ML20137P008

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Removing Values of cycle-specific Core Operating Limits from TS & Relocating Values to Operating Limits Rept
ML20137P008
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 03/24/1997
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20137N986 List:
References
NUDOCS 9704090108
Download: ML20137P008 (12)
v • d • e


Text

. . . . . _ . . _

ATTACHMENT B BRAIDWOOD STATION Proposed Changes to Appendix A Technical Specifications of facility ,

Operating Licenses NPF-72 and NPF-77 hevised Pages:

IV V

1-4 B 2-1 B 2-2 3/41-1 3/4 1-14 3/4 1-15 3/4 1-20 3/4 1-21 3/4 1-22 3/4 2-1 3/4 2-2 3/4 2 3 3/4 2-4 3/4 2-5 3/4 2-8 8 3/4 2-2 B 3/4 2-4 6-22 6-22a F i 9704090100 970324 4 DR ADOCK 0500 A-11

1

] 2.1 SAFETY LIMITS

[/ BASES j 2.1.1 REACTOR CORE t The restrictions of this Safety Limit prevent overheating of the fuel and

.possible cladding perforation which would result in the release of fission

'producu to the reactor cool' ant.

Overheating of the fuel cladding is prevented

. by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is

{ glightly. above the coolant saturation temperature.

~~

i:

Operation above the upper boundary of the nucleate boiling regime could i' result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer

. coefficient. DNB is not a directly measurable parameter during cperation and l

therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonunifore heat flux distri-butions.

. The local DNB heat flux ratio (DNBR) is defined as the ratio of the ,

heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNBR thermal design criterion is that the probability that DNB will 2 not occur on the most I or limiting rod is at least 95% (at a 95% confidence level) l for any Condition II event. (< >

t l(

In meeting this design basis, uncertainties in plant operating parameters, <

nuclear and thermal parameters, and fuel fabrication parameters are considered. 5 As described in the UFSAR, the effects of these uncertainties have been >

t statistically combined with the correlation uncertainty. Design limit DNBR

-values have been determined that satisfy the DNB design criterion.

> (-

~

-Thtdesign DNBR values are 1.34 and 1.32 for a typical cell and a thimble' cell, respettivelytor OFA* fuel,' and 1.33 for a typical cell anA.I Tota, ,

thimble cell for the VANTAE5 fuel (1.25 for the typJral-and' thimble cells)

In addition, margin has been maint11ned4Qoth-dels gns by meeting safety

. analysis DNBR limits of 1.49 for a typical ceTW47 for a thimble cell for -

OFA fuel, and 1.67 and.,L6frfor,a typical cell and a thTable-ca respectively

' for. the VAKTAGE-5-fuel (1.50 for the typical and thimble cells) safetralialyses. n tfo

{

The curves of Figure 2.1-1 (Figure 2.1-la THERMAL POWER, Reactor Coolant System pressure)"and average { tem which the minimum design DNBR is no less than the design DNBR value, or the average liquid enthalpy at the vessel exit is less than the enthalpy of saturated -

/ M J.

e 20st4 mired-fuel- Assemblies-T " Applicable to Unit I and Unit 2 starting with cycle 6.

Unit 1 - Amendment No.  ;

BRAIDWOOD - UNITS 1 & 2 8 2-1 Unit 2 - ^

AmendmentNo.[

, ~ . - , ,~~ - , , , , _ , _ , . . . _ _ , . . , _ _ _ . _ _ _ , _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ . _ _ .

I

+-

  • L .

. l 1

SAFETY LIMITS I

BASES I

  • 'A E S~h REACTOR CORE (Continued - - - - '
7 3g, of 1.i?

l These curves are based on an enthalpy het channel factor, F

-fer OfA iderd 1.50 fcr VASA : 5 fi. l increase 4fr Fh at reduced power,5: ed en th: :g :::1:n'-

fh-1.49(140.3(1=P)]forofA-fue {0 j -Fh-1.52[1+0.3(1-P)3-fc"^.MT^cE5--fuel-4'h re P is th. fractieri ef-RATED-THERMAt-POWER

I .>

Theet limiting heat flux conditions are higher than those calculated for g the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance When the is within axial power the limits of the imbalance i

f 2 (AI) function of the Overtemperature trip.

is not within the tolerance, the axial power imbalance effect on the Overtem-perature AT trips will reduce the Setpoints to provida protection consistent with core Safety Limits.

2.1.2 REACTOR COOLANT SYSTEM PRES $URE l

The restriction of this Safety Limit protects the integrity of the

' Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere, The reactor vessel, pressurizer, and the RCS piping, valves, and fittings j

are designed to Section III of the ASME Code for Nuclear Power Plants which

permits a maximum transient pressure of 110% (2735 psig) of design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria l

i and associated Code requirements.

The entire ACS is hydrotested at 3110 psig,125% of design pressure, to demonstrate integrity prior to initial operation.

4 i

c BRAIDWo0D - UNITS 1 A 2' 8 2-2 AmendmentNo.[

i i

l_

00WER DISTRIBUTION LIMITS  !

BASES t

! L EAT FL.1X OT CHA1NE 30T CHAWNEL FACTOR. and RCS FLOW RATE AND NUCLEAUW FACTOR (Continued) i

c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and
d. The axial power distribution, expressed in tems of AXIAL FLUX i

DIFFERENCE, is maintained within the limits.  :

1 througk d. above are main ained.FS will be maintained within its limits provide The combination of the RCS flow requirement

~F90~,400 gpm (371,400  !

used in the safety ana.ysi gp)}swillbemetand .

the requirement on F,"{guaral '

i Margin between the safety anC ysis limit DNBR:

8 fuel-typical erid thiele ceHsr-re pectiwly e.d 1.6)1.40  ::d 1.47

d-1.05 f rfer the-OFA-th

1ANTAGE 5 typicai ami ibisis celb i

! and the design limit DNBRs 1.35 ivr ihe ig iu.1 ...d thi;.ble ell y ,]-  %

-ceh. -d 1.33 rd 2.32 rf:il.04 :nd(1.02 5 f;;lf.,t '-

the 0,n f.el t

' y-!::1 ::d thir le C

-re;::th:ly (1.25 f:r th:

th: Y^"T^.0E typk:1 ;d thi;.tle ceNU;)4 is maintained.

b ' ::11:,- -,

2 j

p '

A fraction of this margin is utilized to accommodate tb +-~ - ~s. 3 l

d the a l (less than I. '% per WCAP-8691, riate fuel rod bow DNBR penalty

- -DNBR p:;;1t" $ maxi;;a. ;f 12.5") . Revision 1$prokhe rest of the

design flexibility.and safscy analysis DNBR limits can be used for plant design ,

,( of 97,600 The gpm RCS flow requirempnt is based on the loo) minimum measured flow rate 92,850 gpm) which is used in 1

DrMa@re (Ree(sed Thermal Design Procedure)_,the Laproved-Ther=>l 15.0.3. A precision heat balance is perfomeddescribed in UFSAR 4.4.1 and 7l n.31g D once each cycle and is used to d calibrate the RCS flow rate indicators. Potential fouling of the feedwater i

venturi which might not be detected i heatbalanceinanon-conservativema,nner.could bias the results Therefore, a penaltyfromofthe 0.1%precision is  ;

i assessed for potential fee

' uncertainty of 2.2% (3.5%)pwater venturiA maximum measurement-fouling.has been included ,,

o in the of the RCS flow indicators for flow rate verification. rate to account for poten bias the RCS flow rate measurement greater than 0.1% can be detected byAny foulin monitoring and trending various plant performance parameters. If detected action shall be taken before performing subsequent precision heat balance, measurements, i.e., either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling.

Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect possible flow reductions due to any rapid core crud bulidup.

Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen-tation shall be calibrated within seven days prior to the' performance of the calorimetric flow measurement. This requirement is due to the fact.that the drift effects of this instrumentation are not included in the flow measurement uncertainty analysis. This requirement does not apply for the instrumentation whose drift effects have been included in the uncertainty analysis.

1

(

' Applicable to Unit 1.and Unit 2 starting with cycle 6. 4 Unit 1 - Amendment No. 5[

. BRAIDWOOD - UNITS 1 &'2 . B 3 /4 2-4 _ . . ___ Mnit ? - a- d-- + u- rd

i 2

ADMINISTRATIVE CONTROLS REPORTI4 RE0VIREMENTS (Continued)

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

  • 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted prior to May I of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections

, IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT ** l 6.9.1.7 A Radioactive Effluent Reler.se Report covering the operation of the facility during the previous year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive

' liquid and gaseous effluents and solid waste released from the facility. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR'50.36a and Section IV.B.1 of A: '9ndix I to 10 CFR Part 50.

HONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, i including documentation of all challenges to the PORVs or RCS safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear. Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than ,

the 15th of each month following the calendar month covered by the report. )

g OPERATING LIMITS REPORT l

?> ' frc9.1 L Operating limits shall be established and documented in the OPERATING <

LIMITS REPORT-before each reload cycle or any remaining part of a_ reload '

cycle. The analytical methods used to determine the  ;

those previously reviewed andijiproved3y the-NRCTn'. operating'11mits shall be Topical Reports:

l. WCAP-9272-P-A,"Westinghoiiseh S tions-Methodology" dated p

July 1985* '

  • ~*N

,,A single submittal may oe made for a multi-unit station.

A single submittal may be made for a multi-unit station. The eutrittal should combine those sections that are comon to all units at the statNn; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

4 s

ADMINISTRATIVE CONTROLS

/* _

OPERATING LIMITS REPORT (Continued) -

2. AP-8385, " Power Distribution Control and Load Following Procedures-Topltal Report" dated September 1974.

3.

NFSR-0016,s"ign Nuclear Des s Methods" dated July 1983. Commonwealth Edison Company

4. NFSR-0081, " Commonwealth Edison Company Topical Report on-Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," dated 5.

July 1990. N /

Comed letter from D. Saccomando to the Office-of Nuclear Reactor Regulation dated December 21,'1993, transmit' ting an attachment that documents applicable sections of WCAE-11992/11993 and Comed application of the UET methodology addrnssed in "Additiqnal Information Regarding T Application for Amendment to .ac111ty Operating Licenses-Reactivity Controls Systems." N The operating limits shall be determined so that all appT1 cable limits (e.g.,

fuel thermal-mechanic'al limits, core thermal-hydraulic limitsrsECCS limits, nuclear limits sn'ch as shutdown margin, and transient and accidehtsa nalysis limits) of the safety analysis are met. The OPERATING LIMITS REPOR ,

3 including any mid-cycle revisions or supplements thereto, shall be prov up.on' issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. 's.

.N l(

4 i

t t

E 1

^

i i

4 4

!( BRAIDWOOD - UNITS 1 & 2 6-22a AMENDMENT NO. 6

ATTACHMENT B BYRON STATION i

Proposed Changes to Appendix A Technical Specifications of facility l Operating Licenses NPF-37 and NPF-66 l 1

Revised Pages:

IV V

1-4 B 2-1 B 2-2 i 3/4 1-14 3/4115 3/4 1-20 3/4 1-21 3/4 1-22 l 3/4 2-1 j 3/4 2-2  ;

1 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-8 B 3/4 2-2 B 3/4 2-4 6-22 6-22a l

I I

i A-12

r 2.1 ~ SAFETY LIMITS- J

('

l BASES i

-2.1.1- REACTOR CORE '!

-The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission- l Products to the reactor coolant. Overheating of the fuel cladding is ,)

prevented by restricting fuel operation to within the nucleate boiling regime

'where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coclant saturation temperature. ,

Operation above the upper. boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer .

coefficient. DNB is not'a directly measurable parameter during operation and-therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been i related to DNB. This relation has been developed to predict the DNB flux-and-the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio (DNBR). is defined as the ratio of the heat flux.

that would cause DNB at a particular core location to the local heat: flux, and is indicative of the margin to'DNB.<

The DNBR thermal-design criterion is that'the probability th'at DNB'will 7 not occur on the most limiting rod is at least 95% (at a 95% confidence level) <

for any Condition I or II event.

{ (

In meeting this design basis, uncertainties in plant operating parar.sters, nuclear and thermal parameters, and fuel fabrication' parameters .d are considered. As described in the UFSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty. Design

(

)

) limit DNBR values have been determined that satisfy the DNB design criterion.

NT - -

' and 1.Rdesign DNBR values 32 fbr a-typical cell and area1.25 thimble for the typical cell, and thimble respectively cells or4f fuel, ,

and 1.33 for_ a typicilwell and 1.32 for a thimbl csM r;the VANTAGE 5. j fuel *). N Injaddition, margin his-bee both designs"by meeting: .> q _

safety analysis 'DNBR limits' o ~ he'typ al and thimble cells (1.49 7 1 for a typical cel d7 or a thimble cell- 0FA-fue and 1.67 and 1.65 0 nd h mble cell, respectively for th VANTAG '< -

4 The curves of Figure 2.1-1 -(Figure 2.1-la") show the' loci of points of THERMAL POWER,: Reactor Coolant System pressure and average temperature for (

which the minimum design DNBR is no less than the design DNBR value, or the average u enthalpy at the vessel exit is less than the enthag. saturated -  !

D AJ M 5 '

W s ,s

  • Not' applicable to Unit ~1. Appli able to Unit 2 until completion of cycle 5. S

&i- , , = = = , . .

BYRON ' UNITS 1 & 2 B 2-1 Amendment No. l T

- e - ,e-- , - - - . , - -- _ _ _ . _

L j- SAFETY LIMITS BASES REACTOR CORE (Continued) /La., fj~. h

  • m M4 ak

.m alue These curves are base on enthalpy hot channel' actor, Ffg, W e fcr 0F^. fuel er,d 1.50 for "ANTAO 0 fuel. An- allowance -i+-included fem t

' increase 4tt- F g at reduced power based er. the egressiefu-N r3H I* U O' II IUI OIA I"*l

-FN _ 7, g 1 g,3 ,

f gg 3 7 )

3H

-Where-P-is-the frectier, ef RATED THERMAL F0WER-These limiting heat flux condition's are. higher than those calculated for

'the range of all control. rods fully withdrawn'to the maximum allowable control l ,

rod insertion assuming the axial power imbalance is within the limits'of the fl (AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtem-

-(~. perature,AT trips will . reduce the Setpoints to. provide protection consistent '

with core Safety Limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE

'ThersstrictionofthissafetyLimitprotectstheintegrityoft'd h Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of~ radionuclides contained in the reactor coolant from reaching the containment atmosphere.

.an .

Thelreactor' vessel, pressurizer, and the RCS piping, valves, and fittings , l are designed to Section III-of the ASME. Code.for Nuclear, Power, c Plants w~hich .

permits 2aimaximum'transientt pressure"of 110%'(2735' psig) .

of design" pre'ssure'l ""~'", <

The Safety Limit of 2735 psig,is therefore consistent with the design criteria-

.and associated Code. requirements. .

The' entire RCE is hydrotested at 3110 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

s

n- e n -

y'[.

f

~ BYRON - UNITS 1 & 2: 'B 2-2 . AMENDMENT NO.

a C  !

POWER DISTRIBUTION LIMITS BASES

~ KEAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEA CHANNE_ FACTOR (Continued)
c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and i 1
d. The axial power distribution, expressed in terms of AXIAL FLUX

' DIFFERENCE, is maintained within the limits. i FL will be maintained within its limits provided the Conditions a.

[371, h d. above are maintained. The combination of Jhe RCS flow requirement throuk00gpm(390400gpm DNBR used in the s,afety an)a]

and the requirement on Fu guarantee that the lysis will be met. h, tMable-ceMs-Margin between the safety analysis limit DNBRs 'I.50 fer-the-typical-and ,

-respectively-an{4d9 and the desi :r.d-h47 for theu OfA fuel typica" ar.d thimb d-h67-and-h 65-for-the-VANTAGE-T-ty

-Ir32-for-thegn limit DNBRs 'I.25 for the typical OFA fsel t e..pical-and-thimble cethimble VANTAGE 4-fuel-typical ypical erd thisle respectivel end-tMable-ceHs, ceHs and nd 1.32 ivi the is maintained. .

A fraction'of this margin is utilized to accommodate -the-transition-core-

-and the ap 4NBR-penalty-jmeximum-of-12r55)

(less than 1.o% per WCAP-8691, .Revision

=.ha rest of1)prohriate fel rod bow DNBR p the margin between design and safety analysis DNBR limits can be used for pynt Asign i

flextbility.  :

The RCS flow'requirepent is based'on the loop minimum measured flow rate Procedyre pm (97,600 gpm ) which is used in the Revised Thermal Design of 92,850 (

( -

15.0.3-4 sd precision prevad-Thermal-Besign Procedure described in UFSAR 4.4.1 and < l.

heat balance is performed once each cycle and is.used G to cali) rate the RCS flow rate indicators. Potential fouling of the feedwater venturi, which might not be detected, could bias the results from the precision heat balance in a non-conservative manner. Therefore a penalty 1 of 0.1% is assessed for potential fee water venturi fouling. Amaximum measurement uncertainty of 3.5%

measured flow rate to account for(2.2%p) has been included in,j-the potential undetected feedwater venturi lo fouling and the use of the RCS flow indicators for flow rate verification.

Any fouling which might bias the RCS flow rate measurement greater than 0.1% 1 l

can be detected by monitoring and trending various-plant performance:

parameters. If detected i precision heat balance me,asurementsaction i.e. shall be taken, before performing subsequent  !

! either the effect of fouling shall be quantified and compensated for in, thelCS flow rate measurement, or the ' '

venturi shall be cleaned to eliminate the fouling, cw 4

Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect possible flow reductions due to any rapid core crud buildup.

i' Surveillance Requirement 4.2.3.5 specifies that the measurement instrumentation shall be. calibrated within seven days prior to the performance of the calorimetric flow measurement. This requirement is due to the fact that the drift effects of this instrumentation are not included in the flow

' measurement uncertainty analysis. This requirement does not apply for the

' instrumentation whose drift effects have been included in the uncertainty analysis.

4

  • Not applicable to' Unit 1. Applicable to Unit 2 until completion of cycle 5. C

(

BYRON - UNITS 1 &'2 B 3/4 2-4 Amendment No [

!g

  • i '

l i

ADMINISTRATIVE CONTROLS l REPORTING REOUIREMENTS (Continuedl ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

  • 6,9.1.6 The Annual Radiological Environmental Operating Report covering the prior to May of operation 1 ofthe facility each year. during the previous calendar year shall be submitted The report shall include summaries, interpreta-Monitoring Program for the reporting of tions, and analysis of trends the resuits of the Radiological Environmental period.

The material provided shall be consistent with IV.B.3, and IV.C of the objectives Appendix outlined I to 10 CFR in (1) the ODCM and (2)

Part 50. ..,

Sections IV B 2 l

ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **

i 6.9.1.7 A Radioactive Effluent Release Report covering the operation of the The report shall include a summary of the quantities o .

gaseousshall provided effluents be 1and solid waste released from the facility. The material and (2) in conforma (nc)e with 10 CFR 50.36a and Section IV.B 10 CFR Part 50.

I MONTHLY OPERATING REPOR_T_

6.9.1.8 Routine reports of operating statistics and shutdown experience ,

shall be submitted on a monthly basis to the Director, Off ,

Management, U.S. Nuclear Regulatory Commission,, D.C. Washington 20555, with a the 15th of each month following the calendar month cover .

WVy >

DPERATING LIMITS REPORT 6h(0perating LIMITS REPORf4e limits shall be established and documented in the OPERAT N '

The analytical m[ ore o each reload cycle or any remaining part of a reloa cycle.

previously reviewed and appraveddetermine the operating ligits-strsT1 be those etliods-mQto the NRC in TogicaPReports:

1.

WCAP-9272-P-A, "Westinghou July 1985. Reltrr SafetrEv uations Methodology" dated

2. WCAP-838 jS d September 1974.istribution Control and Load Following dures-Topical

,,A single submittal may be made for a multi-unit station *A s .

combine those sections that are common to all units at the station; , for howeverThe su anits with separate radwaste radioactive material from each unit.

systems, the submittal shall specify the releases o BYRON - UNITS 1 & 2 6-22 AMENDMENT NO

~'

I ADMINISTRATIVE CONTROLS NDPERATINGLIMITSTEPORT(Continued) 3 /

Nuclea M-0016, " Commonwealth Edison Company Topical Report on Benchmark R

4. NFSR-0081ggnMethods"datedJuly1983.

Commonwealth Edison Company Topical Report on Benc rk of PWR Nuclear July 1990.Design' Methods Using the Phoenix-P and ANC Compu Codes," dated

5. N Comed dated letter from D. Sacto ando to the Office oclear Reactor Regulation December 21, 1994, t(raftsnitting an at ment that documents applicable sections of WCAP-11992/11993 Ed anhat ication of the UET methodology addressed in " Additional Informati seg ing Application for Amendment to facility Operating Licenses-Rep vity C Sqls Systems."

The operating limits shal] Atfdetermined so that all applicable limits (e g fuel thermal-mechanic 14imits, core thermal-hydraulic limits . . ,

nuclear limits su y ECCS limits, limits) of th afetyasanalysis shutdown margin, are met. and transient and acciden( analysis The OPERATING LIMITS REPOR7 M ncluding forjac reload cycle, to the NRC Document Control Desk with c J gional Administrator ard Resident Inspector.

N

_ -

Retrieved from "https://kanterella.com/w/index.php?title=ML20137P008&oldid=2831312"