ML19261A558

Requests Tech Spec Revisions Necessary to Complete First Refueling & Begin Cycle 2 Operation.Revisions Submitted as Suppl to 781229 Reload Licensing Submittal.W/Encl Revisions & Class III Amend

ML19261A558
Person / Time
Site:	Brunswick
Issue date:	01/17/1979
From:	Utley E CAROLINA POWER & LIGHT CO.
To:	Ippolito T Office of Nuclear Reactor Regulation
References
GD-79-166, NUDOCS 7901260199
Download: ML19261A558 (19)

Text

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Carolina Power & Light Companj

. January 17, 1979 FILE: NG-3514(B) SERIAL: GD-79-166 Office of Nuclear Reactor Regulation ATTENTION: Mr. T. A. Ippolito, Chief Operating Reactors Branch No. 3 United States Nuclear Regulatory Commission Washington, D. C. 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 DOCKET No. 50-325 LICENSE NO. DPR-71 FUEL CYCLE No. 2 - RELOAD LICENSING

Dear Mr. Ippolito:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90 and PIrt 2.101, Carolina Power & Light Company hereby requests revisions to the Technical Specifications for its Brunswick Steam Electric Plant, Unit No. 1. These revisions are necessary to complete the first refueling of Unit No. 1 and begin Cycle 2 operation, and we are submitting them to the Staff as a supplement to our December 29, 1978, 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 NED0-24166 document and is, therefore, incorporated into our revised Technical Specification pages. The new and the revised graphs of MAPLHGR vs. Planar Average Exposure were prepared using tabulated data from GE's NEDO-24165 document.

The Unit No. 1 vfueling outage began on January 13, 1979, and ascent to power is presently schat ad to begin on March 25, 1979. Due to the over-lapping of outages for Unite 1 and 2, we wish to stress the importance that the Staff review be completed and the appropriate license amendments be issued in a time period that allows Unit No. 1 to return to power in accord with this schedule.

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

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IllSTRUMENTATION 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMEilTATION LIMITING CONDITION FOR OPERATIO*1 3.3.1 The control rod withdrawal block instrumentation shown in Table 3.3.4-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4-2.

APPLICABILITY: As shown in Table 3.3.4-1.

ACTION:

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

2. The redundant trip system is demonstrated OPERABLE within 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 channel is restored to OPERABLE status, and the inoperable channel is restored to OPERABLE status within 7 days, or

3. For the Rod Block Monitor only, THERMAL POWER is limited such that MCPR will remain above 1.07 assuming a single l error that results in complete withdrawal of any single control rod that is capable of withdrawal.

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

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

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

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

BRUNSWICK-UNIT 1 3/4 3-39

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES .

2.2.1 REACTOR PROTECTION SYSTEM IflSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints speci-fied in Table 2.2.1-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolr.nt system are prevented from exceeding their safety limits,

l. Intermediate Range Monitor, heutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip set-ooint of 120 divisions is active in each of the 10 ranges. Thus as the IRM is ranged uo 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 IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM's are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the rod being wi thdrawn i r. bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above 1.07. Based on this. analysis, the l IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15" of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. This margin accom-modates the anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup, is not much colder than that already in the system, temperature coefficients are small and control rod patterns are constrained by the RSCS and RWM. Of all BRUNSWICK-UNIT 1 B 2-9

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

S < (0.66W + 54%) T S

RB 1 (0.66W + 42%) T wnere: S and 5 are in percent of RATED THERMAL POWER, W = Loch recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T 1 1.0),and Design TPF for 8 x 8 fuel = 2.45 Design TPF for 8 x 8R fuel = 2.48 APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.

ACTION:

With 5 or S RA exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and S oB 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 POWER to 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 />.

SURVEILLANCE REQUIREMENTS 4.2.2 The MTPF for each class of fuel shall be determined, the value of T calculated, required:

and the flow biased APRM trip setpoint adjusted, as

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 />,

b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions 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 ROD PATTERN for MTPF.

BRUNSWICK-UNIT 1 3/4 2-4

Bases Table E 3.2.1 't SIGt41 FICA!iT IrlPUT PARAMETERS TO THE l

LOSS-OF-C00LAt4T ACCIDEriT AtiALYSIS FOR BRdilSWICK-ViilT 1 Plant Parameters ;

Core lhermal Power ........... . . . . . . . . 25 31 IMt , which cc,rresponds to 105% of rated steam flot. *

{

Vessel Steam Output .......... .

6

. 10.96 x 10 Lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure.... .... 1055 psia Recirculation Line Break Area ,

for Large Breaks - Discharge

- Suction 2.4 ft' (DBA) 1.9 ft (807. DBA) 4.2 ft2 Number of Drilled Bundles 560 l

" "" #8 PEAK TECHflICAL If1ITI AL SPECIFICATI0fi DESIGri MIfilMUM LIfiEAR HEAR AXIAL CRITICAL FUEL BU! IDLE GEfiERATIOl1 RATE PE AKIfiG POWER

' FUEL TiPES GEOMETRY (kw/ f t ) FACTOR RAT 10 "

ALL 8x8 13.4 1.4 1.2 A more detailed list of input to each model and its source is presented in Section II of Reference 1.

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

, ** T o account for the 27 uncertainty in bundle power required b, Aprendix K, 9'T calculatien is per f orrmd wi t h an MO7 e f 1.15 (i.e., 1. 2 d ivided by 1.t+0.! )

et a bundle with an initial !!CPE of 1.20.

3RUtiSWICK - UtilT 1 B 3/4 2-2

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

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

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

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

The control rod system is analyzed to bring the reacor subcritical at a rate fast enough to prevent the MPCR from becoming less than 1.07 l during the limiting power transient analyzed in Section 14.3 of the FSAR.

This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MPCR remains greater than 1.07. The occurrence of scram times longer than those specified l

should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods o# time with a potentially serious problem.

Control rods with inoperable accumulators are declared inoperable and Speci fication 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less , reactivity insertion BRUNSWICK - UNIT 1 B 3/4 1-2

POWER DISTRIBUTION LIMITS BASES

1. General Electric Company A...lytical Model for Loss-of-Coolant 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, NEDD-20360, Revision 1, November 1974

5. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NEDO-10302).

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

BRUNSWICK - UNIT 1 B 3/4 2-6

' s SPECIAL TEST EXCEPTIONS 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS LIMITING CCNDITION FOR OPERATION 3.10.3 The requircments of Specifications 3.9.1 and 3.9.3 and Table 1.2 may be suspended to permit the reactor node switch to be locked in the Startup position and to allow up t o three control rod.s to be withdrawn for l shutdown raargin demonstrations provided at l e a s t. the following require-nents are satisfie' J.

a. The source range nonitors are OPERABLE with the RPS circuitry shorting links removed per Specification 3.9.2,

b. The rod worth ninimizer is OPERABLE per Specification 3.1.4.1 and is programmed for the shutdown margin denonstration, and

c. The " notch-override" control shall not be used during movement of the control rods.

APPLICABILITY: CONDITION 5, during shutdown margin demonstrations.

ACTION With the requirements of the above specification not satisfied, irrae-diately restore the reactor mode switch to the Refuel position.

SURVEILLANCE REQUIREMENTS 4.10.3 Within 30 ninutes prior to the performance of a shutdown margin denonstration verify that;

a. The source range monitors are OPERABLE per Specification

3. 9.2, a nd

b. The rod worth ninimizer is OPERABLE with the required pronram, per Speci fication 3.1.4.1, BRUNSWICK - UNIT 1 3/4 10-3

T

_ABLE 3.3.4-2 en

o g CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS w

5 h TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE E

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

a. Upscale (Flow Biased) < (0.66 W + 425) T* < (0.66 W + 425) T*

b. Inoperative fiA fli PF fiA KTPF

c. Downscale s 3/125 of full ;cale > 3/125 of full scale

d. Upscale (Fixed) [12%ofRATEDTHERMALPOWER [125ofRATEDTHERMALPOWER

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

a. Upscale < (0.66W + 40;) T* < (0.66 W + 40% ) T* l

b. Inoperative NA MTPF NA MIPF

}{ c. Downscale > 3/125 of full scale > 3/125 of full scale SOURCE RA1GE MONITORS (C51-SRM-K600A,B,C,D)

a. Detector not full in NA 5

"A 5

b. Upscale < 1 x 10 cps < 1 x 10 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 MA NA

b. Upsmale < 103/125 of full scale < 108/125 of full scale

c. Inoperative NA NA

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

• T = 2.45 for 8 x 8 fuel T = 2.48 for (8 x 8)R fuel

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FUEL TYPE 8DRB28311 (8x8R)

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLEGR)

VERSUS PLANAR AVERACE EXPOSURE FIGURE 3.2.1-4

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FUEL TYPE 8DRB265L (8x8R)

MAXIMUM AVERACE PLANAR LINEAR llEAT GENERATION RATE (MAPLllGR) .

VERSUS PLALiR AVERAGE EXPOSURE FIGURE 3.2.1-3

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FUEL TYPE 2 (8x8)

EiXIMUM AVERAGE PLA';AR LINEAR l! EAT GENERATION RATE (MAPLilGR)

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FUEL TYPE 1 (8x8)

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLilGR)

VERSUS PLA.!AR AVERAGE EXPOCURE FIGURE 3. 2.1- 1

I PC'.ER DIS TRIBUT IC'; L 111 TS 3/4.2.3 M I '; I:*0. ' CRITIC /1 PU.Ei, R AT I O L IMIT II.S CONDIM r0k Gi'E!KT IM

3. 2. 3 The M'NIMUM CR!TICAL POWER 'lATIO (FCPR), as a function of core flow, shali ,e equai to or greater than :tCpg <. tne t' 7 s r.c. . in figure 3.2.3-1 where

?!CPR - 1.22 tor lioC2 t o (U C - 2000 'r.-lD/t)*

!!CPR = 1.21 tor (EOC2 - 2000 'rJD/t) to (EOC2 - 1 '

,?fdD/t)

':C P;' - 1.2M !or (FuC2 - 1000 .lD / t ) to I:OC2 APPLIU4!LITY:

CCNDI T I. .'; 1, nhe . T Hi f&'_ 'tT.E R 255 RATED THERMAL

' u.. u A C.T.1.0.'. :

With MCPR less thy 'rn: o ppl i e. :.L l e l i ni t de te r: i r.ed t rc:, F i gu re 3.2.3-1, initiate : o r r e c. t i v e :c t ion ' i th i' 15 ii ma t c" ,:nd car;i ra cor rec ti', -

ac* ion so that .Ric om.:I tu or qcea ter than Lt.e irolicable limit w . I h i n 4 'ua" or re d o..e T H F ' . '., c i ' .' ,. R t o l e v, t h a r. J b uf RAIED THEk"AL i>/.nR within the n4 .t

. . i.uars SURV E il i S.C! Ri C,ili I'" ';TS

. 1' .1 MCPR ',ha l l tm de te n . i ned t o L;e eauat to or greater th.n the applicable l mi t ce ten ined f ra . Figure ' .9. a-l :

a. At leas t or.ce per .'a huurs ,

b. Wher ner THERVAL IUWER has been increased by at least 155 01 U ATl 0 THi!d'3! F f '.-?! D and eteady state operating conditions havt been estaulished, ar.J c Init.ially ar.J at least cnce 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 witn a LIMITING CONTROL ROD PATTERN for "CPR.

• BOC2 Bea,inninR of Cycle 2 EOC2 Cad o: Cyc 1, . 3 BRUNS'.'IC K-UNIT 1 3/4 2-0

IiWER DISTRIBUTIC', LIMITS r , ,.

i- .J 3/4.2.2 APr " 5F T PU:',TS The tuel e l eold in e integrity seifety l im i t s of Specification 2.1 were based on

.i Tu l'Ai. Pl.AK I M; F/u;1 t lK u t 2.43 lor 8:;8 luel and 2.43 for 8xSR inel. the scram setting and rod block f unctions of the ;H " instrur ents must be adjusted to ensure that t he 'iCPR does r.J t become les s

• na r 1.0 in the degraded situation. Tne scra.- s et t i r.g s and red blac k s e t t i ns; 3 are adjusted in a c c o r d a r,c t. with, tne forc:ula in this specificaticn when the combination o f THER"T,L PC.;E P u.d p ea t. flux indicnes a TOTM PE/J.It;C FACTOR greater inan 2.03. The ne'hcd used to determine the design TPF shall be con-sinent .;ith the >thod used to determine the 'iTPF-3/4.'>3 I . U""'

- C R I T I CM v'A  % i10 Tne required operating limit MCPP's at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding intejrity Satety L l ahnurual oparational transier.ts'j yit "CP;, of 1.07, FCr any and Cperating abnormal an analysis of tran-sient analysis evaluaticn eith the initial condition of the reactor Leing at the ',teadj state operating linit, it is required that the resul ting MCPR Joes not decrease Leicw the Safety Lmit "CPR at any tine during the transient assum n; instrument trip setting as given in S pec i fica ticn 2.2.1, To assure that the fuel cladding 'ntegrity Safety Limit is r.ot exceeded during any anticipated abncrmal operacioral transient, the most limiting transients have been analy:ed to datermine which result in the largest reduction in CRITICAL PCUER 1ATIO (CPR), The type of transients evaluated

,eere losa of flo.., increase ,o pressure and power, positive reactivity insertion, and coolant tem erature decrease.

The limiting transient which determines the required steady stat. MCPR l i:li t is the turbine trip with failure of the turbine by pass. This transient yield the largest . MCPR. When added to the Safety Linit MCFR of 1.07 the required minimum operating limit MCPR of Specification l 3.2.3 is obtained- Prior to the analysis of abnormal operational tran-sients an initial fuel bundle MCPR was determined. This parameter is based on the bundle fic s calculatcd by a GE rulti-channel steady state flow distribution nodel as described in Section 4.4 of NEDO-20360 \"> and on core paraneters shown in Reference 3, response to Itens 2 and 9.

BRU"SWICK - UNIT 1 B 3/a 2-3

R E ACT I V I T Y C0 tit RO.L S Y S T.ly._S ROD BLOCK MONITOR LIMITING CONDITION _EUR OPERATION ._ ____ _ _ _ , _ _ _ _ . ____

3.l.4.3 Both Rod Black Monitor (RBM) channels shall be OPERABLE.

T@ PL IC ABI L I T Y_ :

CONDITICU 1, when THEPMAL POWER is greater th_n the preset po.,er level of the RWM and RSES.

AC T 10ti:

d. With one RCM Channel inoperdble, PO!.TR OPERATIOil may continue provided that either

1. The inoperanle RBM channel is restored to OPERABLE status within ?4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

2. The redundant RBM is demonstrated CPERABl.E within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at lea r+ 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 restared to GPiRACLE status, and the inoperable RBM is restored to WilRABLE status within 7 days, or

3. THERMAL P0.!ER is limited such that MCPR will remain above 1.07 assuning a single error that results in complete l withdrawol of any single control rod that is capable of withdrawal, Otherwise, trip at least one rod block monitor channel,

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

SURVEILLANCE REQUIRLMENTS ___ _

4.1.4.3 Each of the above required RBM channels shall be demanstrated OPERABLE by perforunce of a CHANNLL FUNCTIONAL TEST and CHANUf L CALI-BRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.

BRUNSWICK-UNIT l 3/4 1-17

2 .1 S A F E T Y L I M I.T.S.

BASES

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

Tm fuel clcddirg, reattor pressure vessel and priracy system p u i. 3 a. Un prir ciN1 barriers tc tne release of radioactive '1at-erials to the envir m . Eifety lir .ts are established to protect the integrity of m se 'urriers during r,ceral plant operations and antici-pated transients iho fuel ciaddina :ntegrity linit is set such T.nat no f uel d. 7;e is calculated to occur if the linit is not. violated. Be-cause fuel dram is not directly otrervable, a step-back approach is used to establis' a Ea fety Linit such that the i l N I KJ" C R I T I C A L P G'.-lE R RATIO (CPR) is no less than 1.07.  !'CPR I.07 represents a conserva- l tive margin relative to th. conditicns required to raintain fuel cladding integrity. The .~ fl cladding is cne of the physical barriers which separate the radiacctive oatr: rials f ecr tne environs. The integrity of this Lladding L.:rr ier in related t o i ts relative freedon f rom perfor-ations ce crack.ing. Altho..qh some corrosion or use related cracking na) cccur during the life of tr.e claddin1, fission product migratica frcm this source is increnentally cumulative and continuously neasurable.

Fuel cladding perf nca t ion s , hov,evc r , can resul t from thernal stresses which occur frac reactor operation sienificantly above design conditions and the Linitin' % fety , s t e. Settings. 'clhi l e fi ss ion product ni gra-tion teori cldJina ;mrforation is just as leasurenle as that from use rel a ted crack i r.g , t he then tally caused cladding perforations signal a tnreshold, beyoaa which still 7reater thermal stresses may cause gross rather tnan incrwentel claddina Jetericration. Therefore, the fuel cladJing Safety L ti t is defined with a narr;in to the conditions which would produce onset of transition hoiling, PCPR of 1.0. These con-ditions represent a significant departure from the conditicn intended by design for planned operation.

2.1.1 THER'GL PBER (Low Pressure or Low Flow)

The use of the GEXL correlation is not valid for all critical power calcuidtions at pressures oeloa 300 psia or core flows less than 101 rs rated flow. Therefore the fuel cladding integrity linit is established by other neans. This is done by establishing a limiting condition on core THER"AL PC'.ER with the following basis. Since tha pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low po..or and fio.ss will always be greater than 4.5 psi. Analyses s ho.s that with a flow of 23 x 10' lbs/hr bundle flow, bundle pressure drop is nearly independent of Lundle pae: e,d has a value of 3.5 psi, Thus, t."e Sundle flow with a , 5 psi crivinn head will be greater than 23 x 103 lbs/hr Full scale ATLM test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assenbly critical power at this flow i s approxinately 3. 35 M'-lt . With the design peaking factors, this corresponds to a IHERMAL PCWER of more than 509 of RATED THERMAL POWER. Thus, a THER"AL PCWLR linit of 251 of RATED THERMAL PC',!ER for reactor pressure below S00 psia is conservative.

BRUNSWICh-UNIT 1 B J-l

u. -

1 2.0 SAFETY LIMITS T.';D LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMI_TS THERMAL. PCWER (Loa Proscure or Lc. Flc..';

2 .1.1 THERMAL PC'lER shall not exceed 25L of RATED THE RML POWER with the renctor .essel stean dom pressure less than 800 ps;a or core flow less than 10 . o f ca tec f I .re.

APPLICABIllTY: CONDIT IC';5 1 and 2.

ACT 10'.

With THERMAL PO'-lER exceeding 25 of RATED THERMAL PC'ER and thc reactor sessel s team dc:Je pressure less than i"20 psia cr core low less than 101 of ratej tiow, Le it a t leas t HOT SHaTDC'..N wi thin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> .

THERMALLOWLR(HiohPressureandHiohFlc.d 2.1.2 The MINIMUM CRITICAL PCWER RATIO (MCPR) shall not be less tnan 1.07 with the reactor vessel stean. dote pre .sure greater than 800 psia l and core flu grea Ne than 10~ of rated flce.

APPLICABILITY:

CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1 07 and the reactcr vessel steam dome pressuro l ureatcr than 800 psia and core flcw greater than 10' of ratcd flos. be in at least HOT SHUIDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

REACTOR COOLANT B STEM PRESSURE The reactor coolant system pressure, as measured in the reactor h 2.1.3 vessel stean do:ne, shall not exceed 1325 piig.

APPLICABILIT(: CONDITIONS 1, 2, 3 and 4 ACTICN-With the reactor coulant system pressure, as neasured in the reactor vessel s tean: done, above 1325 psig, t,e in at least HOT SHUTCOWN with reactor coolant system pressure < 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

BRUNSWICK-UNIT 1 2-1

};r. T. A. Ippolito January 17, 1979 If your staff has any questions concerning the attached information, we will be glad to discuss them e!.ther by telephone or at a meeting with repre-sentatives of your staff.

Yours very truly,

/?

"Q g-' .A) Q- , (/ Qg4 E. E. Utley Senior Vice President Pouer Supply JAM /mf Enclosures Sworn to and subscribed before me this 17th day of January, 1979 Notary Public My Commission Expires: October 4, 1981

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