ML20107A864

Proposed Tech Specs for Siemens Power Corp Fuel Transition

ML20107A864
Person / Time
Site:	LaSalle
Issue date:	04/08/1996
From:	COMMONWEALTH EDISON CO.
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ML20013A388	List: ML20107A835 ML20107A864 ML20107A897
References
NUDOCS 9604150172
Download: ML20107A864 (86)

Text

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INDEX DEFINITIONS SECTION 1.0 DEFINITIONS ffGf 1-1 1.1 ACTI0N...........................................................

EXPOSURE.......................................... 1-1 1.2 AVERAGE PLANAR 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . .

CALIBRATION.............................................. 1-1 1.4 CHANNEL 1-1 1.5 CHANNEL CHECK.................................................... .

1-1 1.6 CHANNEL FUNCTIONAL TEST..........................................

1-2 1.7 CORE ALTERATION..................................................

1-2 1.8 CORE OPERATING LIMITS REP 0RT.....................................

1-2 1.9 CRITICAL POWER RATI0.............................................

1-2 1.10 DOSE EQUIVALENT I-131............................................

1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY..................................

TIME............... 1-2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE N M 4 TIME........ 1-2 1.13 END-OF-CYCLE RECIRCULATION PLHP TRIPyST 1.14 hC[IO)l0[LflTJhG/ f 0[Rf f lyIITp...... . ......... 1-3 1-3 1.15 FRACTION OF RATED THERMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-3 1.16 FREQU EN CY N0 TAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.17 GASEOUS RADWASTE TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-3 1-3 1.18 IDENTIFIED LEAKAGE...............................................

ISOLATION SYSTEM RESPONSE TIME...................................

1-3 1.19 1-4 1.20 DELETED..........................................................

LIMITING CONTROL R00 PATTERN.....................................

1-4 1.21 1-4 1.22 LINEAR HEAT GENERATION RATE......................................

1-4 1.23 LOGIC SYSTEM FUNCTIONAL TEST...............f . ....... ... .

1.24 hijdfj/FJd $'rfd )I/ ylyfT///0yd[D)fpl( J ..

. . 8. . 1-4 1-4 1.25 MEMBER (S) 0F THE PUBLIC..........................................

1-4 1.26 MINIMUM CRITICAL POWER RATI0.....................................

LA SALLE - UNIT 1 I Amendment No. 110 9604150172 960400 PDR ADOCK 05000373 P PDR

J 1

i INDEX

! LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l'

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1 SECTION PAGE i

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM IASTRUMENTATI0N.................... 3/4 3-1 l

3/4.3.2 ISOLATION ACTUATION INFTRUMENTATI0N.......................... 3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTR! MENTATION...... 3/4 3-23 j 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION li A1WS Recirculation Puep Trip System Instrumentation.......... 3/4 3-35

! End-of-Cycle Recirculation Pump Trip System Instrumentation............................................ 3/4 3-39 3/4.3.5 REACTOR CORE ISOLATION C00 LING SYSTEM ACTUATION l INSTRUMENTATION............................................ 3/4 3-45

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! 3/4.3.6 CONTROL ROD WITHDRAWAL BLDCK INSTRUMENTATION................. 3/4 3-50 j 3/4.3.7 MONITORING INSTRL' MENTATION i

Radiation Monitoring Instrumentation......................... 3/4 3-56 Seismic Monitoring Instrumentation........................... 3/4 3-60 Meteorol ogical Monitoring Instrumentation. . . . . . . . . . . . . . . . . . . . 3/4 3-63 i

l Remote Shutdown Monitoring Instrumentation...................

Accident Monitoring Instrumentation..........................

3/4 3-66 h

h 3/4 3-69 3 '

Source Range Monitors........................................ 3/4 3-72 v'

{ {TraversingIn-coreProbeSystem.............................. 3/4 3-7 i Deleted...................................................... 3/4 3-74

' Fire Detection Instrumentation............................... 3/4 3-75 Deleted...................................................... 3/4 3-81 Explosive Gas Monitoring Instrumentation. . . . . . . . . . . . . . . . . . . . 3/4 3-82

! Loose-Part Detection System.................................. 3/4 3-85 l 1 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION l INSTRUMENTATION............................................. 3/4 3-86 4

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1 LA SALLE - UNIT 1 V Amendment No. 85 i

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. _ . . 9

e DEFINITIONS

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CORE ALTERATION

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2 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of i

the movement of a component to a safe conservative pas,1 tion.

j CORE OPERATING LIMITS REPORT

] 1. 8 The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle.

! These cycle-specific core operating limits shall be determined for each

! ' reload cycle in accordance with Specification 6.6.A.6. Plant operation' f

within these operating limits is addressed in individual specifications.

l CRITICAL POWER RATIO 9)#NN 74 i

i 1. 9 The CRITICAL POWER RATIO (CPR) shall be the ratio of at power in the 1 assembly which is calculated by application of the correlation to

! cause some point in the assembly to experience boiling transition, l

{

divided by the actual assembly operating power. l j DOSE EQUIVALENT I-131 j

1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131, .

! microcuries/ gram, which alone would produce the same thyroid dose as the l

quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this i

j calculation shall be those listed in Table III of TID-14844, " Calculation 4

of Distance Factors for Power and Test Reactor Sitas."

l 4

E-AVERAGE DISINTEGRATION ENERGY i

1.11 I shall be the. average, weighted in proportion to the concentration of i i

sach radionuclide in the reactor coolant at the time of sampling, of the Me 4

sum of the average beta and gamma energies per disintegration, in ,

j for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant, j EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME s

1 1.12 The ENERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME l sh interval from when the monitored parameter exceeds its ECCS actuation j setpoint at the channel sensor until the iCCS equipment is capable of j performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.

! Times applicable.

where shall include diesel generator starting and sequence loading delays The response time may be esasured by any series of sequential, overlapping or total steps such that the entire response time

{ is measured.

j END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 4

l 1.13 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE l TIM that time interval to energization of the recirculation pump circuit 3 .

i LA SALLE UNIT 1 1-2 l Amendment No. 70 l

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DEFINITIONS END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME (Continued) breaker trip coil from when the monitored parameter exceeds its trip setpoint at the channel sensor of the associated:

a. Turbine stop valves, and

b. Turbine control valves.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

hCTIONOFLIMITINGPOWERDENSITY 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR . existing at a given location divided by the specified LHGR limit for that bundle Qype.

FRACTION OF RATED THERMAL POWER 1.15 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

FREOUENCY NOTATION .

1.16 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM I

1.17 A GASE0'JS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary i coolant system offgases from the primary system and providing for delay i

or holdup for the purpose of reducing the total radioactivity pticr to

release to the environment.

IDENTIFIED LEAKAGE 4  ;

1.18 IDENTIFIED LEAKAGE shall be

a. Leakage into collection systems, such as pump seal or valve -

packing leaks, that is captured and conducted to a sump or collecting tank, or i

b. Leakage into the containant atmosphere from sources that are I both specifically located and known either not to interfere t with the operation of the leakage detection systems or not to '

be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter eneeds.its isolation actuation setpoint at the 4

channel sensor until the isolation valves travel to their required i

positions. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any i series of sequential, overlapping or total steps such that the entire

response time is measured.

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LA SALLE UNIT 1 1-3 Amendment No. 102

DEFINITIONS 1.20 DELETED LIMITING CONTROL R00 PATTERN 1.21 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit,- i.e., operating on a limiting l value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.22 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat aeneration per unit length of fuel rod. It is the integral of the heat flux over the heat j transfer area associated with the unit length. ,

LOGIC SYSTEM FUNCTIONAL TEST  !

l 1.23 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, 1.e, all relays and contacts, all trip units, solid state logic elements etc. of a logic circuit, from sensor through and including the actuated ,

THE LOGIC SYSTEM FUNCTIONAL TEST may be device to verify OPERABILITY.

performed by any series of sequential, overlapping or total system steps such that the antire logic system is tested. '

MAXIMUM FRACTION OF LIMITING POWER DENSITY l.2 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall b 6t pep [highestvalueoftheFLPDwhichexistsinthecore.

b MEMBERS (S) 0F THE PUBLIC 0F THE PUBLIC shall include all persons who are not occupationally 1.25 MEMBER associate (S)d with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this categoryThis are l persons who enter the site to service equipment or to make deliveries.

category does include persons w;o use portions of the site for recreational, occupational, or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smalle ,

exists in the core. ,

OFFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of in offsite doses resulting the calculation from of gaseous radioactive gaseous and liquid effluents, Setpoints, and in the conduct and liquid effluent monitoring Alarm / Trip of the Environmental Radiological Monitoring Program. The ODCM shall

1) the Radioactive Effluent Controls and Radiological also contain (Monitoring Programs required by Technical Specification Environmental descri tions of the information that should be Section 6.2.F.4 and (2)Radiolo ical Environmental Operating and Semi-included in the Annual Annual Radioactive Effluent Re ease Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

1-4 Amendment No.110

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} 2.1 SAFETY LIMITS i

BASES

2. 0 The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the 3 environs. Safety Limits are established to protect the integrity of these barriers during nomal plant operations and anticipated transients. h fuel cladding integrity Safety Limit is set such that no fuel damage is calculated i

to occur if the limit is not violated.' 8ecause fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that i the MCPR is not less than 1.07. . MCPR greater than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel .]

cladding integrity. The fuel cladding is one of the separate the radioactive materials from the environs. physical barriersof The integrity which this j cladding barrier is related to its relative freedom from perforations or crscking.

i Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incre-i i

mentally cumulative and continuously measurable. Fuel cladding perforations,

however, can result from thermal stresses which occur from reactor operation i significantly above design conditions and the Limiting Safety System Settings.

While fission product migration from cladding perforation is just as measurable

as that from use related cracking, the themally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross i

rather than incremental cladding deter oration. Therefore, the fuel cladding i

Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a signif-i 1 icant depWre from the condition intended by design for planned operation.

f 2.1.1 TrmAL POWER. Low Pressure or Low Flow

[

a The use of the GEXL correlation is not valid for all critical power calculations flow.

at pressures below 785 psig or core flows less than 10% of rated i

4 Therefore, the fuel cladding integrity Safety Limit is. established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure' drop in the bypass region is essentially all elevation head, the core pressum drop at low power and flows j will always be greater than 4.5 psi. Analyses show that ~with a bundle flow of 28 x 108 lbs/hr, bundle pressura drop is nearly independent of bundle power i

and has a value of 3.5 psi.

I will be greater than 28 x 10s 1bs/hr.Thus, the bundle flow with a 4.5 psi driving head  ;

! Full scale ATLAS test data taken at pres- j i sures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 mit. With the design peaking factors, this

corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus,

- a THERMAL 785 POWER limit of 255 of RATED THERMAL POWER for reactor pressure below psig is conservative.

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i LA SALLE - UNIT 1 B 2-1 Amendment No. 40

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l Insert #1 k

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For certain conditions of pressure and flow, the ANFB correlation is not l

} valid for all critical power calculations. The ANFB correlation is not valid' for 8

i bundle mass velocities less than 0.10 X 10 lbs/hr-ft"(equivalent to a core flow of l less than 10%) or pressures less than 590 psia. Therefore, the fuel cladding j

integrity Safety Limit is established by other means. This is done by establishing i a limiting condition on core THERMAL POWER with the following basis. Since j the pressure drop in the bypass region is essentially all elevation head, the core i

pressure drop at low power and flows will always 8 be greater than 4.5 psi. . l Analyses show that with a bundle flow of 28 X 10 lbsthr (approximately a mass velocity of 0.25 X 10' lbsthr-ft"), bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. 8 Thus, the bundle flow with a 4.5 psi i

driving head will be greater than 28 X 10 lbsthr. Fuli4cale ATLAS test data taken j ct pressures from 14.7 to 800 pela indicate that the fuel assembly critical power at 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 j POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

O e

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$AFETY LIMITS BASES t

2.1.2 THERMAL POWER. High Pressure and High Flow .

The fuel cladding integrity Safety 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. Howevsr, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critica' power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR 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 a is. determined using the General Electric Thermal lysis 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.

Thebgsesfortheuncertaintiesinthecoreparametersaregivenin NEDD-20340 and in NEDD-10958-A,the basis for the uncertainty in the GEXL correlation is given The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during an as the distribution used in the analysis.y fuel cycle would not be as severe

a. " General Oectric BWR Th rmal Analysis Bases (GETAB) Data, Correlation and Design Application," NEDD-10958-A.

b. General Electric " Process Computer Performance Evaluation Accuracy" NEDD-20340 and Admondment 1 NEDD-20340-1 dated June 1974 and December 1974, respectively.

h5NN LA SALLE - UNIT 1 B 2-2 Amendment No. 58 l

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Insert #2 The Safety Limit MCPR is determined using the ANF Critical Power Methodology for boiling water reactors (Reference 1) which is a statistical model that i combines all of the uncertainties in operation parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is l determined using the SPC-developed ANFB critical power correlation. l The bases for the uncertainties in system-related parameters are presented in NEDO-20340, Reference 2. The bases for the fuel-related uncertainties are found in References 1,3-5. The uncertainties used in the analyses are provided in the cycle-specific transient analysis parameters document.

1. Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors / Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects /NRC Correspondence, XN-NF-524 (P)(A) Revision 2 (as supplemented) November 1990. -

2. Process Comruter Performance Evaluation Accuracy, NEDO-20340, General Electnc Company, June 1974.

l

3. ANFB Critical Power Correlation, ANF/ EMF-1125 (P) (A), (as supplemented), ,

Advanced Nuclear Fuels Corporation, April 1990.

4. Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19 (P)(A) Volume 1 Supplement 3, Supplement 3 Appendix F, and Supplement 4, t November 1990. '

. 5. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A) Volume 1 (as supplemented)

March 1983.

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! 2.2 1 REACTOR PROTECTION EYSTEM INSTEDMENTATION BRTPOTNTS se Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the inlues at which the reactor trips are set for each parameter.

The Trip Satpoints have been selected to ensure that' the reactor core and reactor coolant system are prevented from == aading their Safety Limits during natual operation and design basis anticipated operational occurrences and to assist in

! mitigating the consequences of arr4dants. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable value is

, acceptable on the basis that the difference between each Trip setpoint and the Allowable value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

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{ 1. Yne m.d4 =en enna. wan4 ter. riu,. utem

. N The IRM system consists of 8 chambers. 4 in each of the reactor trip systems.

The IRM is a 5 decada 10 range instrument. Se trip setpoint of 120 divisions of

scale is active in each of the 10 ranges. Thus as the 3RM is ranged up to i

accommodate the increase in power level, the trip setpoint is also ranged up. Se IRM instruments provide for overlap with both the APRM and SIGE systems.

n e most signifiaast source of reactivity changes sharing the power increase j is due to control rod withdrawal. In order to ensure that the IRM provides j

the required protection, a range of red withdrawal m M d==ts have been analysed.

l The results of these analyses.are in Section 15.4.1.2 of the FIAR. The most severe j case involves an initial condition in teiich THEIGmL pcMER is at approximately 14 of RATED THERMAL PCMER. Addicianal conservatism was taken in this analysis by 1

j *k ' assuming m= resul the 3RM abannel closest to the control rod being withdrawn is bypassed.

y f this analysis show that the reactor is shutdown and peak power is j j limited of RATED TEEIGEL POWER with the peak fuel enthalpy well below the

fuel failure threshold of 170 onl/gn. Based on this analysis, the ERM provides l protection against local control red errors and continuous withdrawal of control rods in sequence and provides h % protection for the APRM. ~

. 2. Avermee power manne Monitor Por operation at low pressure and low flow during STRRTUP, the APAGI scram

• i setting of 15% of RA2ED TREIGEL 70MER provides adequate thermal margin between the l -

setpoint amt. the Safety Limite. Se margin =htes tho anticipated mans:tvers associated with power plant startup. Effects of increasing pressure at sero or low

void content are minor and onid water from sources available sharing startup is not i auch solder than that already in the system. Temperature coefficients are small i and control rod pattazas are constrained by the Riet. Of all the possible sources l l et reactivity input uniform control rod withdrawal is the most probable cause of 4

significant power incrosse.- Because 1

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1 l LA SALLE - 13 TIT 1 3 2-9 1==ad==nt No. 104 i

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REACTIVITY CONTROL SYSTEM 3/4.1.2 REACTIVITY AN0MALIES v l LIMITING CONDITION FOR OPERATION ce tre/ * *#

<* 15

/ /* d*b N' deviation in reactivity from that predicted i larg r than l

expected for normal operation and should therefore evaluated.

==> T e a t ' "v j

MODE 1, of th cont a withdra a toad sta o ion is typ achieved. er ese condi

, the compar between predi and monit to re ivity ro es effective i of the activit . In , con 4ptf rd t ically being withd during a sta ,n MODES 3 4, all control ra a fully inse , , theref reactor s in east active , moni ng i ty is not In

. 5 1 loadi ts in a conti y changing s, ivity.

1 ts .1. ensure fue vesen are perfo vi ( the of safety ly , and SEM ion is requ og the fi startup followi tions that 1d have al reactivity i (e.g. ual vesent, trol i ing).

rep 1 , trol rod

, requ LC0 3.1.1, vid a l rect compari the predi monitored co j ctivity at itions; , reactivit ly not requ dur these ti s.

W ":

should an y develop be asured and radicted core react , the core ty difference t resto vi in the limi continued i is wi n the co design ation

i . Restorat

' n the limit id orund an eval' ion f the design and saf

y. This ana ysis to in A for nation nornelly 7-~ 3w reas-. the core

\ -

y v- e r.ie = - ! .:-: -

.; ;, - /07/er '

POWER DSSTRIBUTION LIMITS (Continued)

,3/4.2.3 MINIMUM CRITICAL POWER RATIO -

SURVEILt.ANCE REQUIREMENTS

• +tt- MCP with  %

4.23.1 ~ l

a. t 8" = 0.86 prior to performance of the initial scram time measurementsl )

. for the cycle in accordance with Specification 4.1.3.2, or j 1

b. t determined within '72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time sur eillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT.

l

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. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at  !

least 15% of RATED THERMAL POWER, and  !

j

c. l 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 oper-  ;

ating with a LIMITING CONTROL ROD PATTERN for MCPR.

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l LA SALLE - UNIT 1 3/4 2-4 Amendment No. 70

Insert #3 4.8.3,%

The applicable MCPR limit shall be determined from the COLR based on: )

a. Technical Specification Scram Spesd (TSSS) MCPR limits, or

b. Nominal Scram Speed (NSS) MCPR limits if scram insertion times determined per surveillance 4.1.3.2 meet the NSS insertion times identified in the COLR.

1 Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of completion of each set of scram testing, the results will be compared against the nominal scram speed (NSS) insertion times specified in the COLR, to verify the applicability of the transient analyses. Prior to initial scram

time testing for an operating cycle, the MCPR operating limits used shall be based on the Technical Specification Scram Speeds (TSSS). -

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INSTRUMENTATION

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VERSING IN-CORE PROBE si m n ,

{ LIMITING CONDITION FOR OPERATION .

! 3.3.7.7. The traversing in-core probe (TIP) system shall be' OPERABLE  ! with: '

5 I.~ Movable detectors, drives and readout equipment to mape core in

the veriuired measurement locations and '

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{ b. Indexing equipment to allow all required detectors be calibrated in a common location. .

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, APPLICABILITY: When the traversing in-core probe is used or:

l a. Recalibration of the LPRM detectors, and J ,

l *b. Monitoring the APLHGR, LHGR, MCPR, or liFLPD.. ~

EllgH

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a. With one or more TIP measurement locations operable, required l measurements may be perfomed as descri in 1 and 2 below, provided the l j

i reactor core is operating in an octant s tric control rod pattern, and the total core TIP uncertainty for the sent cycle has been measured to be less than 8.7 percent.

} 1. TIP data for an inoperable messorement location may be. replaced by 1

data obtained from that . string's redundant j the substitute TIP data was obtained from an(symmetric) operable measurement counterpart if j location. /

/

2. TIP data for an inopera 'e measurement location may be replaced by j

data obtained from a 3 imensional BWR core simulator code normalized with available opera, ng measurements, provided the total number 5cf-

simulated channels asurement locations) does not exceed

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a) All channels of a single TIP machine, or i b) A total o five channel's if more than one TIP machine is l involve . ~

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b. Otherwise, wir the TIP system inoperable, suspend use of the system for the above ap .icable monitoring or calibration functions.

c. The provi ons of Specification 3.0.3 are not applicable.

1 1 l SURVEILL E REOUIREMENTS 4.3.7. The traversing in-core probe system shall be demonstrated OPERABLE by

no izing each of the above required detector outputs within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior j

! o e for the above applicable monitoring or calibration functions.

Only the detector (s) in the required measurement location (s) are required to be OPERABLE.

LA SALLE - UNIT 1 3/4 3-73 AMENDMENT NO. 94 l

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3/4.3.7.7 AND 3/4.3.7.8 INTENTIONALLY LEFT B'. ANY ll l

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

! RECIRCULATION LOOPS -

i

LIMITING CONDITION FOR OPERATION i

! ' 3.4.1.1 Two reactor coolant system recirculation loops shall be in

{ operation.

l APPLICABILITY: OPERATIONAL CONDITIONS I a"d 2 1

j a. With only one (1) reactor coolant system recirculation loop in

! _ operation,' comply with Specification 3.4.1.5 and:

i fed,e, 4 .

Within four (4) hours: .

j 4V5&6E MMI a)

Place ths recirculation flow control system in the Master j //NEAf #54T Manual mode or lower, and .

N #

b) Increase the MINIMIM CRITICAL POWER RATIO (MCPR) Safety l [/)/4//(rA) / jam %) Limit by 0.01Go 1.0B per Specification 2.1.2, and 48 b' # '

c) Increase the MINIMM CAL POWER RATIO (MCPR) Limiting f //s a ition for.0peration.by 0.01 per p ification 3.2.3, ,

i 0 g) Reduce the Average Power Range Monitor (APRM)

! [. 4) (OA6

'" O Rod Block and Rod Block Monitor Trip setpoints{ and cram and -

j4/M gggg7 j

Allowable Values to those applicable to single recirculation loop operation per Specifications 2.2.1 and

{

iL 3.3.6. -

f G)

' 2. ot entise, be in at least HDT SHUTDOWN within the next twelve (12) hours. .

b. With no reactor coolant recirculation loops in operation: '

l

' ~

1. Talre the ACTION required by Specification 3.4.1.5, and l

l

! 2. Be in at least HDT SHUTDOWN within the next six (6) hours.

l

! I l l '

i i 3

! LA SALLE - UNIT 1 3/4 4-1 Amendment No. 103 i i  !

I I

! i i

~ ~ ~ ~ " ~ - ~ ~ ~ ' ~

nem..ivie, tur. i nut a v s e r_m j BASES I a

{ 3/4.1.3 CONTROL RODS (Continued) j l In addition, the automatic CRD charging water header low pressure scrae 1 (see Table 2.2.1-1) initiates well before any accumulator loses its full capa-bility to insert the control rod. With this added automatic scram feature, the surveillance of each individual accumulator check valve is no longer i

' necessary to demonstrate adequate stored energy is available for normal scram action.

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled-and therefore this check must be perforised prior to achieving criticality after j completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity. The subsequent check is performed as a backup to the l initial demonstration.

In order to ensure that the control rod patterns can be followed and I therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

i The control rod housing support restricts the outward movement of a

{ c:ntrol rod to less than 3.65 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount.of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive j h using. ,

t . -

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

314.1.4 CONTROL R0D PRnenaM CONTROLS Control rod withdrawal and insertion sequences are established to assure l that the maximum insequence individual control rod or control rod segments i

which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fual enthalpy greater than 280 cal /gm in the event of a I

control rod drop accident. The specified sequ'ences are characterized by homogeneous scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 105 of RATED THERMA'. POWER, there is no possible rod worth i

which, if dropped at the design rate of the velocity limiter, could result in a

' peak enthalpy of 280 cal /ge. Thus requiring the RWM to be OPERABLE when THERMAL POWER is less than or equal to 105 of RATED THERMAL POWER provides adequate control.

The RWM provide automatic supervision to assure that out-of-sequence rods j will not be withdrawn or inserted.

1 i

The analysis of the rod drop accident is presented in Section 15.4.9 of l the FSAR an

.s _=- ,d the -atechniques s of the analysis are presented in '-dr! 7:;:-'.

i e q ;?-- -+ , a-q c --- : -c L )w.pp.go.jp ng Nuslect mci lNoQ [v 0*Yiy WI** Anndars - kosimic )wr%)r,fu l Dery a. J Analrsix ,

i LA SALLE - UNIT 1 8 3/4 1-4 Amendment No. 89 l

1 J

x- m e , tunexut simm j

i BASES 4

{ 3/4.1.4 CONTROL R00 PROGRAN CONTROLS (Continued) j The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power i operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.. This system backs l up the written sequence used by the operator for withdrawal of control rods.

3/4.1.5 STANDBY LIOUID CONTROL SYSTEM

The standby liquid control system provides a backup capability for

! bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To

! meet this objective it is necessary to inject a quantity of boron which

! produces a concentration of 660 ppe in the reactor core in approximately 50 to j

125 minutes. A normal quantity of 4587 gallons not of solution having a 13.4% l

) sodium pentaborate concentration is required to meet a shutdown requirement of 3%. There is an additional allowance of 255 in the reactor core to account for imperfect mixing. The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required

! pumping rate is 41.2 gps. The minimum storage volume of the solution is j

cstablished to allow for the portion below the pump suction that cannot be

! inserted and the filling of other piping systems connected to the reactor i v:ssel. .

The temperature requirement on the sodium pentaborate solution is i

n:cessary to maintain the solubility of the solution as it was initially mixed l to the appropriate concentration. Checking the' volume of fluid and the j

temperature once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for j injection.

d j

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer j periods of time with one of the redundant components inoperable.

l Surveillance requirements are established on a frequency t' hat assures a i

high reliability of the system. La:e the solution is established, boron concentration will r.ot vary unless more boron or water is added, thus a check on the temperature and volume once each 24' hours assures that the solution is availabic for use.

! Replacement of the explosive charges in the valves at regular intervals i

j will assure that these ' valves will not fall because of deterioration of the charges.

~

l

! 1.

j C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis 1

for Large BWR's," G. E. Topical Report NEDD-10527, March 1972 j 2. C. J. Paone, R. C. Stirn and R. M. Young, Supplement I to NEDD-10527, i

July 1972 J. M. Haun, C. J. Paone and R. C. Stim Addendum 2, " Exposed Cores,"

1 (3. Supplement 2 to NED0-10527, January 1973 LA SALLE - UNIT I B 3/4 1-5 Amendment No. 89 i

l 3 /4. 2 POWER DISTRIBUTION LIMIT 5 BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE ptANAR LINEAR NEAT CENERATION RATE C4 h/

"his specificaties assdras that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit.

specified in 10 CFR 50.45. The specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations.

The peak cladding temperature accident is primarily a function of(the average heat generation rate of all thePCT) follow rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHER 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 in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHER) is this LHGR -

=8 =f divided by- local peaking factor.

of the highest m&GB +~nD However, the~ i. General rte (GE) calculational models (SAFER /GESTR descri in Referen which are consistent with requirements of Ap ix K to 10 CFR 50, have established that.AP are not expect values limited by LOCA/ECCS considerations. APLHGR limits are still required, _ to assure that fuel _ rod mechanical integrity is maintained. They are specified for all r=./ 2 fuel types in the Gore-CW

? :P- ' i:P "

' 6?BMT/NTr1Hdi$n[5l6K1~* Jased on the fuelanical design analysis.

The purpose of the power- and flow-depen MAPLHGR factors specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow and core pcwer conditions. At less than 100% of rated . flow or rated power, the required MAPLHGR is the minimum of either (a) the product of the rated MAPLH9R limit and the powerednt MAPLHER factor or (b) the product of the rated MAPLHGR limit and the flow-dependent MAPLHGR factor. The power- and flow-dependent MAPLHGR factors assure that the fuel remains within the fuel design basis during transients at off-rated conditions. Methodology forestablishingthesefactorsisdescribedinReference#l7

$ o. S E f $ $ E 6

{

l LA SALLE - UNIT 1 B 3/4 2-1 Amendment No. 103 i

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

i

)

i l Insert #4 i

i SPC Fuel

~

This specification assures that the peak cladding temperature of SPC fuel following a postulated design basis loss-of-coolant accident will not exceed the peak cladding temperature (PCT) and maximum oxidation limits specified in 10CFR50.46. The j calculational procedure used to establish the AVERAGE PLANAR j LINEAR HEAT GENERATION RATE (APLHGR) limits is based on a

! loss-of-coolant accident analysis. The analysis is performed using

! calculational models which are consistent with the requirements of i APPENDIX K to 10CFR50. The models are described in Reference 1.

The PCT following a postulated loss-of-coolant accident is primarily

a function of the average heat generation rate of all the rods of a fuel j assembly at any axial location and is not strongly influenced by the

! rod-to rod power distribution within the assembly.

i

} The AVERAGE PLANAR LINEAR HEAT GENERATION RATE i (APLHGR) limits for two-loop operation are specified in the CORE OPERATING LIMITS REPORT (COLR).

l For single-loop operation, an APLHGR limit corresponding to the i product of the two-loop limit and a reduction factor specified in the i COLR can be conservatively used to ensure that the PCT for single.'

loop operation is bound by the PCT for two-loop operation.

i.

/

4 l

J

d, .

i i POWER DISTRIBUTION MMris I

i BASES 1

l 3 /4.2. 2 DELETED

! 3 /4. 2.3 MINIMUM CRITICAL POWER RATfD 4

The required operating limit MCPRs at steady state operating conditions as specifled.in Specification 3.2.3 are derived from the established fuel cladding i

integrity Safety Limit MCPR, and an analysis of abnormal operational i transients. For any abnormal operating transient analysis evaluation with the

initial condition of the reactor being at the steady-state operating limit, it

is required that the resulting MCPR does not decrease below the Safety Limit j MCPR at any time during the transient assuming instrument trip setting given in j 5pecification 2.2. .

To assure that the fuel cladding tateprity Safety Limit is not exceeded during any anticipated abnormal operationa transient, the most limiting transients have been analyzed to determine which result in the largest reduc-tion in CRITICAL POWER RATIO (CPR). 'lha type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, .and j coolant temperature decrease. The limiting transient yields the largest delta l l

MCPR. When added to the Safety Limit EPR, the required minimum operating l

limit MCPR of Specification 3.2.3 is obtained and presented in the CORE n OPERATING LIMITS REPORT. f i Analyses have been performed to determine the effects on CRITICAL POWER (

1 TIO (CpR) during a transient assuming that certain equipment is out of l

service. A detailed description of the analyses is provided in Reference 5.

The analyses performed assumed a single failure only and establishpd the licensing bases to allow continuous plant operation with the analy2ed equipment j out of service. The following single equipment failures are included as part

of the transiant analyses input assumptions

I) main turbine bypass system out of service,

2) recirculation pump trip system out of service, 1

hserf $ 5 kere I

i

l i

2 l

1 j LA SALLE - UNIT I B 3/4 2-2 Amendment No. 103 1

4 i

1 inserts 5 The purpose of the power- and flow-dependent MCPR limits (MCPR,and MCPRt respectively) specified in the CORE OPERATING LIMITS REPORT (COLR) is to define operating limits dependent on core flow and core power. At a given power and flow operating condition, the required MCPR is the maximum of either the power-dependent MCPR limit or the fic; d.pe.Mest MCPR limit. The required MCPR limit assures that the Safety Limit MCPR wB not be violated.

The flow dependent MCPR limits (MCPRr> are established to protect the core fmm inadvertent core flow increases. The cc,re flow increase event used to establish the limits is a slow flow runout to maximum flow that does not result in a scram fmm neutron flux overshoot exceeding the APRM neutron Aux-high level (Table 2.2.1-1, -

ltem 2). A conservative flow control line is used to deAne several core power /Ilow state points at which the analyses are performed. MCPRtlimits are established to support both the automatic and manual modes of operation. In the automatic mode, MCPRtlimits are established to protect the operating limit MCPR. For the manual mode, the limits are set to protect against violation of the safety limit MCPR.

The power-dependent MCPR limits, (MCPRe), are ant =Mahed to pmtect the core from plant transients other than core flow increases, including pressurization and the localized control rod withdrawal error events.

Analyses have been performed to determine the eNocis of assuming various equipment out-of-service scenarios on the (CPR) during transient events. Scenarios were performed to allow continuous plant operation with these systems out of service. Appropriate MCPR limits and/or penalties are included n the COLR for esch of the equipment out-of service scenanos identfled in the COLR. In some cases, the reported Emits or penalties are based on a cycle-independent analysis, while in other cases, analyses are performed on a cf_ ;+ -- basis.

References 2-6 descrbe the methodology and codes used to evaluate the potentially bounding non-LOCA transient events identified in Chapter 15 of the UFSAR.

MCPR limits are presented in the CORE OPERATING UMITS REPORT (COLR) for both Nominal Scram Speed (NSS) and Technical W Scram Speed (TSSS) insertion times. The negative reactivity insertion rate resulting fmm the scram plays a major role in providing the required ,,,vi.CJun against violating the Safety Limit MCPR during transient events. Faster scram insertion times provide greater protection and allow forimproved MCPR performance. The appucation of NSS

, MCPR limits takes advantage of improved scram insertion rates, while the TSSS MCPR limits provide the necessary pivi.CJun for the slowest allowable average scram insertion times identified in bpecification 31.3.3. If the scram insertion times determined per surveillance 4.1.3.2 meet the NSS insertion times, the appropriate NSS MCPR limits iden9 fled in the COLR are applied, if the scram insertion times do not meet the NSS insertion criteria, the TSSS MCPR limits are applied.

A

. POWER DISTRXBUTION SYSTEMS i

BASES hNIMUMCRITICALPOWERRATIO(Continued) l 3) safety / relief valve (5/RV) out of service, and

' 4) feedwater heater out of service (ccrresponding to a 100 degree F reduction in feedwater temperature).

i For the main turbine bypass and recirculation pump trip systems, specific cycle-independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for j Operation (LCO) values are established to allow continuous plant operation with

! these systems out of service. A bounding end-of-cycle exposure condition was

, used to develop nuclear input to the transient analysis model. The bounding i exposure condition assumes a more top peaked axial power distribution than the

nominal power shape, thus yielding a bounding scram response with reasonable l conservatisms for the MCPR LCD values in future cycles. The MCPR LCD values

shown in the CORE OPERATING LIMITS REPORT for the main turbine bypass and i recirculation pump trip systems out of service are valid provided that these j limits bound the cycle specific results.

i j The analysis for main turbine bypass and recirculation pump trip systems l inoperacle allows operation with either system inoperable, but not both at the i same time.

1 . .

1 For operation with the feedwater heater out of service, a cycle specific j analysis will be performed. With reduced feedwater tamperature, the Load Reject i 4

Without Bypass event will be less severe because of the reduced core steaming i

rate and lower initial void fraction. Consequently, no further analysis is

! needed for that event. However, the feedwater controller failure event becomes i more severe with a feedwater heater out of service and could become the limiting l

transient for a specific cycle. Consequently, the cycle specific analysis for the feeewater controller failure event will be performed with a 100 degree F

! feeowater temperature reduction. The calculated change in CPR for that event j- will then be used in determing the cycle specific MCPR LCO value.

I j In the case of a single S/RV out of service, transient analysis results

showed that there is no impact on the calculated MCPR LCD value. The change in CPR for this operating condition will be bounded by reload licensing calcu-lations, and no further anal /ses ce required. The analysis for a single Si.9V i out of service is valid in conjunction with dual and single recirculation loop

! operation. ,

1 i The evaluation of a given transient'begins with the system initial parameters i i shown in FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient

computer program. The codes used to evaluate events are described in

! ~

!t i LA SALLE UNIT 1 B 3/4 2-3 Amendment No. 70 l 1

l f _ _ _ _ _ ._

l I POWER DISTRIBUTION SYSTEMS 1

1 BASES i u l j (

f MINIMUM CRITICAL POWER RATIO (Continued) m/

l l

4 I i NEDE-24011-P-A-US (Reference 4). The outputs of these programs along with the l 1 initial MCPR form the input for further analyses of the thermally limiting i bundle (Reference 4). The principal result of this evaluation is the recuctioni s j in MCPR caused by the transient.

The need to adjust the MCPR operating limit as a function of scram time  !

! arises from the statistical approach used in the implementation of the 00YN l j computer code fcr utlyzing rapid pressurization events. Generic statistical  :

I analyses were performed for plant groupings of similar design which considered )

i the statistical variation in several parameters, i.e., initial power level, i l CRD scram insertion time, and model uncertainty. These analyses, which are described further in Reference 2, produced generic Statistical Adjustment j Factors which have been applied to plant and cycle specific ODYN results to

yield operating limits which provide a 951 probability with 95% confidence

! that the limiting pressurization event will not cause McPR to fall below the fuel cladding integrity Safety Limit.

{

l As a result of this 95/95 approach, the average 205 insertion scram time

aust be monitored to assure compliance with the assumed statistical distribu-

! tion. If the mean value on a cycle cumulative, running average, basis were to i exceed a 5% significance level compared to the distribution assumed in the j as a ODYN functionstatistical analyses, of the mean the MCPR 20% scram time, tolimit mustconservative a more be increasedvalue linearlyIch wh

! reflects an NRC determined uncertainty penalty of 4.45. This penalty is j applied to the plant specific CDYN results, i.e. without statistical adjust-j ment, for the limiting single failure pressurization event occurring at the

limiting point in the cycle. It is not applied in full until the mean of all

! , current cycle 20% scram times reaches the 0.86 seconds value of Specifica-

) i tion 3.1.3. 3. In practice, however, the requirements of 3.1.3.3 would most j likely be reached, i.e. , individual data set average > 0.86 secs, and the required actions taken well before the running averagt exceeds 0.86 secs.

{

! The 5% significance level is defined in Reference 4 as:

l ..

J 1,s,.1..S(N,/b,)2/2 s i

i {' i=1 i

{ where p = mean value for statistical scram time distribution j

. to 20% insertad = 0.672 I e a standard deviation of above' distribution = 0.016 Ng = number of rods tasted at 80C, i.e. , all operable rods i n l INg= total number of operable rods tasted in the

i=1 current cycle

• J LA SALLE UNIT 1 B 3/4 2-4 Amendment No. 70 I f -

i I

l POWER DISTRIBUTION SYSTEMS-i .

j BASES 1

l MINIMUMCRITICALPOWERRATIO(Continued)

1 i The valm for f used in Specification 3.2.3 is 0.687. seconds which is l conservative for the,following reason

l l l

\

i For simplicity le formulating and implementing the LCO, a conservative j a 1 value for I N, of 5g8 was used. This represents one full core data set ,

1-1 . . l st B0C plus one fell core data set following a 120 day outage plus twelve 1

los of core, lg reds, data sets. The 12 data sets are equivalent to i 24 operating months of surveillance at the increased surveillance -

l frequency of one set per 60 days required by the action statements of 3 Specifications 3.1.3.2 and 3.1.3.4. l t

! That is, a cycle length was assumed which is langer than any past or l l contemplated refueling interval and the number of rods tested was maximized in I i order to simplify and conservatively reduce the criteria for the scram time at which MCPR penalizatine is necessary.

The purpose of the power- and flow-dependent EPR limits specified in the i CORE OPERATING LIMITS REPORT is to define operating limits at other than rated f core flow and core power conditions. At a given power and flow operating condition, the required E PR is the maximum of either the power-dependent E PR },

> limit or the flow-dependent EPR limit. The required E PR assures that the

. Safety Limit EPR will not he violated. Methodology for establishing the Qower- and f1= t-- ' t EPR limits is described in Refgnce 6.g At THERMAL POWER levels less than or equal to 25i M RATED THERMAL POWER, the reactor will be operating at minimum recirculaMon pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, oP4 rating plant experience indicates that the resulting EPR value is in ocess of requirements by a considerable margin. During initial start-ar testing of the plant,.a EPR evaluation will be made at 255 of RATED THERhtb POWER level with minimum recirculation pump speed. The llCPR margin will thus be demonstrated such that future MCPR evaluation below this power level wilr be shown to be unnecessary.

The daily requinment for calculating EPR when THERMAL POWER 1: gre.:ter than or equal to 255 of RATED THERMAL POWER is sufficient since power distribution shifts are very slow uhen there have not been significant power or control rod changes. The requirement for calculating EPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

LA SALLE - UNIT I B 3/4 2-5 Amendment No.103

j POWER DISTRIBUTION SYSTDes BASES -

! 3/4.2.4' LINEAR HFAT GENERATION RATE i 4E h/

Tie specification assures that the LINEAR HEAT GENERATION RATE (LNGR) in

any rod is less than the desien linear heat men ration even if fuel sellet .

{ densifwation is sontulatedd a r ty 3ptcifi is ed th i al 's p en 'n ett 3. . of icaVrepo 73

! une 1 arl nere ng v nati n 1 s e i o t and pa ass

• a con ence at no an ne j ~) exc s des LI c_ ION d to r 'ikia Refergnces:

L j'b Sed N bere]

General Doctric Company Analytical Model for. Loss-of-Coolant i

[I Analysis 'in Accordance with 10 CFR 50, Appendix K, NED0-20566A, j September 1986.

! 2. " Qualification of the One-Dimensional Core Transient Model for

Boiling Water Reactors," General Doctric Company Licensing Topical

! Report NED0 24154 Vols. I and II and NEDE-24154 Vol. III as sup-i plemented by letter dated September 5,1980, from R. H. Buchholz

! (GE) to P. 5. Check (NRC).

J i 3. "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-i Coolant Accident Analysis," General Doctric Company Report NEDC-32258P, October 1993.

4. " General Electric standard' Applica' tion for Reactor Fuel,"

{ NEDE-24011-P-A (latest approved revision).

5. " Extended Operating Domain' and Equipment Out-of-servicegfor LaSalle l County Nuclear Station Units 1 and 2," NEDC-31455, November 1987.

l 6. " ARTS Improvement Program . Analysis for LaSalle County Units 1 and 2,"

l G

k eneral Doctric Company Report NEDC-31531P, December 1993.

l /L.sse/ # '7 4eee]

i 1Ae eM*dr & A/ de.,r.-#cd., are /;seu.,f r fle Go cral E/edtrc sk da<d &A'c /,, ,

-for fo* doe Fue/ (Gss74p) Asar- 29oy-/-4.

[Ae & ESTAR hscuses //e, he7f48 used fo~ .

Sisure L//GA remi r beb~ tb / erg Ly.

j 4

LA SALLE - UNIT 1 B 3/4 2-6 Amendment No. 103

)

Insert #6 l

l' SPC Fuel The Linear Heat Generation Rate (LHGR) is a measure of the heat generation rate per unit length of a fuel rod in a fuel assembly at any axial location. LHGR limits are specified to ensure that fuel integrity limits are not exceeded during normal operation or anticipated l operational occurrences (AOOs). Operation above the LHGR limit l followed by the occurrence of an AOO could potentially result in fuel damage and subsequent release of radioactive material. Sustained operation in excess of the LHGR limit could also result in exceeding the fuel design limits. The failure mechanism prevented by the LHGR limit that could cause fuel damage during AOOs is rupture of the fuel rod cladding caused by strain from the expansion of the fuel ,

pellet. One percent plastic strain of the fuel cladding has .been '

defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur. Fuel design evaluations are performed to demonstrate that the mechanical design limits are not exceeded during continuous operation with LHGRs up to the limit defined in the CORE OPERATING LIMITS REPORT. The analysis also includes allowances for short tann transient operation above the LHGR limit.

At reduced power and flow conditions, the LHGR limit may need to be reduced to ensure adherence to the fuel mechanical design bases during limiting transients. At reduced power and flow conditions, the LHGR limit is reduced (multiplied) using the smaller of either the flow-dependent LHGR factor (LHGRFAC,) or the power-dependent j LHGR factor (LHGRFAC,) corresponding to the existing core flow l and power. The LHGRFAC, multipliers are used to protect the core during slow flow runout transients. The LHGRFAC, multipliers are  !

used to protect the core during plant transients other than core flow transients. The applicable LHGRFAC, and LHGRFAC, multipliers are I specified in the CORE OPERATING LIMITS REPORT. i

i l

l Insert #7 j .

! 1. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors j EXEM BWR ECCS Evaluation, ANF-91-048(P)(A), Advanced Nuclear Fuels j Corporation, January 1993.

2. Exxon Nuclear Methodology for Bolling Water Reactors, Neutronic Methods for Design and Analysis, XN-NF40-19 (P)(A), Volume 1 (es supplemented) Exxon  !

j Nuclear Company, March 1983.

4

3. Exxon Nuclear .'_':M-RR-yy for BoBng Water Reactors, THERMEX Thermal 4

Limits Methodology Summary Description, XN-NF40-19 (P)(A), Volume 3

Revision 2 (as supplemented), Exxon Nuclear Co.. re iry, January 1987.

l 4. Exxon Nuclear Plant Transient Methodology for BoBng Water Reactors, XN-NF-

79-71 Revnion 2 (P)(A) (as supplemented), Exxon Nuclear Company, March j 1988.

l 5. COTRANSA2: A Computer Program for Bolung Water Reactor Transient Analyses, ANF-913(P)(A) Volume 1 Revision 1 (as supplemented), Advanced Nuclear Fuels Corporation, August 1990.

8. XCOBRA-T: A Computer Code for'BWR Transient Thermal-HydrauBc Core

! Analysis, XN-NF44-105(P)(A) Volume 1 (as supplemented), Exxon Nuclear j Company, February 1987. . .

7. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF49-98(FXA)

Revision 1 (as supplemented), Siemens Power Corporation - Nuclear Division, l

1 May 1995.

! 8. LaSane County Station Units 1 and 2 SAFER /GESTR - LOCA Loss-of-Coolant l Accident Analysis, NEDC-32258P, General Electric C ,virany, October 1993.

$ 9. ARTS improvement Program analysis for LaSaNo County Station Units 1 and 2, NEDC-31531P, General Electric Company, December 1993.

I .

1 1

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i l

l

1 TNfvstmfrNv1TTON e mAers I l

j 1/d 1 A WFf*TRL"UT.17 TON FDNF TRTF ACTUATTON TutvWTfMFWy1vTffM q

f i The anticipated transient without scram (A'ars) recirculation pump trip system provides a means of limiting the consequences of the unlikely 1 occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of l j ' study ev d ts in General Electric company Topical Report NEDO-10349, dated j j March 1971 and NEDO-24222, dated December, 1979, and, Appendix G of the FSAR. I i

i The end-of-cycle recirculation pump trip (EDC-RPT) system is a part of

! the Beactor Protectice system and is an essential safety supplement to the j reactor trip. The purpose of the EDC-RPT is,te recover the loss of thermal

! margin which occurs at the end-of-cycle. The physical phenomenon involved is i that the void reactivity feedback due to a pressurization transient can add I -

positive reactivity to the reactor system at a faster rate than the control i rods add negative scram reactivity. Each EDC-RPT system trips both recircula-

} tion pumps, reducing coolant flow in order to reduce the void collapse in the 1 core during two of the most limiting pressuriaation events. The two events

! for which the EOC-RPT protective feature will functica are closure of the i turbine stop valves and fast closure of the turbine crentrol valves.

l --

l A eric ya , whi ovid or coat;, stied on wi one o

th ip ens o e FT sys ,e, has per ed. e

j. ana sis armi ho cyc1 www- FOWER 0

I PE) t endi for ation ( ) valu which t be if a

OC- syst is rahle. ese es that to r eti mar to e McFR afety 1 t exis in the t of anal d tr isnt l w the on in able. anal resul are fu er cuss the as f speci tion .3.

j i A fast closure sensor from each of two turbine control valves provides j input to the EOC-RFT systemt a fast closure sensor from each of the other two j

]

turbine contrci valves provides input to the second EDC-RPT system.

similarly, a position switch for each of two turbine stop valves provides

! input to one EDC-RPT systems a position switch from each of the other two stop j valves provides input to the other EDC-RPT system. For each EDC-RPT system, l 4

the sensor relay sentacts are arranged to form a 2-out-of-2 logic for the fast l j closure of turbine control valves and a 2-out-of-2 logic for the turbine stop '

j valves. The operstian of either logic will actuate the EOC-RFT system and

{ trip both recirculation pumps.

1 j Each EOC-RFT system may be manually bypassed by use of a keyswitch which

• j is administrative 1y controlled. The manual bypasses and the automatic j Operating typass at less than 30% of EATED THERNAL FOWER are annunciated in the control room.

Specified' surveillance intervals and surveillance and maintenance outage times

} have been determined in accordance with the followings i

l 1. MEDC-30851F-A, *Teckuncal specification Ia.irovement Analyses for BWR j Reactor Protection System *, March 1988.

i j -i L

l j

(Z ssert # 7 a

l i

1

(

i j LA ELLLE - UNIT 1 3 3/4 3-3 m=-a ht no. 104 1

1 Insert #8 Analyses were performed to support continued operation with one or both trip systems of the EOC-RPT inoperable. The analyses provide MINIMUM CRITICAL POWER RATIO (MCPR) values which must be used if the EOC-RPT system is inoperable. These MCPR limits are included in the COLR and ensure that adequate margin to the MCPR safety limit exists with he EOC-RPT function inoperable. Application of these limits are discussed further in the bases for .

Specification 3.2.3.

I i

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INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.5 ACCIDENT MDNITORING INSTRUMENTATION The OPERABILITY of the accident monitorint instrumentation ensures that sufficient information is available on selectec plant and assess important variables following an accident. parameters to monitorThis capa sistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons j Learned Task Force Status Report and Short-Tern Recommendations." ,

'~

l 3/4.3.7.6 SOURCE RANGE MONITORS The sourco range monitors provide the operator with infomation of the status W the neutron level in the core at very low power levels during startup 4 ,

i and shutdown. At these power levels reactivity additions should not M made i without this flux level information a,vailable to the operator. When the inter-i mediate range monitors are on scale adequate in - ;.i .. ,'s available without the SRMs and they can be retracted. =a cL.A 3 l f 3/4.3.7.7 JTRAVERSING IN-CORE ' PROBE SYsiuV I The OPERABILITY of the traversing in-core probe (TIP) system with the Ispecified minimum complement of equipment ensures that the measurements  !

I ebtained from use of this equipment accurately represent the spatial neutron i flux distribution of the reactor core.

The specification allows use of substituted TIP data from symmetric l channels if the control rod pattern is symmetric since the TIP data is adjusted

{ by the plant computer to remove machine dependent and power level dependent 2

bias. The source of data for the substitution may also be a 3-dimensional BWR l i

core simulator calculated date set which is normalized to available real data.

Since uncertainty could be introduced by the simulation and nomalization i

process, an evaluation of the specific control rod pattern and core operating '

state must be performed to ensure that adequate margin to core operating limits j Q maintained.

l 3/4.3.7.8 DELETED 1

{ 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capa-bility is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the potential for damage to safety-related-equipment and is an integral element in the overall facility fire protection program.

~

i j

In the event that a portion of the fire detection instrumentation is t

inoperable increasing the frequency of fire watch patrols in the affected areas is re, quired to provide detection capability until the inoperable j

instrumentation is restored to OPERABILITY.

J

~

j LA SALLE - UNIT 1 B 3/4 3-5 Amendeent No. 85 i

i i

ADMINISTRATIVE CONTROLS Semiannual Radioactive Effluent Release Renort (Continued)

Any changes to the 0FFSITE DDSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the '

change In addition, a report of any major changesto(s) thewas made effective.

radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

1

5. tore Dneratine timits Renart

! a. Core operating limits shall be established and documented in the CORE OTRATING LIMITS REPORT before each reload cycle or eny i

remaining part of a reload cycle for the following:

i E #e<* o f (1) The Average Planar Linear Heat Generation Rate (APLNGR) for Technical specification 3.L I.

i o,,,/yzal c~~ &D 3 '

p~

' (2) The minimus Critica gI,,(a M .p 1 -n. ti= , Mr '-' , Ratio ont MCPR (MCPR lim)its'tand xt,1, power and

{

' g a.a flow

=.R.@deendentMM ts) for Technical $pecification (3) The Linear Heat3.2.4.

Specification Generation Rate (LHER) for Technical (4) The Rod Block Monitor Upscale Instrumentation setpoints for Technical Specification Table 3.3.5-2.

i b. The analytica1' methods used to determine the core operating i

! / limits shall be those previously reviewed and approvou by the NRC in the latest approved revision or supplement of the topical

[73ferf-79 j reports describing tie methodolo Unit I, the topical reports are:gy. For LaSalle County Station

{

L 417 NEDE-240!I-P-A, " General Electric Standard Application for (if/ Reactor Fuel,' (latest approved revision). .

.g f Commonwealth Edison Tonical Report NFSR-0085, " Benchmark of y) BWR Nuclear Design Mottods," (latest approved revision).

l

.pr comm.anwealth Edison Topical Re ort NFSR-0085, Supplement L

. ' Benchmark of BWR Nuclear Desi Methods - Qcad Cities N Gamma Scan Comparisons " (late t approved revision).  !

e ,,pr Commonwealth Edison Topical Report NFSR-0085, Supplement 2, g Licensing Analyses," (latest approved revision)." Benchma

\ (p) r%.e e-lN Edison TcfI c*/ AePoff NMWb l .<geajag o f CAM 0/hKroBOM BW Nuclece ycy y flo/.r, (/a/esf offw/wwo.2 LA SALLE - UNIT 1 6-25

Amendment No.103

}

4

4 j insort #9 i

t i

l 1. ANFB Critical Power Correlation, ANF/ EMF-1125(P)(A), (as supplemented).

l 2. Letter, Ashok C. Thadani(NRC) to R. A. Copeland (SPC)," Acceptance for  ;

i Referencing of ULTRAFLOW" Spacer on 9x9-IX/X BWR Fuel Design, l j July 28,1993.

i l 3. Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling l

Water Reactors, ANF-524(P)(A) Revision 2 (as supplemented). l 1

j

4. COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis, ANF-913(P)(A), Volume 1, Revision 1 (as supplemented).

i j 5. HUXY: A Generalized Multirod Hestup Code with 10 CFR 50, Appendix K j Hestup Option, ANF-CC-33(P)(A)(as supplemented),

f 6. Advanced Nuclear Fuel Methodology for Boiling Water Reactors, j XN-NF40-19(P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and .

j Supplement 4.

i 7. Exxon Nuclear Methodology Boiling Water Reactors: Application of the ENC i Methodology to BWR Reloads, XN-NF40-19(P)(A), Volume 4, Revision 1,

June 1986,

8. Exxon Nuclear Methodology for Botng Water Reactors THERMEX: Thermal

. Limits Methodology Summary Description, XN-NF-80-19(P)(A), Volume 3, Revision 2, January 1987, t

9. Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF4547(P)(A)

10. Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced

. Nuclear Fueis Corporation 9x9-IX and 9x9-9X BWR Reload Fuel, ANF49-014(P)(A), Revision 1 (as suppiemonted).

11. Volume 1 - STAIF - A Computer Program for BWR Stability in the Frequency 1 .

Domain, Volume 2 - STAIF - A Computer Program for StVR Stability in the Frequency Domain, Code Qualificabon Report, EMF-CC-074(P)(A).

I 12. RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, XN-NF41-58(P)(A), Revision 2 (as supplemented).

j 13.XCOBRA-T: A computer Code for BWR Transient Thermal-Hydraulic Core

Analysis, XN-NF-84-105(P)(A), Volume 1 (as supplemented).

14. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors

EXEM BWR Evaluation Model, ANF-91-048(P)(A).

1 15. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for

, Design and Analysis, XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2,

Exxon Nuclear Company, Richland, WA 99352, March 1983.

l

Inserf #9 (continued)

16. Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-71(P).

17. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A).

l 1

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i s

l

DESIGN FEATURES 1

i g'$ sv? *136d'b 1

5.3 REACTOR CORE

./ p /g FUEL ASSEMBLIES  !

5.3. he reactor core shall contain 764 fuel assemblies. Each assembly onsists of a matrix of Zircalloy clad fuel rods with an initial composition i of slightly enriched uranium dioxide, UD . Fuel assemblies shall be limited l

tothosefueldesignsapprovedforusei$ SWR's. a

k CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 crucifors shaped control rod assenclies. I

' The control material shall be baron carbide power (8 C) and/or hafnium metal. -

The control rod assemoly shall have a nominal axial $bsorber length of 143 inches.

1 5.4 REACTOR COOLANT SYSTEM J

j DESIGN PRESSURE AND TEMPERATURE

! 5.4.1 The reactor coolant system is designed and shall be maintained: '

s. In accordance with the code requirements specified in Section 5.2

1. of the FSAR, with allowance.for normal degradation pursuant to the applicable Surveillance Requirements, f b. For a pressure of:

4 i

1. 1250 psig on the suction side of the recirculation pumps.

2. 1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3. 1500 psig from the discharge shutoff valve to the jet pumps.

c. For a temperature of 575'F. .

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation

. system is

• 21,000 cubic , feet at a nominal T ,, of 533*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

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l , LA SALLE - UNIT 1 .

5-4 Amendment No. 70

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

Design Features 4.0 4.0 DESI N -

4. Site Loca oca 4.2 r Co 4.2.1 uel Anna lian 7/4' The reactor shall contain 1 assemblies. Each assembly shall consist of a matrix ircalley - ?"4 ." 1 rods with an initial camposition of na ni er slightly enriched uranium dioxide (UD,) as fuel materia ( nf -+ 7:f:L timited substitutions of zirconium al r or stainless steel filler rods for fuel rods, in accordance approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limitad to those fuel designs that have analyzed with applicable lett staff approved codes and astbods shown by tests or analyses to comply with all safety design bases. limited number of lead test assemblies that have not campi representative testing may be placed in non11 siting core regions.

4.2.2 conten1 an M 6 a 11.s de- "!3 # NV J The actor core amblies. The cent shall containteria[l shall be193].crucw orn ide,shaped hafnimeceWirol red tal) as approved Igic. ,

x /

Fuel Storag 4.3.1 tiemlity 4.3.1.1 The fuel storage racks igned and shall be

[

a.

mai F 1 assemblies havi with:

marisam [k-infinity of ]in moreal reactor confiyation at 'mid itions]

[averageU-55 chment of '4.5] meight

);

b. k s 0.95 fully flooded with tad water, which

. i,ncludes allosence for ies as described la

, 7[5ecti .1 of the FIAR];

, (continued)

SWR /6 STS- 4.0-1 Rev 1, 04/07/95

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J lEDfl DEFINITIONS SECTION 7 E8SE 4

1.0 DEFINITIONS 1-1

! 1.1 ACTI0N...........................................................

1-1

^

1.2 AVERAGE PLANAR EXPOSURE..........................................

RATE....................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION 1-1

1.4 CHANNEL CALIBRATION..............................................

1-1 1.5 CHANNEL CHECK....................................................

J 1-1

1.6 CHANNEL FUNCTIONAL TEST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 -2

1.7 CORE ALT ERAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4

1-2 i y

1.8 CORE OPERATING LIMITS REP 0RT.....................................  !

1-2 l

! 1.9 CRITICAL POWER RATI0............................................. l 4 1-2 1.10 DOSE EQUIVALENT I-131............................................

l 1-2 1.11 E-AVERAGE DISINTEGRATIONENERGY..................................

l 1-2 TIME...............

l 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE 1-2 l p.13 END-0F-CYCLE RECIRCULATION PUMP TRIP " SYSTEM RESPONSE TIME..../...

! b.14 FRACTION OF LIMITING POWER DENSITYg 8.?.'.U. .?. .1-3. . . . . . . . . . .

1-3

1.15 FRACTION OF RATED THERMAL P0WER..................................

1-3

, 1.16 . FREQUENCY N0TATION...............................................

1-3 1.17 GASE0US RADWASTE TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

l 1-3

' 1.18 IDENTIFIED LEAKAGE...............................................

1-3 TIME...................................

1.19 ISOLATION SYSTEM RESPONSE i' 1-3 l.20 DELETED..........................................................

1-4 l 1.21 LIMITING CONTROL R0D PATTERN.....................................

i 1-4 1.22 LINEAR HEAT GENERATIONRATE......................................

1-4 TEST............... p ...................

1.23 LOGIC SYSTEM FUNCTIONAL 1.24 BUM FRACTION OF LIMITING POWER DENS %

"ETN...... 1-4 1-4 j 1.25 MEMBER (5) 0F THE PUBLIC..........................................

1-4 RATI0.....................................

! 1.26 MINIMUM CRITICAL POWER s

i I Amendment No. 95 1 LA SALLE - UNIT 2 1

1

i i INDEX 1

i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS 1,

.! SECTION i PAGE .

, 3/4.3 INSTRUMENTATION 1

4 3/4.3.1 J REACTOR PROTECTION SYSTEN INSTRtMENTATION. . . . . . . . . . . .3/4 . . .3-1 j 3/4.3.2 4 ISOLATION ACTUATION INSTRLMENTATION. . . . . . . . . . . . . . . . . .3/4 . . .3-9

) 3/4.3.3 1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...... 3/4 3-23

! 3/4.3.4 4

RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION i

ATWS Recirculation Pump Trip System Instrumentation..........

1 3/4 3-35 End-of-Cycle Recirculation Pump Trip System 4

. Instrumentation............................................ 3/4 3-39 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION

.i INSTRUMENTATION............................................ 3/4 3-45 j 3/4.3.6 f ' CONTROL R00 WITHDRAWAL BLOCK INSTRLMENTATION. . . .3/4 . . .3-50 f 3/4.3.7 MDNITORING INSTRtmENTATION 2

Radiation Monitoring Instrumentation......................... 3/4 3-57 Seismic Monitoring In'strumentation...........................

~

3/4 3-60 t

)

Meteorological Monitoring Instrumentation. . . . . . . . . . . . . . . . . . . . 3/4 3-63 j

L Remote Shutdown Monitoring Instrumentation................... 3/4 3-66 i

Accident Monitoring Instrumentation.......................... /4 3-69 l Source Range 1

- - - . - Monitors........................................

- 3/4 3-72 Traversir.g In-core Probe System..............................

/4 3-73 )

j

_W I _.be-l = k #

Deleted....................................... ~............... 3/4 3-74 i '

i i

Fire Detection Instrumentation............................... 3/4 3-75 Deleted......................................................

1 i 3/4 3-81

! Exploxive has Moni tori ng Instrumentation. . . . . . . . . . . . . . . . . . . . . 3/4 3-82 4,

3 i

Loose-Part Detection System.............................'..... 3/4 3-85 3/4.3.8 FEEDWATER/ MAIN TURRINE TRIP SYSTEM ACTUATION INSTRUMENTATION............................................. 3/4 3-86 J

LA SALLE - UNIT 2 V 1 Amendment No. 69 e

i I

i

i j DEFINITIONS CORE ALTERATION

1. 7 CDRE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incere instruments or reactivity controls within the reactor pressure vessel with the vessel' head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT

1. 8 The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. ,

These cycle-specific core operating limits shall be determined for each  !

reload cycle in accordance with Specification 6.6. A.6. Plant operation within these operating limits is addressed in individual specificptions.

CRITICAL POWER RATIO g m er / CN 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of hat power in the l assembly which is calculated by application of the correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

005E EQUIVALENT I-131 1.10 005E EQUIVALENT I-131 shall be that concentration of I-131, i microcuries/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131 I-132. I-133,1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shalT be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

I-AVERAGE DISINTEGRATION ENERGY . .

1.11 I shall be the average, weighted in proportion to the concentration of I each radionuclide in the reactor coolant at the time of sampling, of 05e sus of the average beta and gassa energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIE 1.12 The EMERGENCY CORE C00 LING SYSTEM (ECCS) RESPONSE TIME shall be that time  !

interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of perfaming its safety function, i.e., the valves travel to their required

, positions, pump discharge pressures reach their required values, etc.

Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is asasured. -

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TI E 1.13 The END-0F-CYCLE RECIRCULATION PL99 TRIP SYSTEM RESPONSE TIME shall be l that time interval to energization of the recirculation pump circuit breaker trip coil from when the monitored parameter exceeds its trip setpoint at the channel sensor of the associated:

a. Turbine stop valves, and

b.

• Turbine control valves.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LA SALLE - 11 NIT 2 1-2 Amendment No. 54

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

(

DEFINITIONS r m k

i

? - ./~ ^

Q E c. tty'~

[ FRACTION OF LIMITING POWER 'r DENSITY [ w 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing

at a given location divided by the specified LHGR limit for that bundle j type.

I FRACTION OF RATED THERMAL POWER 1

i 1.15 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured T POWEL divided by the RATED THERMAL POWER.

t 1 FRE00ENCY NOTATION 4 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM l

1.17 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary ' '

coolant system offgases from the primary system and providing for delay l or holdup for the purpose of reducing the total radioactivity prior to i release to the environment. )

) l

) IDENTIFIED LEAKAGE 1.18 IDENTIFIED LEAKAGE shall be:

f 4 a. Leakage into collection systems such as pump seal or valve i packing leaks, that is captured,and conducted to a sump or t

collecting tank, or l

i b. Leakage into the containment atmosphere from sources that are  !

4 both s)ecifically located and known either not to interfere with tie operation of the leakage detection systems or not to i be PRESSURE BOUNDARY LEAKAGE.

j l ISOLATION SYSTEM RESPONSE TIME l 1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the l

i channel sensor until the isolation valves travel to their required

! positions. Times shall include diesel generator starting and sequence ~

! loading delays wnere applicable. The response time may be measured by any j series of sequential, overlapping or total steps such that the entire i response time is measured.

1.20 DELETED i

l i

l 2 . , , . , , , , . . . 1_3 Amendment No. 95 1

J l

DEFINITIONS 1

- LIMITING CONTROL ROD PATTERN 1

j 1.21 A LIMITING CONTROL ROD PATTERN shall c be a pattern which results in the i core being on a thermal hydraulic limit, i.e., operating on a limiting i value for APLHGR, LHGR, or MCPR.

i LINEAR HEAT GENERATION RATE l l

4 j 1.22 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over Jhe heat trans.fgr //0 Mc l

ef L/fGS toarea

ts fwlassociated SMc

Iw R~-+, with asthe unitred sincr+ 13ngth.lbl@

7 -Me CME 09649T is wh fr ULMV8 LOGIC SYSTEM FUNCTIONAL TEST #E/b/Z l.23 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components,  !

i.e., all relays and contacts, all trip units, solid state logic elements, 4 etc. of a logic circuit, from sensor through and including the actuated l

! device to verify OPERABILITY. THE LOGIC SYSTEM FUNCTIONAL TEST may be l l performed by any series of sequential, overlapping or total system steps j j

such that the entire logic system is tested. I

! MAXIMUM FRACTION OF LIMITING POWER DENSITY 1

! 1.24 he MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the  ;

l highest value of the FLPD which exists in the core.

l

! MEMBER (S) 0F THE PUBLIC

- 1.25 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupation-ally associated with the plant. This category does not include employees i of the licensee, its contractors, or vendors. Also excluded frpm this

category are persons who enter the site to service equipment or to make I deliveries. This category does include persons who use portions of the i site for recreational, occupational, or other purposes not associated with j, the plant.

, MINIMUM CRITICAL POWER RATIO i.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l exists in the core.

OFFSITE DOSE CALCULATION MANUAL

! 1.27 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of

the Environmental Radiological Monitoring Program. The ODCM shall also

contain (1) the Radioactive Effluent Controls and Radiological 4

Environmental Monitoring Programs required by Technical Specification i Section 6.2.F.4 and (2) descriptions of the information that should be i included in the Annual Radiological Environmental Operating and Semi-

! Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

4 LA SALLE - UNIT 2 1-4 Amendment No. 87 ll

l i

! 2.1 SAFETY L M TS l

1 l l BASES i I .

i l

! The fuel cladding reactor pressure vessel, and primary systas piping are the principal barrlers to the release of radioactivo estarials to the I

environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel .

! cladding integrity Safety Limit is set such that no fuel damage is calculated l 1

to occur if the limit is not violated. Because fuel de mge is not directly i i

observable, a step-back approach is used to establish a safety Limit such that -

the MCPR is not less than LO7 for two recirculation loop operation and 1.08 for

single recirculation loop operation. . MCPR praater than LO7'for two recircula-tion loop operation and 1.08 for single ree' reulation liep operation mpresents l a conservative margin relative to the conditions required to maintain fuel

cladding integrity. The fuel cladding .is one of the i separate the radioactive materials from the environs.pQical barriersofwhich he integrity th i I cladding barrier is related to its reistive freedse from perforstions or is ,

I' l cracking. Althouph some corrosion et use related cracking any occur during i the life of the c adding, fission product migration free this source is incre .

i mentally cumulative and continuously measurable. Fuel cladding perforations  ;

• however, can result from thersal stresses which occur from reactor operation, l

• s fonificantly above design conditions and the Limiti Safety System settings. '

W 1e fission product aiFion free cladding perfe ion is just as asesurable .

I the therme11y caused cladding perforations i

, as signalthat from use a threshold related beyond cracking,ill which st greater therus1 stresses any cause - gros

rather than incremental cladding deter oration. Therefore the fuel cladding i i

' Safety Limit is defined with a margin to the conditions 'shIch would produce onset of transition boiling, ICPR of LO. These conditions mpresent a signif-icant departure from the condition intended by design for planned opersti .' l 2.1.1 THE1tMAL POWER. Low Pressure or Low Flow l r =

/~

f The use of the GEXL correlation is not valid for all critica1 puer '

calculations at pressures below 785 psig er core flows less than 205 of rated i flow. Therefore the fuel cladding integrity safety Limit is established by ,

) other means. Thls is done by establishing a limitinp condition en core THERMAL i basis.

l POWER 1s withallthe essentially following!1on head, the core pressure drop at low power andS elevet 1 '

flows will alueys. be- than 4.5 psi. Analyses show that with a bundle flow of 28 x 30" 2bs bundle pressure drop is nearly independent of bundle power and has.a value L E psi. Thus, the bundle flow with a 4.5 psi driving  ;

head will be greater then 2B x 10s 2bs/ttr. Full scale ATLA 5 test data taken j at pressures from 24.7 psia ta 800 1a indicate that the fuel assembly critical i power at this flow is as 3.35 fttt. tiith the desi king factors,

this corresponds to a of more than 505 of IRA POWER.

l Thus, a THDDEL POWER limit of 255 of RATED THERMALPOWEt for reacter pressure l below 785 psig is conservativa.

i.;

~

d I

_bd # 1 -

LA SALLE - UNIT 2 B 2-1 W No. 2 l

1

i. _

1 i

i 4

, insert #1 l ti n is not For certain conditions of pressure and flow, the ANFB cor

! valid for all critical power calculations. The ANFB t t a corecorre flowaof bundle mass velocities less than 0.10 X 10' lbs/hr-ft" the fuel cladding l

tablishing (equivale i less than 10%) or pressures less than 590 psia. b i g basis. Since Therefore, l integrity Safety Limit is established by ll other means. This head, the core f ti

! a limiting condition on core THERMAL ter than 4.5 psi.

t ly a mass POW pressure drop at low power and flows willl always independent beofgrea Analyses show that with a bundle flow of 28 ith aX 4.510' psi lbsthr (

j velocity of 0.25 X 10' lbsthr-ft'),S bundle test data taken press l r

driving head will be greater than 28 X 10' lbsthr. Full 4cale A /

at pressures from 14.7 to 800 psia indicate that the fuel assemW this flow is approximately 3.35 Mwt. f RATED THERMAL l POWER for l corresponds to a THERMAL POWER of mo

! reactor pressure below 785 poig is conservative. -

1 d

i l

i t

i

!4 1

i

)

i . . .

i s'

SAFETY LIMIT _5 BASES f

2.1.2 THERMAL POWER. Hioh Pressure andl damage Hioh Flow The fuel cladding integrity Safety Limit Since the is set such that no parameters during reactor operation, f is calculated to occur i which result in fuel damageif are thenot limit is not directly violated.

from nucleate observable fuel damage

)

the theriaal and itydraulic conditions resulting in a departure i l

boiling have been used to mark the beginning of the could would occur. not necessarily result in damage to BWRd fuel ted as a convenient rods, j t and i which boiling transition is calculated to occur hasin been a op an uncertainty l

! in the procedures used to calculate the critic limit. Therefore, the fuel cladding integrity l for which l in the value of the critical power. to avoid boiling

? Safety Limit is defined as the CPR in the limiting e and all fuel assemb y j

l acre than 99.9% of the fuel rods in the core are ex l uncertainties.

l i ed using the General Electric llThermal of the Analysis The Safety Basis, LimitGETAB"MCPR , which is a statistical is model determ that dn tocombines calculate a I l

i uncertainties critical power.

in operating parameters and the procedure

! determined using the General Electric CHtical' Quality (X) Bo EXL correlation.

l Sef' f W h 6/~C )

e l tion

~.

a

" General Electric BWR Theriaal Analysis Bases (GETAB) D and Design Application," NEDD-10958-A.

Amendment No. 41 B 2-2 LASALLE - UNIT 2

Insert #2 The Safety Limit MCPR is determined using the ANF Critical Power Methodology for boiling water reactors (Reference 1) calcu! ate critical power. The probability of the occurrence of boiling transition is determined using the SPC <leveloped ANFB critical power correlation.

The bases for the uncertainties in system-related parameters are presented in NEDO-20340, Reference 2. The bases for the fuel-related uncertainties are found References 1,3-5. The uncertainties used in the analyses are provided in the cycl) specific transient analysis parameters document. \

1. Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors / Advanced Nuclear-Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects /NRC Correspondence, XN-NF-524 (P)(A) Revision 2 (as supplemented) November 1990.

2. Process Computer Performance Evaluation Accuracy, NEDO-20340, General Electric Company, June 1974.

3. ANFB Critical Power Correlation, ANF/ EMF-1125 (P) (A), (as supplemented),*

Advanced Nuclear Fuels Corporation, April 1990.

4. Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19 (P)(A) Volume 1 Supplement 3, Supplement 3 Appendix F, and Supplement 4 November 1990.

5. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A) Volume 1 (as supplemented)

March 1983.

• Until ANF/ EMF-1125 Supplement 1 Appendix C (ANFB Critical Power Correlation Application for Co-Resident Fuel) is approved by the NRC, cycle specific evaluations are submitted (e.g. EMF-96-021, Application of the ANFB Critical Power Correlation to Co-Resident Fuel for i.aSalle Unit 2 Cycle 8).

l l

1 I

SAFETY LIMITS 8ASES

-w

~

, I ERMAL POWER. High Pressure and High Flow (Continued) l Thebgsasfortheuncertaintiesinthecoreparametersaregiv l

in NED0-10958-A,the NED0-20340 and The powerbasis distribution for the is based uncer on a na typical 764 assembly k ed power-i core in which the rod pattern was arbitrarily chosen to produce a s ew l

I distribution levels.

having the greatest number of assembl as the distribution used in the analysis.

I ~_

i  !

i l

}

i i

i .

i j .  !

, u I

1 l

l .

1 l

t i l

l I

~

a.

" General Eiectric BWR Themal Analysis Bases (GETA8) Data, .

and Design Application," NEDD-10958-A.

b.

General Electric " Process Computer dated Performance June 1974 and Evaluation Ac NEDD-20340 and Admondment 1, NED0-20340-1  :

December 1974, respectively.

8 2-3 Amendeent No. 41 LASALLE - UNIT 2

, , em . w n+m a . x -l --

earn nar e a hmes eme1wf's g.o mem n m ecified in The Reactor Protection syntas instrissentation i are set for setpoints each sp Table 2.2.1-1 are the values at idnish + " the 4ne their reactorsafetytr ps'the Trip parameter. ii ted operational occur-core and reactor coolant system are prevented from -

idents. operation Limits during normal aparation and design basis ant c pa rences and to assist in mitigating the consequences i igierence of.accbut between, within its sp with a trip set less conservative than its Trip setpo ntf each Trip setpoint and the Allowable value is equaallow .

~ ~ '1" #4 *

• = == "4 *" - --

1. Ine = H 4 ="a tor trip Thu IItM system consists of 8 chambers, 4 in each of the reacThe trip

'the 23tu is a 5 decada 10 range instrument. Ttnas as the ZItM is j systems.

divisions of scale is active in each of the 10 ranges. rip setpoint is l ranged up to accommodate the increase in power level, the tTh also ranged up.

and sitM systems.

ii hanges during the power

'the most significant source of react v ty cIn order to ensure that the IIIM 1 accidents have increase is due to oestrol red wisia_r'.

gr ;

provides the required protection, a range of red ** w1 ' -rThe j/[oresn been analysed. conservatima FRAR. The most severe esse involves PoMER. an initial condician in

• M==taly 14 of RASED TMERMAL closest to the control Powsm is at .,,

was taken in this analysis by assuming the IRMThe results analysis of show that the rod being withdranst is bypassed. f RATED TMERMAL POWER reactor is shutdoun and peak power is limited lure threshold of 170 local

• with cal /gs.

the peak fuel enthalpy well -1 of below controlthe rodsin fuel' Based sequence and an this control rod errors and continuous PRM. wit

.providas backup protection for the A "4 '"

~

2. AEaraa= ===" = n=2 h APIIM scram For operation at low pressure and low flow during h al eargin between t e STANTUP,

.*f RASED TEIRMAL POIER provides dequate t eenh tes the anticipated setting of 15% The margin the setpoint and the Safety Limits. Effects of increasing pressure maneuvers associated with power plant startup. available at during sero or law stareg void is not nachcontant solder than arethat aimer already and sold la the water systas.4==d byfrom, Temperature soefficients are amall and controlBecause l +1e amass of signifiennt power increase.

red patte the stuu.

witimb r i is the most p .

b n==ad==nt No. 90 B 2-9

• A SAI.1,I - tP *

~

. l REACTIVITY CONTROL SYSTEM 3/4.1.2 REACTIVITY ANONALIES s

LIM u iNG COMITION POR OPERATION c r ihc./ r.

rd ess.J,,,v .'

3.1.2 The reactivity equivalence of the difference between the ac N and the p g g p }{not M N delta Uk. '

l OPERATIONAL CONDITIONS I and L

' APPLICABILITY:

1 ACTION:

! With the reactivity different by more than 5 delta k/k: l l l

a. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to detem

! 1s explained and corrected. ,

l t

b.

Otherwise, be in at least MT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I, I

j SURVEILLM kiiiiruiRJT5 WYfs~~

4.1.2 The reactivity equivalence of the difference between the actualM

' shall be verified to be less than or 1ENSTTY and the '*

(MJ*reh equal to 5 delta k/

l core AtrnATIoNs, and

a. Durine the fint startu, foiio ine b.

At least once per 31 effective full power days during POWER OPERATION. ,

k .

1 1

1 1

i J

s t ,,

3/4 1-2 LA SALLE - UNIT 2

i 1

1 i

! . 1 3/4.1 REACTIVITY CONTROL SYSTEMS l d __

j 8ASES j t

i I

! 3/4.1.1 SHUTDOW MARGIN i

j A sufficient SHUTDOWN MARGIN ensures that (1) the react subcritical from all operating conditions, (2) the reactivity transients j associated with postulated accident conditions are controllable within l acceptable limits, and -(3) the reactor will be maintaine l

Since core reactivity values will vary through core life as a function of 2

fuel depletion and poison burnup, the demonstration of SHUTDOWN MAR performed in the cold, xenon-free corditionl and ltd shall s l .The valuevalue of A in units of maximum core of X delta during reactivity K is thethe difference operatingbetweencycle and thet l beginning-of-life core reactivity.

must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration ofThethe SWTDOWN SHUTDOWN The highest worth rod any be determined analytically or by test.

MARGIN is demonstrated by an insequence control rod withdrawal at the 4

beginning-of-life fuel cycle conditions, and, if neces Observation of suber.iticality in this be reduced as a function of exposure. condition assures subcriticality with

! withdrawn.

l ' This reactivity characteristic has been a basic assumption in the l analysis of plant performance and can be best demonstrated at the tim j

loading, but the margin aust also be determined anytime g a control rod is incapable of insertion.

• • # f 3/4.1.2 REACTIVITY ANOMALIES IN requi at for the r is small, refuh '

Since SHUTDOWN ons is necessa and the l 1 condi to the icted co patterns check o from e comparisons o a

c in reactiv can be in checks are. an imposi the coup sons are a y done, f for no l Si ations. A change is 1 r than is l l nonnel is magni should be the ly evalua . A l

j o a change reactor operatio would not the design itions of cha large as tulated transi .

i an s on the sa side of the

]

8 3/4 1-1 l LA sALLE - UNIT 2

! l

• 1

• l

Lu;. '. . ; i., A..M in es,Ap-4 =

N J8f anert (---+t_.;) -

tto- The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety i analyses. Large differences between monitored and predicted

/Nr"h core reactivity may indicate that the assumptions of the DBA r<*/re- "I and transient analyses are no longer valid, or that the

/# C*' k uncertainties in the Nuclear Design Methodology are larger 4 than expected. A limit on the difference between the 7

cc4 co 7,/ =/

ag c,g Jp4 monitored core and the predicted core has been establis ed based on engineering ju gmant of 15 Ak/k

> 15 j ^^g ,/g, k J/r., deviation in reactivity from that predicted is larg r than 4

pg* or nomal operation and should therefore

]

I

.'""L I M* " ?" , MODE 1, 1

of th cont afw withdra a teady/ l 1

4 sta o ion is typ achieved. Under ese condi s, the compar between predi and monito co re ivity ro es effective i of the

, activi oma . In MOD , con rod re t ically '

! being withd during a sta a MODES 3 4, all l control a fully insert , , theraf , he reactor s in east active , moni ing ty is not ces . In $ 5 1 loadi re( ts in a contin y changing co activity.

i ts(LCD .1. ensure fue vesen are j of safety s, and perfo wi the ly EM ton is requ ng the fi startup followi rations that 1d have al reactivity l (s. . 1 vesent, trol rep 1 t, trol rod s ing). , LEO 3.1.1, vid a i rect comparis the predi monitored co

- ivity at itions; fore, reactivit ly not requ dur these iti s.

1. s kU _ _"N Should an y develop bet asured and radicted core react , the core ity difference st resto wt in the limi ure continued tion is vi n the co design sumptidas. Rastorat n the limit 1d perforund by an eval ion f the

• l design and saf analysis to dethmi reasons for

y. This untion normally v the core j

! t t4.o m

.",T:l - * !.1 ^ --

1, M /^7/*5 '

1./

h

POWERDISTRIB0TIONLIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO .

SURVERLANCE REQUIREMENTS O I j

d.2.' MCP , with:

y.p. I 0.86 prior to performance of the initial scram time measurement t ,y,= for the cycle in accordance with Specification 4.1.3.2, or a.

b. t,y, determined within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of eac surveillance test required by Specification 4.1.3.2.

shall be determined to be ecual to or greater than the applicable MCPR t lim specified in the CORE OPERATING LIMITS REPORT.

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.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <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 MCPR.

ASGrf S MJ -

Amendment No. 54 3/4 2-4 LA SALLE - UNIT 2

4 Insett#3 l].do3r h The applicable MCPR limit shall be determined from the COLR based on:

a.

Technical Specification Scram Speed (TSSS) MCPR limits, or b.

Nominal Scram Speed (NSS) MCPR limits if scram insertion times determined per surveillance 4.1.3.2 meet the NSS insertion times identified in the COLR.

Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of completion of each set of scram tasting, the results will be compared against the nominal scram speed (NSS) insertion times specifie ,

COLR, to verify the applicability of the transient analyses. Prior to initial scram time testing for an operating cycle, the MCPR operating limits used shall be -

based on the Technical Specification Scram Speeds (TSSS).

1 l

t

! I l '

I INSTRUMENTATION -

1 l

TRAVERSING IN-CORE PROBE SYSTEM _

i o

LIMITING CONDITION FDR OPERATION f The traversing in-core probe (TIP) system shall be: O a core in i

1 3.3.7.7.

a.

Movable detectors, drives and readout equipment to map l the required measurement locations and be calibrated

^

j '

b.

Indexing equipment to allow all required detectors-I in a common location. l l

i When the traversing in-core probe is used or:

APPLICABILITY:

l

• a. Recalibration of the LPRM detectors, and l

• b. Monitoring the APLHGR, LHGR, MCpR, or MFL j '

ACTION: noperable, required measure- l l With one or more TIP measurement locations d 2 below, provided the reactor

a. ments may be performed as described in I control rod lattern, and the

) core is operating in an octant symmetriant cycle has seen measured to be 1 total core TIP acertainty for the

! less than 8.7 percent.

asurement location may be. replaced by l

1. TIP data for an inoperable ng's endundant (symmetric) counterpar!

data obtained from that st obtained from an operable measurement >

i the substitute TIP data l location. ,

j able measurement location may be replaced by

TIP data for an inop I

1 2.

data obtained from ating measurements, provided t j with simulated available chann o s (measurement locations) does not e .

l j els of a single TIP machina, or All cha

' a) .

hine is l '

b) A to .1 of five channels if more than one TIP mac inv ved.

ith the TIP system inop9rable, suspend use of the

b. Otherwise applicable monitoring or calibration functions.

t the abov f

The p visions of specification 3.0.3 are not applicable.

c.

i SURVE J4NCE REQUIRinEnT5 PERABLE by i

4.

7a The traversing in-core probe system shall be demonst i liz t4on f . ciions.

or the above a,,u cabie monitor 4n, or can l

t o use ,ing each of the above required J i ed

• 0nly the detector (s) in the required measurement locatio l to be OPERABLE. Amendment No. 78 j 3/4 3-73

' a S ALLE ""~~ ?

f

3/4.3.7.7 AND 3/4.3.7.8 INTENTIONALLY LEFT BLANK PAGE 3/4 3-74 DELETED LASALLE -UNIT 2 3/4 3-73 NEXT PAGEIS 3/4 3-75

3 /4.4 REACTOR C00LANT SYSTEM 3/4.4.1 RECfRCULATION SYSTEM RECIRCULAT10N LOOPS LIMIfiNg wsDITION FOR OPERATTON ~

'3.4.1.1 Two reactor coolant system recirculation loops shall be in op APPLICABILTTY: OPERATIONAL CONDITIONS 1 and 2 EllE a.

With only one (1) reactor coolant system recirculation loop in operation, comply with Specification 3.4.1.5 and:

Within four (4) hours:

Place the recirculation flou control system in the Master a) Nanual ande or lower, and AVEM(E PMM/ ,

g/pggf // E47 CR m CAL POW R RATIO (MCPR) Safety b) Incmase the n! 1. per Specification 2.1.2, and pp/5/4)T/og g g /.T .' A p BATE Limit by 0.01

• b lacrease the MINDEM CRITICAL PERlER RATIO (MCPR) Lim c) Condition for Operation by 0.01 per Specification 3.2.3, to //'

.jf, ****

g jy V e Average Power Range Monitor (APRM) Scram and f .%c f ) Reduce and al and Rod slock Monitor Trip setpoints end "

L r, //s co/5 o/E/AT/ Allowable Values to those applicable to singl l p j f j 8 8 N gg7-3.3.s. .

s emise, he in at least HDT SHUTDOWN within the next twelve 2.

(12) hours. .

b.

With no reactor coolant recirculation loops in operation:

1.

Take the ACTION required by Specification 3.4.1.5, and

2. Se in at least NOT SHU1 DOM within the next six (6)

Amendment No. 88 3/4 4-1 LA SALLE - WIT 2

REACTIVITY CONTROL SYSTDtS BASES 3/4.1.3 CONTROL RODS (Continued) .

In addition, the automatic CRD charging water header low pressure scram (see Table 2.2.1-1) initiates Withwell before this added any accumulator automatic scram feature, loses its fu bility to insert the control rod.

the surveillance of each individual accumulator chec action.

Control rod coupling integrity is required to ensure compliance with analysis of the rod drop accident in the FSAR.

provides the only positive means of detensining th completing CORE ALTERATIONS that could have affected the control The subsequent check is performed as a backup to the coupling integrity.

initial demonstration.

therefore that other parameters are within their limits, position indication system must be OPERABLE.

The control rod housing support restricts the outward movement The of a control rod to less than 3.65 inches in the event of a housing failure.

amount of rod reactivity which could be added by this small meount of rod withdrawal is less than a normal withdrawal increment and will not con to any damage to the primary coolant system.there is no pressure t I housing. ,

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive waar on the s components.

3/4.1.4 CONTROL R00 Pit 0 GRAM C6nik6i.S l Control rod withdrawal and insertion sequences are established t to assure . l

. that the maximum insequence individual control rod or control rod segmen s which are withdrawn at any time during the fuel cycle could not be worth e to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a ihe specified sequences are characterized by control rod drop accident. When THERMAL POW'!R hoogeneous, scattered patterns of control rod witMrawal.

' is greater than 10% of RATED THERMAL POWER, there is no possible r which, if dropped at the design rate of the velocity limiter, could result in a Thus requiring the RWM to be OPERABLE when peak enthalpy of 280 cal /gs.

THERMAL POWER is less than or equal to 105 of RATED THERMAL POWER adequate control.

The RWM provide automatic supervision to assure that out-of-sequence r will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented t-

%,'. 2

.~.;;

hloclesr /}le;, NN:o:g:;;h?v fe{hy kde- Kgwefory . ggf,,7e g,ff ],

for Degy+~/A**W-Amendment No. 74 LA SA' ' - UNIT 2 B 3/4 1-4

SYSTDtS

_ REACTIVITY CONTROL _

BASES 3/4.1.4 CONTROL ROD PROGRAM CONTROLS (Continued)

The RBM is designed to automatically prevent fuel damage in the erroneous Two rod withdrawal channels are from locations of high power density du provided.

operation.

erroneous rod withdrawal soon enough to prevent fuel damage. . Th up the written sequence used by the operator for withdrawal of con 3/4.1.5 STANDBY LIOUID CONTROL SYSTEM The standby liquid control system provides a backup capability for i

bringing the reactor from full power to a cold, Xenon-free To shutdown, j that the withdrawn control rods remain fixed in the rated power pattern.

meet this objective it is necessary to inject a quantity of baron0 which to l

i produces aAconcentration of 660 ppa in the reactor core in approximat normal quantity of 4587 gallons not of solution having a 13.4%

125 minutes.

sodium pentaborate concentration is required to meet c shutdown re .

I 3%. There isThe antime additional requirementallowance was selected toofoverride 25% the in reactivity the reactor imperfect mixing.

insertion rate due to cooldown following the Xenon poison peak and the r The minimum storage volume of the solution is i

' pumping rate is 41.2 gps.

established to allow for the portion below the pump suction that cannot be

,a inserted and the filling of other piping systems connected to the reactor vessel.

l The temperature requirement on the sodium pentaborate solution is necessary to maintain the solubility ofvolume Checking the the solution asthe of fluid and it was initially mix i

to the appropriate concentration. temperature once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assu '

injection.

l I With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the rea periods of time with one of the redundant components inoperable.

Surveillance requirements are Onceestablished the solutiononisaestablished, frequency that boron assures a i

'high reliability of the system.

i concentration will not vary unless more boron or water is added, thus a check

on +he temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solutio i

l available for use.

l 1

will assure that these valves will not fail because of de

! charges.

~

1.

C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis

! for Large BWR's," G. E. Topical Report NED0-10527, March 1972 J

1

2. C. J. Paone, R. C. Stirn and R. M. Young, Supplement I to NEDO-10527, July 1972 ,

)

j i 3. J. M. Haun, C. J. Paone and R. C. Stirn Addendum 2. " Exposed Cores,"

Supplement 2 to NEDD-10527, January 1973

] Amendment No. 74 LA SAL' e - UNIT 2 B 3/4 1-5

l 3 /4.2 POKR DISTRIBUff0N LIMITS 1 I

! sAsEs i - . l The specifications of this section assure that the peak cladding l temperature following the postulated design basis loss-of-coolant accident i

will not exceed the 2200'F limit specified in 10 CFR 50.46.

3 /4.2.1 AVERAGE ptANAR_tfNEAR HEAT GENERATION RATE l

i GE f e/

TMs specification ass'ures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. This specification also assures that fuel rod mechanical integrity is maintained during nomal and transient operations.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHER for the highest powered rod which is equal to or less than the design LHER corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady-state gap conductance and red-to-rod local peaking factor.

The Technical Specification AVERAGE PLAMAR LINEAR HEAT GENERATION RAl'E (APLHGR) is this LHER of the highest _ powered rod divided by its local peaking factor. f. gg .py ,

However, the c., .t General El c (GE) calculational mode; s in Reference , which are consistent with 4he -

(SAFER /GESTR descri requirements of App ix K to 10 CFR 50, have established that APLHER values are not expected limited by LOCA/ECCS considerations. APLHER limits are still required, @ownsep, to assure that fuel rod mechanical integrity is maintained. They are specified for all fuel types in the.Coas CME

^;=t i - ' 4 =i + r " based on the fuel a anical design analysis.

OfSM7?Nr L////f~5Pi'MR T LHER factors specified in The purpose o* the power- and flow-de the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow and core power conditions. At less than 100% of rated flow or rate:i power, the required MAPLHER is the minimum of either (a) the product of the rated MAPLHGR limit and the power-dependent MAPLHER factor or (b)'the product of the rated MAPLHGR limit and the f1= f-ependent MAPLHER factor. The power- and flow-dependent MAPLHGR factors assure that the fuel remains within I

the fuel design basis during transients at off-rated conditions. Methodology forestablishingthesefactorsisdescribedinReference#l9 5* Gr6 l LA SALLE - UNIT 2 8 3/4 2-1 Amendment No. 88 l

l Insert #4 i

SPC Fuel i

l This specification assures that the peak cladding temperature of j SPC fuel following a postulated design basis loss-of-coolant accident will not exceed the peak cladding temperature (PCT) and l maximum oxidation limits specified in 10CFR50.46. The calculational procedure used to establish the AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) limits is based on a j loss-of-coolant accident analysis. The analysis is performed using e calculational models which are consistent with the requirements of APPENDlX K to 10CFR50. The models are described in Reference 1.

j The PCT following a postulated loss-of-coolant accident is primarily i a function of the average heat generadon rate of all the rods of a fuel l assembly at any axial location and is not strongly influenced by the

] rod-to rod power distribution wif.nin the assembly.

1 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE i

5 (APLHGR) limits for two-loop operation are specified in the CORE OPERATING LIMITS REPORT (COLR).

For single-loop operation, an APLHGR limit corresponding to the i product of the two-loop ilmit and a reduction factor specified in the COLR can be conservatively used to ensure that the PCT for single-loop operation is bound by the PCT for two-loop operation.

I i

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l POWER OfSTRIBUTf0N shitris

' BASES ,

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. 3/4.2'.'/ DELETED '

3 /4.2.3 MINIMUM CRITICAL POWQMHD l

The required operating liWit EPRs at steady-state operating conditions i

as specified in Specification 3.2.3 are derived from the established fuel i cladding integrity Safety Limit MCPR and an analysis,of abnomal operational l transients. For any abnomal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it

is required that the resulting MCPR does not decrease below the Safety Limic l

MCPR at any time during the transient assuming instrument trip setting given i in Specification 2.2.

To assure that the fuel cladding inteprity Safety Limit is not exceeded during any anticipated abnomal operationa , transient, the most limiting l

i transients have been analyzed to detemine which result in the largest reduc-i tion in CRITICAL POWER RATIO (CPR).

The type of transients evaluated'were i loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest I

delta MCPR. When added to the Safety Limit NCPR, the required eini!'en j operating limit MCPR of Specification 3.2.3. is obtained and presented % the j CORE OPERATING LIMITS REPORT.

l Analyses have been performed to detemine the effects on CRITICAL POWER

! TIO (CPR) during a transient assuming that certain equipment is out of l

service. A detai .ed description of the analyses is provided in Reference 5.

l The analyses performed assumed a single failure only and established the i

licensing bases to allow continuous plant operation with the analyzed l

equipment out of service. The following single equipment failures are i included are part of the transient analyses input assumptions:

l 1. main turbine bypass system out of service, ,

2. recirculation pump trip system out of service, hScrl $5 kam l -

1 i

{ I

! l 5 3/4 2-2 Amendment No. 88 LA SALLE - UNIT 2 l

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f i

The purpose of the power- and flow-dependent MCPR limits (MCPR,and MCPR respectively) specified in the CORE OPERATING LIMITS REPORT (COLR) is to

{

define operating limits dependent on core ficw and core power. At a given power l and flow operating condition, the required MCPR is the maximum of either the l power-dependent MCPR limit or the flow-dependent MCPR limit. The requiredl l

MCPR limit assures that the Safety Limit MCPR will not be violated. I l

' The flow dependent MCPR limits (MCPRo are established to protect the core from l inadvertent core flow increases. The core flow increase event used limits is a slow flow runout to maximum flow that does not result in a neutron flux overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1, 4

Item 2). A conservative flow control line is used to define several core power / flowi l state points at which the analyses are performed. MCP l i

i MCPRrlimits are established to protect the operating limit MCPR. For the manual  !

mode, the limits are set to protect against violation of the safety limit MCPR. /

The power-dependent MCPR limits, (MCPRe), are established to protect the core from plant transients other than core flow increases, including pressurization and the i, localized control rod withdrawal error events.

Analyses have been performed to determine the effects of assuming various equipment out-nf-service scenarios on the (CPR) during transient events. Scenario

! were performed to allow continuous plant operation with these systems out of l service. Appropriate MCPR limits and/or penalties are included in the COLR for f

each of the equipment out-of-service scenarios identilled in the COLR. In some j cases, the reported limits or penalties are based on a cycle-independent analysis, i while in other cases, analyses are performed on a c4 gsA basis.

1 References 2-6 describe the methodology and codes used to evaluate the potentially bounding non-LOCA transient events identified in Chapter 15 of the UFSAR.

MCPR limits are presented in the CORE OPERATING LIMITS REPORT (COLR) for

both Nominal Scram Speed (NSS) and Technical  ; A-+ " *-n Scram Speed (TSSS) j

- insertion bmes. The negative reactmty insertion rate resulting from the scram plays a major role in providing the required pve against violating the Safety Limit j MCPR during transient events. Faster scram insertion times provide greater protection and allow for improved MCPR performance. The application of NSS j

MCPR limits takes advantage of improved scram insertion rates, while the TSSS l j MCPR limits pavide the necessary pro'setion for the slowest allowable average l scram insertion times identified in Specification 3.1.3.3. If the scram insertion times determined per. surveillance 4.1.3.2 meet the NSS insertion times, the appropriate NSS MCPR limits identified in the COLR are apphed. If the scram insertion times do d

not meet the NSS insertion criteria, the TSSS MCPR limits are applied.

9 k

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4

1 POWER OfSTRIBUT20N SYSTEMS l .

, BASES -

3 j MINIMUM CRITICAL POWER RATIO (Continued) i"

3. safety / relief valve (S/RV) out of service, and ,

i

! 4. feedwater heater out of service (corresponding to a 100 degree F

reduction in feedwater temperature).

1 i i For the main turbine bypass and recirculation pump trip systems specific

[ cycle-independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for

! Operation (LCO) values are established to allow continuous plant operation with these systems out of service. A bounding end-of-cycle exposure condition

! was used to develop nuclear input to the transient analysis model. The

{ bounding exposure condition assumes a more top peaked axial power distribution j than the nominal power shape, thus yielding a bounding scram response with l reasonable conservatises for the MCPR LCD values in future cycles. The MCPR LCD values shown in the CORE OPERATING LIMITS REPORT for the main turbine bypass and recirculation pump trip systems out of service are valid provided l that these limits bound the cycle specific results.

i The analysis for main turbine bypass and recirculation pump trip systems t inoperanle allows operation with either system inoperable, but not both at the same time.

. For operation with the feedwater heater cut of service, a cycle specific

! analysis will be performed. With reduced feedwater temperature, the Load i Reject Without Bypass event will be less severs because of the reduced core d

steaming rate and lever initial void fraction. Consequently, no further analysis is needed for that event. However, the feedwater controller failure event becomes more severe with a feedwater heater out of service and could l becomi the limiting transient for a specific cycle. Consequently, the cycle i specific analysis for the feedwater controller failure event will be performed i with a 100 degree F feedwater temperature reduction. The calculated change in l CPR for that event will then be used in determining the cycle specific MCPR 1 LCO value.

1 In the case of a single S/RV Dut of service, transient analysis results

'showed that there is no impact on the calculated MCPR LCD value. The change l in CPR for tr.is operating condition will be bounded by reload licensing calculations and no further analyses are required. The analysis for a single

! S/RV out of service is valid in conjunction with dual and single recirculation l loop operation.

! The evaluation of a given transient begins with the system initial parameters

! shown in FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient j computer program. The codes used to evaluata events are described j

l -

l LA SALLE - UNIT 2 8 3/4 2-3 Ameneent No.54 b_ ____ _ _ _______ ___-- _ __ _ --

I POWER DISTRf8UTTON SYSTEMS BASES MINIMUM CRITICAL. POWER RATIO (Continued) in NEDE-24011-P-A-US (Reference 4). The outputs of these ' programs along with the initial MCPR form the input for further analyses of the thermally limiting bundle (Reference 4). The principal result of this evaluation is the reduction in MCPR caused by the transient.

~

The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the ODYN  !

computer code for analyzing rapid pressurization events. Generic statistical I analyses were performed for. plant groupings of similar design which considered the statistical variation in several parameters, i.e., initial power level,  !

CRD scram insertion time, and model uncertainty. These analyses, which are described further in Reference 2, produced generic Statistical Adjustment Factors wnich have been applied to plant and cycle specific ODYN results to yield operating limits which provide a 955 probability with 95% confidence that the limiting pressurization event will not cause MCPR to fall below the fuel cladding integrity Safety Limit. '

i I

As a result of this 95/95 approach, the average 205 insertion scram time

{ must be monitored to assure compliance with the assumed statip'ical distribu-i tion. If the mean value on a cycle cumulative, running aversge, basis were to

! exceed a 5% significance level compared to the distribution assumed in the ODYN statistical analyses, the MCPR limit must be increased linearly, as a function of the mean 205 scram time, to a more conservative value which reflects i

an NRC determined uncertainty penalty of 4.45. This penalty is applied to the

{ plant specific ODYN results, i.e. without statistical adjustment, for the limit-ing sin cycle. gle failure pressurization event occurring at the limiting point in the It is not applied in full until the mean of all current cycle 20% scram times reaches the 0.86 seconds value of Specification 3.1.3.3. In practice, I however, tne requirements of 3.1.3.3'would most likely be reached, i.e. , indivi-cual cata set average > 0.86 secs, and the required actions taken well before j the running average exceeds 0.86 secs.

I

{'

• The 5% sigt.ificance level is defined in Reference 4 as:

! 1

-j f 1,=,.1...(yi=1hp2 ,

p where , a mean value for statistical scram time distribution to 205 insertad = .572

. =

Ng a standard deviation of above distribution = .016 number of rods tasted at 80C, i.e., all operable

. rods n

1Ng= total number of operable rods tested in the i=1 current cycle LA SALLE - UNIT 2 . t 3/4 2-4 Amendment No.54 l

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! p0WER DISTRIBUTION instris i' BASES j -

i MINIMUM CRITfCAL pMe RATIO (Continued)

The value for f used in Specification 3.2.3 is 0.68f seconds which is j

conservative for the,following reason:

! For simplicity is fomulating and implementing the LCO, a conservative J n

! value for i=1, IN of Eg8 was used. This represents one full core data set 4

at 80C plus one full core data set following a 120 day outage plus twelve a 10% of core,19 reds, data sets. The 12 data sets are equivalent to

! 24 operating months of surveillance at the increased surveillance I frequency of one set per 60 days required by the action statements of i Specifications 3.1.3.2 and 3.1.3.4. -

{ That is, a cycle length was assumed which is longer than any past or

contemplated refueling interval and the number of rods tested was maximized in

! order to simplify and conservatively reduce the criteria for the scram time at which MCPR penalization is necessary.. .

f l

The purpose of the power- and flow-dependent EPR limits specified in the

CORE OPERATING LIMIT 5 REPORT is to define operating limits at other than rated j core flow and core poner conditions. At a given power and flow operating 4 condition, the required MCPR is the maximum of either the power-dependent EP l limit or the flow-dependent MCPR limit. The required EPR assures that the

' Safety Limit EPR will not be violated.' Methodology for establishing the power- and flow-dependent EPR limits is described in Reference 6.

L At THERMAL POWER levels less than or equal to 255 of RATED THERMAL POWER,

the reactor will be aparating at minimum recirculation pump speed and the i

moderator void content will be very small. For all designated control rod

! patterns which may be employed at this point, operating plant experience

! indicates that the resulting MCPR value is in excess of requirements by a

• considerable margin. During initial start-up testing of the plant, a EPR evaluation will be made at 255 of RATED THERMAL POWER level with minimus j recirculation pump speed. The E PR margin will thus be demonstrated such that j n'ture MCPR evaluation below this power level wiU be shown to be sonocessary.

. The daily requirement for calculating EPR when THERMAL POWER is greater than

or equal to 25% of RATED THERMAL POWER is sufficient since power distribution

! shifts are very slow een there have not been significant power 'or control rod 3

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

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! LA SALLE - UNIT 2 B 3/4 2-5 Amendment No. 88 I

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l POWER DISTRIBUTIDN SYSTEMS BASES

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3 /4. 2. 4 LINEAR HEAT RENERATION RATE

&pFw/-

The specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in any rod is less than the design linear heat annee=+4a= = van if fuel pellet _

densification is costula+ C Yhe r pen y spedified Das on

/the a ysis sent n i .2.1 the top al rep 073 (5 two coreene , s a 11 arly creas var tion axia aps i

on top as a9 confi ce t no re t one coeds e de gn LI GEN ON du r

- piki .

References 5

}h N $&rG]

General Electric Company Analytical Nodel for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NED0-20566A',

September 1986.

2. " Qualification of the One-Dimensional Core Transient Nodel for Boiling hter Reactors," General Electric Co. Licensin9 Topical Report NEDO 24154 Vols. I and II and NEDE-24154 Vol. III as sup-plemented by letter dated September 5,1980, from R. H. Buchholz (GE) to P. 5. Check (NRC). , ,

3. "LaSalle County station Units I and 2 SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis," General Electric Co. Report NEDC-32258P, October 1993. g

4. " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A (latest approved revision).

5. " Extended Operating Domain.and Equipment Out-of-service for LaSalla County Nuclear Station Units 1 and 2,* NEDC-31455, November 1987.

6.

• ARTS Improvement Profram Analysis for LaSalle County Station I Units 1 and 2,* General Electric Co. Report NEDC-31531P, December 1993. ,

h*SYY$7henr]

g [g,g ,f .fw/ dec.)G~rek drir tNK*stW h }

f g ,4,.af g/ecbr7C 'S k/?! ** *' t y/

g [gg, ,,g, TAR),]y 7,SED entwa 5-9900S.

L#6A re~ea We hebSEMA He I'*"&

eg Tys, limib-LA SALLE - UNIT 2 8 3/4 2-6 Amendment No. 88

-t

insert #6 SPC Fuel 4 The Linear Heat Generation Rste (LHGR) is a measure of the heat generation rate per unit length of a fuel rod in a fuel assembly at any axial location. LHGR limits are specified to ensure that fuel integrity limits are not exceeded during normal operation or anticipated operational occurrences (AOOs). Operation above the LHGR limit l followed by the occurrence of an ADO could potentially result in fuel damage and subsequent release of radioactive material. Sustained operation in excess of the LHGR limit could also result in exceeding the fuel design limits. The failure mechanism prevented by the LHGR limit that could cause fuel damage during AOOs is rupture of the fuel rod cladding caused by strain from the expansion of the fuel pellet. One percent plastic strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur. Fuel l design evaluations are performed to demonstrate that the mechanical design limits are not exceeded during continuous i

operation with LHGRs up to the limit defined in the CORE OPERATING LIMITS . REPORT. The analysis also includes allowances for short term transient operation above the LHGR limit.

At reduced power and flow conditions, the LHGR limit may need to tie reduced to ensure adherence to the fuel mechanical design bases during limiting transients. At reduced power and flow conditions, the LHGR limit is reduced (multiplied) using the smaller of either the flow-dependent LHGR factor (LHGRFAC,) or the power-dependent LHGR factor (LHGRFAC,) corresponding to the existing core flow I and power. The LHGRFAC, multipliers are used to protect the core during slow flow runout transients. The LHGRFAC, multipliers are used to protect the core during plant transients other than core flow l

transients. The applicable LHGRFAC, and LHGRFAC, multipliers are l specified in the CORE OPERATING LIMITS REPORT.

1 l

i f . Insert #7 i l 1

i 1. Advanced Nuclear Fuels Corporrtion Methodology for Boiling Water Reactors  :

EXEM BWR ECCS Evaluation, ANF-91-048(P)(A), Advanced Nuclear Fueis j Corporation, January 1993, i

2. Exxon Nuclear Methodology & Bomng Water Reactors, Neutronic Methods for i Design and Analysis, XN-NF-80-19 (P)(A), Volume 1 (as supplemented), Exxon

Nuclear Company, March 1983.

3. Exxon Nuclear Methodology for Bogng Water Reactors, THERMEX Thermal i Limits Methodology Summary Description, XN-NF-80-19 (P)(A), Volume 3 '

j Revision 2 (as supplemented), Exxon Nuclear Company, January 1987.

4. Exxon Nuclear Plant Transient Methodolo py for Bolling Water Reactors, XN-NF-

{ 79-71 Revision 2 (P)(A) (as supplemented), Exxon Nuclear Company, March j 1986.

! 5. COTRANSA2: A ComputerPre v , i,hBoiling Water ReactorTransient

! Analyses, ANF-913(P)(A) Volume 1 Revision 1 (as supplemented), Advanced

! Nuclear Fuels Corporation, August 1990.

i* 8. XCOBRA-T: A Computer Code h BWR Transient Theiii._' Nrau8c Cors Analysis, XN-NF-84-105(P)(A) Volume 1 (as supplemented), ifxxon Nuclear Company, February 1987. . .

l

7. Generic Mechanical Design Criteris for BWR Fuel Designs, ANF-89-98(P)(A)

' Revision 1 (as supplemented), Siemens Power Corpora 6on - Nuclear Division, May 1995.

l

8. LaSalle County Station Units 1 and 2 SAFER /GESTR - LOCA Loss <d-Coolant Accident Analysis, NEDC-32258P, General Electric Company, October 1993.

' 9. ARTS Improvement Program analysis for LaSalle County Station Units 1 and 2, NEDC-31531P, General Electric Company, December 1993.

4 9

,,...m..+... .n mAsrs 3 /4. 3. 3 . EMERGENCY CDRE axrLING ff5 TEM ACTDATTON TWITNMENTATION (continued) specified surveillance intervals and surv=111==a= sad maintenance outage times have been determined in accordamos with EEDC-3093&P-A, "Taahaie=1

' specification T=.= - . - ; Esthodology (With Demonstration for ENR ECCS Actuation Instrumentation)", Parts 1 and 2, December 1988, and RE-025 Revision 1,

• Technical specification Improvement Analysis for' the Emergency Core Cooling system Actuation Instrumentation for Lasalle county station, Units 1 and 2*, April 1991. When a channel is placed in am inoperable status solely ior perfarmance of required surveillances, entry lato Zap and required ACTIONS

, may be delayed, provided the associated function ==4=*=4== BCCs initiation

capability.

3 /4. 3. 4 RhinuudTTON MDip vmfP ACTUATfDN --inm.m_sion The anticipated transiest without scram (A2EE) recirculation pump trip system provides a means of limiting the consequenons of the un11 holy occurrence of a failure to scram during an ma*ia4==ted transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NEDD-10349, dated March 1971 and NEDD-24222, dated December,1979, and Appendia G of the F8&R.

The end-of-cycle recirculation ymmp trip (EDC-RFT) sy' stem is a part of

, the Beactor Protection system and is an essential safety supplement to the reactor trip. The purpose of the EOC-EFT is to remover the loss of *h===1 margin which occurs at the end-of-cycle. The physiasi phaaa==aa= involved is that the void reactivity feedback due to a pressert==*ina transiest can add positive reactivity to the reassor system at a faster rate than the control rods add negative scram reactivity. Each EDC-EFT system trips both recircula-tion pumps, reducing coolant flow in order to reduce the void onllapse in the core during two of the most limiting pressurimmei== events. The two events for which the EDC-RPT , mive feature will functies are closure of the turbine stop valves and fast closure of the turbine sontrol valves.

A ~ anal , whi ides -

1. operation one tri et the has been armed.

analy det r- '

- = sycle * '. ' -

- - = 0 I

' (M ) Limi for ation ( values anst if

-RFT tem is la These ensure t margin reactiv the saf t in'the of ysed -

wit he ion able. anal resol dis the e for S .2.3.

A fast closure senser from weh of two turbine control. valves provide sad input to the IOC-RPT systant a fast closure me.aeor from each of the other turbine control valves provides input to the acomed EDC-RFT system.

N similarly, a position switch for each of tuo turbine stop valves provides input to one BOC-RPT systemt a position switch from each of the other two stop valves provides impet to the other EOC-EPT system. Per each EDC-EFT system, the eensor relay contacts are arranged to foam a 2-est-of-2 logia for the fast closure of turbine sentrol valves and a 2-est-of-2 logia for the turbine stop valves. The operation of either logic will actuate the EOC-RFT system and trip both recirculation pumps.

Each EOC-AFT system may be manually bypassed by ese of a keyewitch which is ad=ini=tratively controlled. The manual hypseems and the autamatic Operating typass at less than 30% of R&IED 2EEEE&L PONER are anamaciated in the control room.

LA SALLE - UNIT 2 5 3/4 3-3 a===d===t no. 90

_ - ~ - - - _ . - - - _ . -_ __ .

l l

Insert #8 Analyses were performed to support continued operation with one or both i trip systems of the EOC-RPT inoperable. The analyses provide MINIMUM l l

CRITICAL POWER RATIO (MCPR) values which must be used if the EOC-RPT system is inoperable. These MCPR Ilmits are included in the COLR and ensure that adequate margin to the MCPR safety limit exists with the EOC-RFT function inoperable. Application of these limits are discussed further in the bases for  ;

Specification 3.2.3.

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i l INSTRUMENTATION i

f BASES

{ MONITORINGINSTRUMENTATION(Continued) j 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION i The OPERABILITY of the accident monitoring instrumentation ensures that i sufficient information is available on selected plant i and assess important variables following an accident. parameters to monitorThis cap sistent with the recommendations of Regulatory Guide 1.97 " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During j and Learned Following Task ForceanStatus Accident, Report and Short-Tern Recommendations"." Decem j 3/4.3.7.6 SOURCE RANGE M)NITDRS I The source range monitors provide the operator with information of the j

status of the neutron level in the core at very low power levels during startup

! and shutdown. At these power levels reactivity additions should not be made

! without this flux level infomation a,vailable to the operator. When the inter-j i mediate range monitors are on scale adequate informatien is available without ip the SRMs and they can he setracted. -

!f 3/W.3.E7 MERSING IN-CORE PR08E SYSTEM i F The'0PERABILITY of the ' traversing in-cote probe (TIP) system with the j specified minimum complement of equipment ensures that the measurements obtained j from use of this equipment accurately represent the spatial neutron flux dis-

tribution of the reactor core.

! The specification allows use of substituted TIP data from s channelsifthecontrolrodpatternissymmetricsincetheTIPdSais tric l

j adjustedbythe i dependent bias. The plantsource computer to remove of data for the nachine dependent substitution and any also bepower a level

! 3-dimensional BWR core simulator calculated data set which is normalized to

! available real data. Since uncertainty could be introduced by the simulation f and normalization process, an evaluation of the specific control rod pattern

} } and core operating state must be perfomed to ensum that adequate margin to i gre operating limits is maintained.

3/4.3.7.8 DELETED 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION-OPERABILITY of the fire detection instrumentation , ensures that adequate warnino pability is available for the prompt detection of fires. This

capabili is recuired in order to detect and locate fires in their early j stages. rompt cetection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility i fire protection program.

d In the event that a portion of the fire detection instrumentation is j inoperable increasing the frequency of fire watch patrols in the affected areas is re, quired to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

l 3/4.3.7.10 DELETED 1

LA SALLE - UNIT 2 B 3/4 3-5 Amendment No. 69 1

l

Core Oneratina Limits Renert (Continued)

(1) The Average Planar Linear Heat Generation Rate (APLHGR) for Eh6 4 a,,,,/ 2ed' Technical Specification 3.2.1.g, q (2) The minimum Critical 'i ;idb =#

E g '" ower Ratic ndent MCPR (MCPR)its, lim and op;;wer and x . s tehndent

, + -'_' M Mt ts) for Technical Specif cation f,f.g;ce

[ " Ued - (3) TheSpecification Linear Heat Generation Rate (LHER) for Technical -

3.2.4.

(4) The Rod Block' Monitor Upscale Instrumentation Satpoints for Technical Specification Table 3.3.5-2.

b. The analytical methods used to determine.the core operating limits shall be those previously reviewed and approved by the 7 NRC in the latest bS4d l reports describin roved a revision orFor methodolo su[plement of theStation aSalle County topical g] 2, the topic 1 reports are:gy.

.{47 NEDE-24011-P-A, (af Reactor Fuel," ( latest approved revision).' General Electric Stan J27 Commonwealth Edison Topical Report NFSR-0085, " Benchmark of (g) BWR Nuclear Design Methods," (latest approved revision).

.J37 Commonwealth Edison Topical Report NFSR-0085," Supplement 1, fy) " Benchmark of 8WR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons," (latest approved revision).

Jer Commonwealth Edison Topical Report NFSR-0085, Supplement 2

" Benchmark of BWR Nuclear Design Methods - Neutronic (pf) Licensing Analyses " (latest approved revision (. '

c. Tie core operating limits shall be determined so that all applicable limits La.c., fuel thermal-mechanical limits, core themal-hydraulic 'ial.ts ECCS Limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any mid-cycle -

revisions or supplements thereto, shall be provided upon issuance, for each reload cycle to the U.S. Nuclear tegulatory '

CommissionDocumentControlDeskwithcopiestotheRegional Administrater and Resident Inspector.

B. Deleted. _

(3) Co- om/M Ea4so Toh/ Pepory' NFSt-M

1. Bedwk a f C4tho/hk/0Bof# BWe #de*-

psy fre Nolr, ^(/a fer7' affrove/ redsfo 0.

LA SALLE - UNIT 2 6-25 Amendment No. 88

l i insert #9 l '

! 1. ANFB Critical Power Correlation, ANF/ EMF-1125(P)(A), (as supplementet!).

I l 2. Letter, Ashok C. Thadani(NRC) to R. A. Copeland (SPC)," Acceptance for Referencing of ULTRAFLOWS Spacer on 9x9-IX/X BWR Fuel Design,

! July 28,1993.

! 3. Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling l Water Reactors, ANF-524(P)(A) Revision 2 (as supplemented).

i 4. COTRANSA 2: A Computer Program for Boiling Water Reactor Transient j Analysis, ANF-913(P)(A), Volume 1, Revision 1 (as supplemented).

l l 5. HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K

Heatup Option, ANF-CC-33(P)(A)(as supplemented).

6. Advanced Nuclear Fuel Methodology for Boiling Water Reactors, i XN-NF-80-19(P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and i Supplement 4.

I

7. Exxon Nuclear Methodology Boiling Water Reactors: Application of the ENC 1

Methodology to BWR Reloads, XN-NF40-19(P)(A), Volume 4, Revision 1, June 1986.

8. Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal

, Limits Methodology Summary Description, XN-NF4019(P)(A), Volume 3 j Revision 2, January 1987.

1

' 9. Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-8567(P)(A).

10. Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced

} Nuclear Fuels Corporation 9x9-lX and 9x9 9X BWR Reload Fuel, j ANF-89-014(P)(A), Revision 1 (as supplemented).

i  !

11. Volume 1 - STAIF - A Computer Program for BWR Stability in the Frequency '

i .

Domain, Volume 2 - STAIF - A Computer Program for BWR Stability in the l Frequency Domain, Code Qualification Report, EMF-CC-074(P)(A).

12. RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, j XN-NF-81-58(P)(A), Revision 2 (as supplemented).

4 i

13.XCOBRA-T: A computer Code for BWR Transient Thermal-Hydraulic Core q Analysis, XN-NF-84-105(P)(A), Volume 1 (as supplemented).

a

14. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors

EXEM BWR Evaluation Model, ANF-91-048(P)(A).

i .

j l

l 15. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for J  !

' Design and Analysis, XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, l Exxon Nuclear Company, Richland, WA 99352, March 1983. l I

insert #9 (continued)

16. Exxon Nuclear Plant Transient Methodology for Bolling Water Reactors, XN-NF-79-71(P).

17. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A).

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DESIGN FEATURES i

5.3 REACTOR CORE 87 dre 6TN i FUEL ASSEMBLIES j .,T6 yer / #/0 l b M eactor core shall contain 764 fuel assemblies. Each assembly i iconsists of a matrix of Zircalloy clad fuel rods with an initial composition i I of slightly enriched uranium dioxide, UDs. Fuel assemblies shall be limite j p those fuel designs approved for use in SWR's.

I

! CONTROL ROD ASSEMBLIES i

! ' 5.3.2 The reactor core shall contain 185 cruicform shaped control rod assemblies. The control material shall be baron carbide powder (8 C) and/or hafnium metal. The control rod assembly shall have a nominal axial4 absorber length of 143 inches.

l l 5.4 REACTOR COOLANT SYSTEM i DESIGN PRESSURE AND TEMPERATURE i

i 5.4.1 The reactor coolant system is designed and shall be maintained:

q '

t a. In accordance with the code requirements specified in 5ection 5.2 l* of the FSAR, with allowance for normal degradation pursuant to the j applicable Surveillance Requirements,

b. For a pressure of

• l 1. 1250 psig on the suction side of the recirculation pumps.

l- 2. 1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

I

3. 1500 psig from the discharge shutoff valve to the jet pumps.

l c. For a temperature of 575'F.

VOLUME 5.4.2 Tha totai water and steam volume of the reactor vessel and reci,rculation system is

• 21,000 cubic feet at a nominal T,,, of 533'F.

)

5. 5 METEOROLOGICAL TOWER LOCATION i 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

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• q i .

i k

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LA SALLE - UNIT 2 fr-4 Amendment No.54 a

} .

Design Fostures j . '

• i 4.0 4.0 DESI TURES f 5Aserl W/0

. si ation [T cation of s tion]

4 a s-

,n .

p. ., y 4.2.1 uel Ass ==h11 g

{ 74V i

The reactor shall contain fuel assemblies. Each assembly l shall consist of a matrix of ircIlloy r Tir 7 1 rods with j an initial composition of natural or slightly enr;iched uranium

dioxide (%) as fuel materia'd, -f _t:r 2]. Limited 4

substitutions of zirconium al' or stainless steel filler rods -

for fuel rods, in accordance wt approved applications of fuel

! rod configurations, may be used. Fuel assemblies shall be limited l

l to those fuel designs that have l analyzed with applicable 18tc '

i staff approved codes and methods shown by tests or analyses to

! comply with all safety design bases. A limited number of lead l l

test assemblies that have not comp 1 representative testing may-be placed in nonlimiting core regions.

The h . /4.1 A S.

4.2.2 M Red An===hlien *#

The core shk1 j ass ins.

a 193 . crucifore eh=w Wrol rod 11 be carbide, hafni i

i /

u- .roltem.aterial

c. .

7 j .

4

4. Fuel Stora 1 ,.

4.3.1 iticality '

] '/ 4.3.1.1 The fuel storage are desi and shall be maintained with:

a. Fme assemblies ving a maximus in

! .' /the normal rua core confim tim. at

-infinit cof(offl.31 Aonditi s]

.' / [ average U-l enrichment of %.5] weight ];

i

/ b. 4 inci 5 0. if fully fl with unbora unter, which an allomance uncertainties'as described in

./ [ Sect 4on 9.1 of the ];

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/ /

\ /

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, ./ g inued) i SWR /6 ST5 4.0-1 1,04/07/95 i /

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Attachment C l

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i Evaluation of Significant Hazards Evaluation l l

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C. Evaluation of Significant Hazards Considerations

References:

1. ANF-89-014(P)(A) Rev. I Supplement 1, Generic Mechanical Design for Advanced Nuclear Fuels 9X9 IX and 9X9 9X BWR Reload Fuel.

2. EMF 94 217(P) Rev. I, Siemens Boiling Water Reactor Licensing Methodology Summary.

The fuel supplier for LaSalle is being changed from General Electric (GE) to Siemens Power Corporation (SPC). As a result, certain items in the Technical Specifications are being revised. These changes can be classified in three categories: (a) fuel thermal limits, (b) miscellaneous, and (c) minor changes not related to the SPC transition. Each is discussed below,

s. Fuel therusal liasits l

The fuel thermal limits in the Technical Specifications are LHOR, APLHOR, and MCPR. Each fuel vendor provides LHGR and APLHGR limits for their fuel. As required by the Technical I Specification Surveillance Requirements (SR's), each fuel type will continue to be monitored via its vendor supplied LHOR and APLHOR limits. As such, the change to the Technical Specifications Bases for LHOR and APLHOR will be the addition of background information related to the SPC LHGR and APLHOR. The Limiting Conditions for Operation (LCO), Action Statements and SR's are unaffected since they refer to the Core Operating Limits Report for the fuel type dependent limits.

The CPR is calculated using a Nuclear Regulatory Commission (NRC) approved CPR correlation.

The GE correlation (GEXL) is being replaced by the SPC correlation, ANFB The co-resident OE fuel will be monitored by the ANFB correlation supplemented with bundle geometry dependent facto : to ensure the calculated CPR data is conservative with respect to that which would have been calculated by the OEXL correlation. This mixed core treatment of CPR is being documented in EMF ll25(P) Supplement I Appendix C,"ANFB Critical Power Correlation Application for Co.

Resident Fuel," November 1995. In light of the requested schedule for the approval of these Technical Specification changes, a LaSalle Unit 2 Cycle 8 specific document, EMF-%-021,

" Application of the ANFB Critical Power Correlation to Co-resident GE Fuel for LaSalle Unit 2 Cycle 8, has been submitted to the NRC for interim approval. The Technical Specifications and Bases related to the GE methods for determining the operating limit for MCPR are replaced by the SPC methods.

b. Miscellaneous change The Reactivity Anomaly surveillance is being upgraded to be consistent with SPC methods and NUREG 1434.

c. Minor Changes Not related to SPC Transitlem The Traversing In-core Probe (TIP) uncertainty limits (Specification 3.3.7.7) is being re-located from the Technical Specifications as a line item improvement from the Improved Technical Specifications (NUREG 1434). The same is true for the fuel description in Specification 5. A typographical error is being corrected in the Bases (page B 2 9) related to the power level at which the IRM system terminates the low power control rod withdrawal error event.

Comed has evaluated the proposed Technical Specification amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazard

coraideration established in 10CFR50.92 (c), operation of LaSalle Units 1 and 2 in accordance with the proposed amendment (s) will not represent a significant hazards consideration for the following reasons:

These changes do not:

1. Involve a significant increase in toe probability or consequences of an accident previously evaluated, j The probability of an evaluated accident is derived from the probabilities of the individual precursors to that acciW. Tue consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits will be established consistent with NRC approved methods to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed Technical Specifications amendment reflects previously approved SPC methodology used to analyze normal operations, including anticipated operational occurrences (AOOs), and to determine the potential consequences of accidents.

Licensino Methods and Models The proposed amendment is to support operation with NRC approved fuel and licensing methods supplied from Siemens Power Corporation. In accordance with FSAR Chapter 15, the same accidents and transients will be analyzed with the new fuel and methods as were analyzed by OE for OE fuel. The analysis methods and models are NRC approved (Note the mixed core treatment of CPR is being addressed under separate correspondence). These approved methods and models are used to determine the fuel thermal limits. Traversing In core Probe (TIP) uncertainty are assumptions in the approved Siemens core monitoring methodologies. The SPC core monitoring code enables the site to monitor k,,s weil as rod density to perform the reactivity anomaly surveillance. This is consistent with GE methodology. Therefore, the change in licensing analysis methods and models does not significantly increase the probability of an accident or the consegrences of an accident previously identified. The support systems for minimizing the consequences of transients and accidents are not affected by the proposed amendment.

New Fuel Denien The use of ATRIUM 9B fuel at LaSalle does not involve a significant increase in the probability or consequences of any accident previously evaluated in the FSAR. The ATRIUM-9B fuel is generically approved for use as a reload BWR fuel type (See Reference 1). Limiting postulated occurrences and normal operation have been enalyzed using NRC-approved methods for the ATRIUM 98 fuel design to ensure that safety limits are protected and that acceptable transient and accident performance is maintained.

The reload fuel has no adverse impact on the performance of in core neutron flux instrumentation or CRD response. The ATRIUM 9B fuel design will not adversely affect performance of neutron instrumentation nor will it adversely affect the movement of control blades. The exterior dimensions of the ATRIUM 9B fuel assembly are essentially identical to the OE9B; the ATRIUM-9B fuel assembly for LaSalle uses a standard fuel channel and normal control cell positioning (i.e., no offset). Thus, no adverse interactions with the adjacent control blade and nuclear instrumentation are anticipated. Additionally, given the above mentioned overall envelope similarities, no problems are anticipated with other station equipment such as the fuel storage racks, the new fuel inspection stand and the spent fuel pool fuel preparation machine.

The ATRIUM 9B design is neutronically compatible with the existing fuel types and core components in the LaSalle core. SPC tests have demonstrated that the ATRIUM 9B fuel design is hydraulically compatible with the OE9 fuel. The bundle pressure drop characteristics of the

l ATRIUM 9B bundle are similar to those of the OE9 fuel design, hence core thermal-hydraulic stability characteristics are not adversely affected by the ATRIUM 9B design.

1 An evaluation of'he Emergency Procedures is being performed to ensure that the use of the ATRIUM 9B fuel at LaSalle does not aher any assumptions previously made in evaluating the radiological consequences of an accident at LaSalle Station.

I j Methods approved by the NRC are being used in the evaluation of fuel performance during normal and abnormal operating conditions. The Comed and SPC methods to be used for the cycle specific j transient analyses have been previously NRC approved. The exception is the mixed core treatment of CPR, which is being addressed under separate correspondence.

The description of the fuel is expanded to be consistent with NUREO 1434. The description of the fuel materials, lead test assembly use, and stating that designs must have been analyzed with NRC Staff approved codes does not change existing methods; it only describes them.

l Review of the above concludes that the probability of occurrence and the consequences of an accident previously evaluated in the safety analysis report have not been significantly increased.

Comed has evaluated the proposed License amendment and determined that it does not represent a j

significant hazards consideration. Based on the criteria for defining a significant hazard consideration l

' established in 10CFR50.92 (c), operation of LaSalle Units 1 and 2 in accordance with the proposed amendment (s) will not represent a significant hazards consideration for the following reasons:

hoe changes do not:

i

2. Crwate the possibility of a new or different kind of accident frein any accident previously evaluated:

Creation of the possibility of a new or different kind of accident would require the creation of one i or more new precursors of that accident. New accident precursors may be creat:d by modifications of the plant configuration, including changes in allowable modes of operation.

Licensina Methods and Models N proposed Technical Specification amendment reflects previously approved SPC methodology used to analyze normal operations, including AOOs, and to determine the potential consequences of

accidents. As stated above, the proposed changes do not permit modes of reactor operation which differ from those currently permitted.

New Fuel Desian N basic design concept of a 9x9 fuel pin array with an internal water box has been used in various lead assembly programs and in reload quantities in Europe since 1986. WNP 2 has loaded reload l quantities since 1991. Approximately 650 water box assemblies have been irradiated in the United States through 1995, with a substantially higher number being inaciated overseas. The NRC has reviewed and approved the ATRIUM 9B fuel design (See Reference 1) W similarities in fuel design and operation indicate there would be no expectation of introducing new or different types of '

accidents than have been considered for the existing fuel. Therefore, the use of ATRIUM-9B fuel at LaSalle does not create the possibility of a new or different kind of accident from any accident previously evaluated.

I, I

Comed has evaluated the proposed License amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazard consideration established in 10CFR50.92 (c), operation of LaSalle Units I and 2 in accordance with the proposed amendment (s) will not represent a significant hazards consideration for the following reasons:

Wee changes do not:

l 3. Involve a significant reduction la the margin of safety for the following j rwasons:

b existing margin to safety is provided by the existing acceptance criteria (e.g.,10CFR50.46 I limits). W proposed Technical Specification amendment reflects previously approved SPC methodology used to demonstrate that the existing acceptance criteria are satisfied. The revised methodology has been previously reviewed and approved by the USNRC for application to reload cores of GE BWRs. References for the Licensing Topical Reports which document this methodology, and include the Safety Evaluation Reports prepared by the USNRC, are added to the Reference section of the Technical Specifications as part of this amendment.

Licaa=la- Me*~la and Models b proposed amendment does not involve changes to the existing operability criteria. NRC approved methods and established limits (implemented in the COLR) ensure acceptable margin is maintained. The Comed and SPC reload methodologies for the ATRIUM 9B reload design are

, consistent with the Technical Specification Bases. & Limiting Conditions for Operation are taken

) into consideration while performing the cycle specific and generic reload safety analyses. NRC j

approved methods are listed in Specification 6 of the Technical Specifications.

l 1

Analyses performed with NRC-approved methodology have demonstrated that fuel design and licensing criteria will be met during normal and abnormal operating conditions. Therefore, there is not a significant reduction in the margin of safety.

New Fuel Denien i

The exterior dimensions cf the ATRIUM 9B fuel assembly are essentially identical to the OE9B; the ATRIUM-9B fuel assembly for LaSalle uses a standard fuel channel and normal control cell  ;

positioning; i.e. no offset. Thus, no adverse interactions with the adjacent control blade and nuclear '

instrumentation are anticipated. & change does not adversely impact equipment important to safety and, therefore does not reduce the margin of aarety .

Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve significant hazards considerations.

This proposed amendment most closely fits the example of a change which may either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan.

4 This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant j

relaxation of the bases for the limiting conditions for operations. Therefore, based on the guidance i

provided in the Federal Register and de criteria established in 10 CFR 50.92(c), the proposed enange ,

does not constitute a significant hazards consideration. I 4

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i Attachment D Environmental Assessment Applicability Review 1

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l D. Environmental Assessment Applicability Review Comed has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental assesunent in accordance with 10 CFR 51.21 It has been i determined that the proposed changes meet the criteria for categorical exclusion as provided for under 10 CFR 51.22(c)(9). This conclusion has been determined because the changes requested do not pose significant hazards considerations or do not involve a significant increase in the amounts, and no significant changes in the types of any effluents that may be released off site. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure. j l

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Attachment E Boiling Water Reactor Licensing Methodology Summary, EMF-94-217 1

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