgiculational models which are consistent with 2

the requirements of Appendix K to 10 CFR 50.

A complete discussion of each l

code employed in the analysis is presented in Reference 1.

Differences in this analysis compared to previous analyses performed with Reference 1 are:

(1) The analysis assumes a fuel assembly planar power consistent with 102%

of the >fAPLHGR shown in Figuren 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, and 3.2.1-7; (2) Fission product decay is computed assuming an

(;

energy release rate of 200 MEV/ Fission; (3) Pool boiling is assumed after j

nucleate boiling is lost during the flow stagnation period; (4) The effects i

. C-]

of core spray entrainment and counter-current flow limitation as described in Referenc_a 2, are included in the reflooding calculations.

3pm i

A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.

BRUNSWICK - UNIT 2 83/42-1

Bases Table B 3.2.1-1 i

SIGNIFICANT INPUT PARAMETERS TO THE

}

l l

LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR BRUNSWICK UNIT 2 i

I Plant Parameters; Core l hermal Power.................... 2531 MWe, which corresponds io 105% of rated steam flow

• l Vess el Steam Output.............. 10.96 x 100 lbm/h which corresponds to 105% of rated steam flow

)

Vessel Steam Dome Pressure......... 1055 psia Recirculation Line Break Area for Large Breaks - Discharge 2.4 ft (DSA) 1.9 ft (80% DBA)

]

- Suction i

4.2 ft2 Number of Drilled Bundles 520

_ Fuel Parameters PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAR AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPES GEOMETRY (kw/ f t)

FACTOR RATI0'*

{

~~

t

'aload Core 8x8 13.4 1.4 1.20 l

\\.

/

7x7 18.5 1.5 1.20 4

t i

A more detailed list of input to each model and its source is presented in j

Section II of Reference 1.

.}

• This power level meets the Appendix K requirement of 102%.

• • To account SCAT calcul'ation is performed with an MCPR of 1.18 (i.e.,for the i

for a bundle with an initial MCPR of 1.20.

1.2 divided by 1.02) 4 i

l BRUNSWICK - UNIT 2 l'

B 3/4 2-2 I

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.%*d POWER DISTRIBUTION LIMITS 7

BASES i

3/4.2.2 APRM SETPOINTS The fuel cladding integrity safety limits of Specification 2.1 were based on a TOTAL PEAKING FACTOR of 2.60 for 7 x 7 fuel, 2.45 for 8 x 8 fuel, and 2.48 for 8 x 8R fuel. The scram setting and rod block functions of the

,A APRM instruments must be adjusted to ensure that the MCPR does not become

(,j less than 1.0 in the degraded situatione The scram settings and rod block settings are adjusted in accordance with the formula in this specification

.a q

when the combination of THERMAL POWER and peak flux indicates a TOTAL PEAKING aams FACTOR greater than 2.60 for 7 x 7 fuel, 2.45 for 8 x 8 fuel and 2.48 for 8 x 8R fuel. The method used to determine the design TPF shall be consistent 4

.N with the method used to determine the MTPF.

7 3/4.2.3 MINIMUM CRITICAL POWER RATIO I

The required operating limit MCPR's at steady state operating conditions as specified in Specification 3.2.3 are derived from the established

V fuel cladding integrity Safety L t MCPR of 1.07, and an analysis of l

m abnormal operational transients.

For any abnormal operating tran-sient analysis evaluation with the initial condition of the reactor i

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 instrument trip setting as given in

- x

()

Specification 2.2.1.

5 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).

The type of transients evaluated r

7, were loss of flow, increase in pressure and power, positive reactivity l

insertion, and coolant temperature decrease.

The limiting transient which determines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass.

This transient yields the largest a MCPR.

When, added to the Safety Limit MCPR of 1.07 the required minimum operating limit MCPR of Specification

[

3.2.3 is obtair.ed.

Prior to the analysis of abnormal operational tran-

,9 sients an initial fuel bundle MCPR was determined. This parameter is j

based on the bundle flow calculated by a GE multi-channel steady te

.g flow distribution model as described in Section 4.4 of NED0-20360 and i

on core parameters shown in Reference 3, response to Items 2 and 9.

BRUNSWICK - UNIT 2 B 3/4 2-3

P0'WER DISTRIBUTION L1MITS BASES' MINIMUM CRITICAL POWER RATIO (Continued)

}

For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that j

the MCPR was equal to the operating limit MCPR at rated power and flow.

The Kg factors shown in Figure 3.2.3-1 are conservative for the GE plant operation with 8X8 and 8X8R fuel assemblies because the operating limit r

MCPRs of Specification 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of Kg.

The Kg curves are coc-servative for 7X7 fuel whenever the operating limit MCPR is greater than 1.23 as documented in Appendix C of NEDE 24011-P-A. A correction to the Kg curves is, therefore, necessary whenever the MCPR for the 7X7 fuel is

/

equal to or less than 1.23 in order to ensure that the fuel cladding integrity safety limit is not violated. This correction is made by using a scoop tube set point of 102.5%. The MCPR for 7X7 fuel is then the pro-t duct of the value given in Specification 3.2.3 and the Kg curve based on 112% as shown in Figure 3.2.3-1.

Whenever the MCPR for the 7X7 fuel is _.

greater than 1.23, this correction is not applied.

i At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indicated f

that the resulting MCPR value is in excess of requirements by a considera-ble margin.

With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with mimimum recirculation pump

( )i speed.

The MCPR margin will thus be demonstrated such that future MCPR l

evaluation below this power level will be shown to be unnecessary.

The daily requirement for. calculating MCPR above 25% rated themal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.

f 3.2.4 LINEAR HEAT GENERATION RATE The LHGR specification assures that the linear heat generation rate l

in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

The power spike penalty specified is based on the analysis presented in Section 3.2.1 of the GE topical report NEDM-10735 Supplement 6, and assumes a linearly increasing variation in axial gaps between core bottem and top, and assures with a 95% confidence that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking.

j BRUNSWICK - UNIT 2 B 3/4 2-5

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+ + =. - -.. - - - = ~ -

l l

i POWER OISTRIBUTION LIMITS BASES 1.

General Electric Company Analytical Model for Loss-of-Coolant p

Analysis in Accordance with 10 CFR 50, Appendix K NED0-20566, January, 1976.

2.

General Electric Refill Reflood Calculation (Supplement to SAFE Code-Description) transmitted to USAEC by letter, G. L. Gyorey to V. Stello, Jr., dated December 20, 1974.

3.

Letter from J. A. Jones, Carolina Power and Light Company to B. C. Rusche, NRC transmitting Amendment 31 to the Brunswick

( ;,

Unit 1 Docket No. 50-325, dated November 26, 1975.

4.

General Electric BWR Generic Reload Application for 8 x 8 Fuel, NED0-20360, Revision 1, November 1974.

5.

R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NED0-10802).

/*

6.

Letter from J. A. Jones Carolina Power and Light Company, to B. C. Rusche, NRC dated May 7, 1976.

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_ INSTRUMENTATION v

t 4

..v3 3/4.3.4 CONTROL R00 WITHDRAWAL. BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION f

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 i

with the values shown in the Trip Setpoint column of Table 3.3.4-2.

APPLICABILITY,: As shown in Table 3.3.4-1.

ACTION:

f)

With a l control rod withdrawal block instrumentation ~ channel trip a.

setpoint less conservative than the value shown in the Allowable i

Values column of Table 3.3.4-2, declare the channel inoperable until the channel is restored to OPERABLE status with its Trip 9

Setpoint adjusted consistent with the Trip Setpoint value_.

b.

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

1.

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

- m 2.

The redundant trip system is demonstrated OPERABLE within

+

, c

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable 1

l channel is restored to OPERABLE status, and the inoperable

( )\\

channel is restored to OPERABLE status within 7 days, or l

x 3.

For the Rod Block Monitor bnly, THERMAL POWER is limited i

such that MCPR will remain above 1.07 assuming a single error that results in complete withdrawal of any single control rod that is capable of withdrawal.

4 4

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

. r :==

With the requirements for the minimum number of OPERABLE c.

channels not satisfied for both trip systems, place at least j

one trip system in the tripped condition within one hour.

t d.

The provisions of Specification 3.0.3 are not applicable in

j OPERATIONAL CONDITION 5.

?

SURVEILLANCE REQUIREMENTS l

M 4.3.4 Each of the above required control rod withdrawal block instrumen-ation channels shall be demonstrated OPERABLE by the performance of a t

lHANNEL CHECK, CHANNEL CALIBRATION and a CHANNEL FUNCTIONA the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4 ng BRUNSWICX-UNIT 2

_ 3/4 3_39_ _ _ _ _.

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TABLE 3.3.4-2

=

B CONTROL R0D WITilDRAWAL BLOCK INSTRUMENTATION SETPOINTS j

y 4

'l 5

l Q

TRIP FUtlCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE l--

?

2 1.

APRM (C51-APRM-Cll.A,B,C,D,E,F) a.

Upscale (Flow Biased)

< (0.66 W + 42%)

T*

5 (0.66 W + 42%) T*

b.

Inopera tive fiA MTPF NA MIPF c.

Downscale

> 3/125 of full scale

> 3/125 of full scale d.

Upscale (Fixed) i 12% of RATED TilERMAL POWER 512% of RATED TilERMAL POWER 2

R0D BLOCK MONITOR (CSI-RBM-Cil.A,B) a.

Upscale

< (0.66W + 39%)

T*

< (0.66 W + 39%)

T*

l b.

Inopera tive IIA I4TPF fiA MIPF c.

Downscale

> 3/125 of full scale

> 3/125 of full scale M

~

3.

SOURCE RANGE MONITORS (C51-SRM-K600A,8,C.D) i Y'

j e

a.

Detector not full in NA NA

~

5 5

b.

Upscale 1 1 x 10 cps 5 1 x 10 cps c.

Inoperative NA NA d.

Downscale

> 3 cps

> 3 cps 4.

INTERMLDIATE RANGE MONITORS (C51-IRM-K601A B.C D.E.F,G,II) a.

Detector not full in NA NA b.

Upscale 1 108/125 of full scale 5 108/125 of full scale c.

Inopera tive NA NA

?

d.

Downscale

> 3/125 of full scale

> 3/125 of full scale 1

/

T=2.60 for 7 x 7 fuel.

I T=2.45 for 8 x 8 fuel.

T=2.48 for 8 x 8R fuel.

i i

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TABLE 4.3.4-1 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CifANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH g

TRIP FUNCTION AND INSTRUMENT NUMBER CHECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED

$i 1.

APRM (CSI-APRM-CH.A,B,C,D,E,F) k S/Ufc)M R(b) j a.

Upscale (Flow Blased)

NA i

c),Q NA 1, 2, S Y

b.

Inoperative NA S/U E

c.

Downscale NA S/U

,M NA 1

7 d.

Upscale (Fixed).

NA S/U

,Q R

2, 5 l

j 2.

R0D BLOCK MONITOR (CSI-RBM-CH.A,B) a.

Upscale NA S/U(c) M R

1*

b.

Inopera tive NA S/U

,Q NA 1*

c.

Downscale NA S/U

,M R

1*

j 3.

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

Detector not full in NA S/U

,W NA 2, 5 b.

Upscale NA S/U W

NA 2, 5 S/U((c),W c.

Inoperative NA NA 2, 5 y

d.

Downscale NA S/U c),W NA 2, 5 4.

INTERMEDI ATE RANGE MONITORS (C51-IRM-K601 A,B,C D.E,F,G,H) a.

Detector not full in NA S/U(c) y(d)

NA 2

=l NA W

NA 5

I b.

Upscale NA S/U(c)9(d)

NA 2

NA W

NA 5

c.

Inoperative NA S/U(c) 9(d)

NA 2

NA W

NA 5

d.

Downscale NA S/U(c) y(d)

NA 2

NA W

NA 5

a.

CHANNEL CALIBRATIONS are electronic.

b.

This calibration shall consist of the adjustment of the APRM flow biased setpoint to conform to a calibrated flow signal.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.

I c.

! j' When changing from CONDITION 1 to CONDITION 2, perform the required surveillance d.

l within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering CONDITION 2.

When TitERMAL POWER is greater than the preset power level of the RWM and RSCS.

l l

i i

INSTRUMENTATION l

i j

3/4.3.6 RECIRCULATION PUMP TRIP LOGIC l

l j

LIMITING CONDITION FOR OPERATION I

(

f f

l 3.3.6 The recirculation pump trip (RPT) system consists of two independent systems, either of which will trip the recirculation pumps on turbine control valve fast

{

closure or turbine stop valve closure.

Both trip systems must be operable at all l

.j times in their applicable conditions.

t

]

APPLICABILITY: Condition 1 with thermal power > 30%.

1 1

ACTION:

^

l h')

If the test period for one trip system exceeds two hours the trip a.

system is inoperable and ACTION b. applies.

b.

With one RPI system inoperable for >72 consecutive hours or with l

both systems inoperable operation may continue with an operating MCPR limit as follows:

[

i For 7 x 7 fuel MCPR shall be > l.21 For 8 x 8 and 8 x 8R fuel MCFR shall be > 1.27 SURVEILLANCE REOUIREMENTS 6

l.

4.3.6.1 Each recirculation pump trip system logic shall be demonstrated operable at least once per 31 days by the performance of logic system functional tests ex-j

(

cept that the breakers will not be tripped. Each system may be placed in test for p

two hours to perform monthly functional testing.

(

\\

i l

4.3.6.2 The recirculation pump trip system breakers will be tested during each i

refueling outage.

I I

BRUNSWICK - UNIT 2 a

I 3/4 3-62 4

i l

l l'

5.0' DESIGN FEATURES l

5.1 SITE EXCLUSION AREA d

I]

5.1.1 The exclusion area shall be as :hown in Figure 5.1.1-1.

g LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1, based on the information given in Section 2.2 of the FSAR.

l' )

i

(

. /

5.2 CONTAINMENT 1

i CONFIGURATION 5.2.1 The PRIMARY CONTAINMENT is a steel lined reinforced concrete structure composed of a series of vertical right cyltnders and truncated cones which form a drywell. This drywell ts attached to a suppression chamber through a series of vents. The suppression chamber is a con-crete steel lined pressure vessel in the shape of a torus. The primary containment has a minimum free air volume of (288,000) cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is desfgned and shall be maintained for:

i es

(

i a.

t_

Maximum internal pressure 62 psig.

i b.

Maximum internal temperature:

drywell 300*F.

i suppression chamber 200*F.

c.

Maximum external pressure 2 psig.

5.3 REACTOR CORE 1

FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 560 fuel assemblies. The 7X7 fuel assemblies shall contain 49 fuel rods; the 8X8 fuel assemblies shall con-tain 63. fuel rods; and the 8X8R fuel assemblies shall contain 62 fuel rods.

All fuel rods shall be clad with Zircaloy 2.

The nominal active fuel length of each fuel rod shall be 144 inches for 7X7 fuel assemblies, 146 inches for 8X8 fuel assemblies, and 150 inches for 8X8R fuel assemblies.

Each fuel rod shall contain a maximum total weight of 4,430 grams of UO '

2 BRUNTJICK - UNIT 2 5-1

- - - - -