Text

Proposed Tech Specs,Incorporating New Siemens' Methodologies That Will Enhance Operational Flexibility & Reducing Likelihood of Future Plant Derates
	ML20216C638
Person / Time
Site:	Dresden, Quad Cities, LaSalle
Issue date:	08/29/1997
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML17187B089	List: ML20216C638;
References
	NUDOCS 9709090106
	Download: ML20216C638 (49)
	v • d • e

IABLE OF CONTENTS TOC DEFINITIONS SEGIlQH E section 1 DEFINITIONS ACTION........................................... 11 ERAGE PLANAR EXPOSURE (APE) . . . . . . . . . . . . . . . . . . . . . . . 13 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) . . . 11 CHANNEL.......................................... 11 CH ANNEL CALISRATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 C H AN N EL CH E C K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 CH ANNEL FUNCTION AL TEST . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 C O R E A LTER ATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 CORE OPERATING LIMITS REPORT (COLR) . . . . . . . . . . . . . . . . . . . 12 CRITICAL POWER RATIO (Cih . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 DO SE EQUIVALENT l 131 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 FRACTION OF LIMITING POWER DENSITY (FLPD) (applicable t o G E f ue ll . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 FRACTION OF RATED THERMAL POWER (FRTP) . . . . . . . . . . . . . . . 13 FREQUENCY. NOTATION . . . . ........................... -13 FUEL DESIGN LIMITING RATIO (FDLRX) . . . . . . . . . . . . . . . . . . . . . 13 FUEL DESIGN LIMITING RATIO FOR CENTERLINE MELT (FDLRC) ... 13 IDENTIFIED LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . 4 . . . . . . . 13

. LIMITING CONTROL ROD PATTERN (LCRP) .................. 13 LINEAR HEAT GENERATION RATE (LHOR) ................... 13 LOGIC SYSTEM FUNCTIONAL TEST (LSFT) .................. 1-4 4

QUAD CITIES - UNITS 1 & 2 i Amendment Nos. 177 8 175 970909 9 PDR A K0 7 P P

Definiti3ns 1.0 1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be cchieved. The defined terms appear in capitellred type and shall be applicable throughout these Technical Specifications.

ACTION .

l ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions, hERAGE The AVERAGEPLANAR PLANAREXPOSURE EXPOSURE (APE) (APE) shall be applicable to a specific planar height and is l

equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified l height divided by the number of fuel rods in the fuel bundle.

AVERAG PLANAR UNEAR HEAT GENERATION RATE (APU4GR)

The AVERAGE PLANAR UNEAR HEAT GENERATION MATE (APLHGR) shall be applicable to a specific planer height and is equal to the sum of the UNEAR HEAT GENERATION RATE (s) for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL A CHANNEL shall be en arrangement of a sensor and associated components used to evaluate plant variables and generate a single protective action signal. A CHANNEL terminates and loses its identity where single action signals are combined in a TRIP SYSTEM or logic system.

CHANNEL CAU8 RATION A CHANNEL CAUBRATION shall be the adjustment, as necessary, of the CHANNEL output such that it responds with the necessary range and accuracy to known values of the parameter which the CHANNEL monitors. The CHANNEL CAU9 RATION ahall encompass the entire CHANNEL including the required anneor and alarm and/or trip functions, and shaN include the CHANNEL PUNCTIONAL TEST. The CHANNEL CAU9 RATION may be performed by any series of sequential, overlapping or total CHANNEL steps such that the entire CHANNEL is calibtsted.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of CHANNEL behavior during operation by observation. This determination shall include, where possible, comparison of the CHANNEL indication and/or status with other indications and/or status derived from independent instrument CHANNEL (s) measuring the same parameter.

QUAD CITIES - UNITS 1 & 2 11 Amendment Nos, m a ur

SAFETY UMITS 2.1 2.0 SAFETY UMITS AND UMITING' SAFETY SYSTEM SETTINGS M SAFETY UMITS

THERMAt. POWER. Low Pressure or low Flow 2.1. A THERMAL POWER sheit not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPUCABluTY: OPERATIONAL MODE (s) 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.

THERMAL POWER. Hiah Pressure and Hlah Flow l,0q 2.1.8 The MINIMUM CRITICAL POWER RATIO (MCPR) shat not be less thenhith the reactor vessel steam dome. pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of reted flow. During single recirculation loop operation, this MCPR limit shall be increased by 0.01. .

APPLICABluTY: OPERATIONAL MODE (s) 1 and 2.

I ACTION:

With MCPR less than the above applicable limit and the reactor vessel steam dome pressure greater than or equal to 785 peng and core flow greater then or equal to 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.

QUAD CITIES UNITS 1 & 2 21 Amendment Nos. 1" 8 187 i

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS TdN MCM 'THIs -

THER Al POWER Low Pressure or low Flow '

2.1.A T MAL POWER shall not exceed 25% of RATED THERMAL PC ER with the reactor vessel steem dome pressure less than 785 psig or core flow less than 1'0% of rated flow.

/

APPLICABILITY: O RATIONAL MODE (s) 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 185 psig or core flow le'ss than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with t e' requirements of Specification 6.7.

N N

THERMAL POWER. Hioh Pressure and Hlob Flow N

2.1.8

/ \

The MINIMUM CRITICAL POW 1.10* for Unit 2 with the reactor vesse/R l steam RATIO dome brossure (MCPR) greater shall than or equal notpsig to 785 beandless than 1.07 core flow greater than or equal to 10% of rated flow. During singts recirculation bp operation.

this MCPR limit shall be increased 0.01. )

\

APPLICABILITY: OPERATION MODE (s) 1 and 2

\

ACTION:

\

With MCPR less than the above applicable timh and the reactor vessel steam dome pressure Greater than or equal t '785 psig and core flow greater than or equal b 10% of rated flow, be in

et least HOT SHUTD N within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply whh the regulrombts of SpecNication 6.7.

- Y '

I* Applicable

/" Unit 2 for cycle 15 only.

M W'T*E QUAD CITIES UNIT 2

\

21a Amendment No.174 l

BASES Nb b

/ c' approsc Auch of the data indicates that BWR fuel r:an survive for en extended period in an /

environments f transition boiling. / /

C Safety Limit s 1.07, based on General Electric methods for calculating t Safety Umit. The nit 2 MCPR Safety Umit is 1.10', based on Siemens Power Corporation (SPC))

methods for calcul ' g the MCPR Safety Umit. / '

L14 Reseter Cholant System Pressure g{[b The Safety Umit for the r ctor coolant system pressure has been pelected such that it is at a pressure below which it can shown that the integrity of the syyrtem is not endangered. The reactor coolant system integr : an important barrier in the pr ention of uncontrolled release of fission products. It is essential t et the integrity of this syste be protected by establishing a pressure limit to be observed for all perating conditions and honever there is irradiated fuel in the reactor vessel.

The reactor coolant system pressure Safety Umit of 134) psig, as measured by the vessel steam space pressure indicator, is equivalent to'y75 psig at t,he lowest elevation of the reactor vessel.

l The 1375 psig value is derived from the design pressufes of the reactor pressure vessel and coolant system piping. The respective desigrtgres s are 1250 psig at 575'F and 1175 psig at 560'F. The pressure Safety Limit was chosen es lower of the pressure transients permitted by the applicable design codes, ASME Boiler and Pr's ute Vessel Code Section til for the pressure vessel, and USASI B31.1 Code for the reactor e ant system piping. The ASME Boller and Pressure Vessel Code permhs pressure transie s u to 10% over design pressure (110% x 1250

= 1375 psig), and the USASI Code permits ansients p esau%reup to 20% over design pressure (120% x 1175 = 1410 psig). The Safety it press of 1375 psig is referenced to the lowest elevation of the reactor vessel. The desig pressure fo the recirculation suction line piping (1175 psig) was chosen relative to the reactor vessel design pre'asure. Demonstrating compliance of peak vessel pressure with the ASME overpres4ure protection lim't (1375 psig) assures compliance of the suction piping with the USASI limit (1 10 psif. Evaluation ethodology to assure that this Safety Umit pressure is not exceeded for an reload 4 'ocumented the specific fuel vendor. The design basis for the reactor pressur vessel mdes evident the bstantial margin of protection against failure at the safety press e limit of 1375 psig. The vessel has been designed for a general membrane stress no gre or than 26.700 psi at an intomabpressure of 1250 psig; this is a factor of 1.5 below the yield s ngth of 40,100 psi at 575'F. At pressure limb of 1375 psig, the general membrane stress ill only be 29,400 pel, still safely belo the yield strength.

The relationships of stress)evels to yield strength are comparable for the, ary system piping end provides similar mar ' of protection at the established pressure Safe Umit.

The normal operating essure of the reactor coolant system is nominally 1000 psig. Both pressure relief and safety relief valves have been installed to keep the reactor 'wessel peak pressure below 1375 psig. However no credit is taken for relief valves during the postui ed full closure of all MSIVs without 4 direct (valve position switch) scram. Credh, however, is tak for the neutron flux scram. The itect flux scram and safety valve actuation provide adequate margin below the '

allowable peak essel pressure of 1375 p$.

( lie to Unh 2 cycle 15 o R ETE QUA CITIES . UNIT 2 823a Amendment No.174

\

POWER DISTRIBUTION LIMITS APLHGH 3/4.11.A 3.11 UMITING CONDITIONS FOR OPERATION 4.11 SURVEILLAN::E REQUIREMENTS A.- AVERAGE PLANAR LINEAR HEAT A. AVERAGE PLANAR UNEAR HEAT GENERATION RATE GENERATION RATE All AVERAGE PLANAR UNEAR y The APLHGRs shall be verified to be equal GENERATION RATES (APLHGR))for eac to or less than the limits specified in the fysis of fuel as a function of AVERAGE CORE OPERATING UMITS REPORT.

bdC hNAR EXPOSURE'shall not exceed the limits specified in the COME OPERATING 1. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , LIMITS REPORT.

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

OPERATIONAL MODE 1, when THERMAL 3. Init! ally and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months POWER is greater then or equal to 25% of when the reactor is operating with a RATED THERMAL POWER. UMITING CONTROL ROD PATTERN for APLHGR.

l l ACTION: 4. The provisions of specification 4.0.D l are not applicable.

With an APLHGR exceeding the limits specified in the CORE OPERATING UMITS REPORT:

1. Initiate corrective ACTION within 15 minutes, and 4

2. Rostore APLHGR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4 4

QUAD CITIES UNITS 1 & 2 3/4.11 1 Arnendment Nos, maw

Reperting Requirements 6.9 ADMINISTRATIVE CONTROLS (14) ANFB Critical Power Correlation, ANF 1125(Pl(A) and Supplements 1 and 2 Advanced Nuclear Fuels Corporation, April 1990.

(15) Advanced Nuclear Fuels Corporation Critical Power Methodology for Bolling Water Reactors / Advanced Nuclear Fuels Corporation Critical Power Methodology for toiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects /NRC Correspondence, ANF 524(P)(A),

Revision 2, Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.

(16) COTRANSA 2: A Computer Program for toiling Water Reactor Transient Analyses, ANF 913(P)(A) Volume 1 Revision 1 and Volume i Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.

(17) Advanced Nuclear Fuels Corporation Methodology for toiling Water Reactors EXEM BWR Evaluation Model, ANF 9104s(P)(A), Advanced Nuclear Fuels Corporation, January 1993.

(18) Commonwealth Edison Topical Repert NFSR 0091, ' Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods.* Revision 0, Supplements 1 and 2. December 1991, March 1992, and .May 1992, respectively; SER letterdated_ March W1993.

. ' i19)* Comed l9tter, ' Comed Response to NRC Staff Request for Additional D ,

/ Information (RAll Regarding the Application of Siemens Power Corporation \

hleIti f

)

ANFB Critical Power correlation to Coresident General Electric Fuel for LaSalle Unit 2 Cycle 8 and Quad Cities Unit 2 Cycle 15. NRC Docket No.'s 50 373/374 and 50 254/265*, J.B. Hosmer to U.S. NRC, July 2,1996, i'

( transmitting the topical report, Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15, EMF 96-051(P), Siemens Power Corporation + Nuclear Division, May 1996, and f

'ormation. "

J.nserf d --> .

c. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.

6.9.8 Special Reports Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

e lC W

' Applicable to Unit 2 for cycle 15 only.

QUAD ESMtTS+& 616a Amendment Nos. 177 4 175

INSERT A QUAD CITIES Sect:3n 6.9.A.6.b Technical Specifications Insert (19) ANFB Critical Power Correlation Application for Coresident Fuel, EMF-l 1125(P)(A), Supplement 1, Appendix C, Siemens Power Corporation, August 1997.

(20) ANFB Critical Power Correlation Unceriminty (or Limited Data Sets, ANF-1125(P)(A), Supplement 1, Appendix D, Siemens Power Corporation, (DATE TO l BE DETERMINED).

4 s

f I

?

Attachment D :

Marked Up Pages and Inserts for Dresden Technical !

Specifications ,

I i

h i

i' t

9 i

a t

k l 42 l

l-- . . - - . - , . . , . . . . - . , - , . = . . . . - - . - . , - . . - - . . . - - . - _ , . . . - - - . . . -,

TABLE OF CONTENTS TOC DEFINITIONS SECTION P. AGE Section 1 DEFINITIONS -

A C TI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . d 1' hERAGE PLANAR EXPOSURE (APE) .... .. ..... .. . ... 1D AVERAGE PLANAR UNEAR HEAT GENERATION RATE (APLHGR) ... 11 CHANNEL......................................... 11 C H ANN EL C AU BRATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 CH AN N EL CHECK . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 CH ANNEL FUNCTION AL TEST . . . . . . . . . . . . . . . . . . . . . . . . . . 12

, C O R E ALTERATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 CORE OPERATING UMITS REPORT (COLR) .................. 1 2-CRITICAL POWER RATIO (CPR) .......................... 1-2 DOSE EQUIVALENT l 131 .............................. 12 FRACTION OF RATED THERMAL POWER (FRTP) . . . . . . . . . . . . . . . 13 FRE QUENCY NOTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 FUEL DESIGN UMITING MATIC (FDLRX) . . . . . . . . . . . . . . . . . . . . . 13 FUEL DESIGN UMITING RATIO for CENTERUNE MELT (FDLRC) . . . . . 13 IDENTIFIED LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 UMITING CONTROL ROD PATTERN (LCRP) .................. 1-3 UNEAR HEAT GENERATION MATE (LHCR) . . . . . . . . . . . . . . . . . . . 13 LOGIC SYSTEM FUNCTIONAL TE.iT (LSFT) .................. 1-3 MINIMUM CRITICAL POWER RATIO (MCPR) . . . . . . . . . . . . . . . . . . 1-4 OFFSITE DOSE CALCULATION MANUAL IDDCM) . . . . . . . . . . . . . . 14 DRESDEN - UNITS 2 & 3 l Amendment Nos, no a m

_]

Definitions 1.0 1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

ACTION ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions. '

b(Ic k,,

AVERAGE PLANAR EXPOSURE (APE)

The AVERAGE PLANAR EXPOSURE (APE) shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified - ]*

height divided by the number of fuel rods in the fuel bundle, j AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR.1 The AVERAGE PLAN.t.A UNEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the UNEAR HEAT GENERATION RATE (s) for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle. .

CHANNEL A CHANNEL shall be an arrangement of a sensor and associated components used to evaluate plant variables and generate a single protective action signal. A CHANNEL terminases and loses its identity where single action signals are combined in a TRIP SYSTEM or logic sysum.

CHANNEL CAllBRATION i

A CHANNEL CAUBRATION shall be the adjustment, as necessary, of the CHANNEL output such that it responds with the necessary range and accuracy to known values of the parameter which the CHANNEL monitors. The CHANNEL CAUBRATION shall encompass the

, entire CHANNEL including the required sener and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHAf#4EL CAU8 RATION may be performed by any asties of sequential, overia'pping or total CHANNEL steps such that the entire CHANNEL is calibrated. '

CHANNEL CHECK '

A CHANNEL CHECK shall be the qualitative assessment of CHANNEL behavior during operation by observation. This determination shall include, where possible, comparison of the CHANNEL indication and/or status with other indications and/or status derived from independent instrument CHANNEL (s) measuring the same parameter.

DRESDEN - UNITS 2 & 3 11 Amendment Nos. iso a 145 f

SAFETY LIMITS 2.1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS M SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow

2.1.A - THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABill',y,:, OPERATIONAL MODE (s) 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.

THERMAL POWER, Hioh Pressure and Hioh Fir.; lp 2.1.B The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less thahith the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow. During single recirculation loop operation, this MCPR limit shall be increased by 0.01.

APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.

I ACTION:

With MCPR less than the above applicable limit and the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow, be in ct least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.

DRESDEN - UNITS 2 & 3 2-1 Amendment No,s. 150 s its

POWER DISTRIBUTION UMITS APLHGR 3/4.11.A 3.11 UMITING CONDITIONS FOR OPERATION 4.11 SURVEILLANCE REQUIREMENTS A. AVERAGE PLANAR UNEAR HEAT A. AVERAGE PLANAR UNEAR HEAT GENERATION RATE GENERATION RATE All AVERAGE PLANAR UNEAR HEAT The APLHGRs shall be verified to be equal to or less than the limits specified in the GENERATION _ RATES (APLHGR)[for e

~

type of fuel as a function of bundle average CORE OPERATING LIMITS REPORT.

@ lek v exposurefiell not exceed the limits specified in the CORE OPERATING UMITS 1. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , REPORT.

2. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least APPUCABlUTY: 15% of RATED THERMAL POWER, and OPERATIONAL MODE 1, when THERMAL 3. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months POWER is greater than or equal to 25% of when the reactor is operating with a RATED THERMAL POWER. UMITING CONTROL ROD PATTERN for APLHGR.

ACTION: 4. The provisions of Specification 4.0.D are not applicable.

With an APLHGR exceeding the limits specified in the CORE OPERATING UMITS ,

REPORT: ,

(

1. Initiate corrective action within 15 minutes, and

2. Restore APLHGR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . .

With the provielons of the ACTION above ,

not met, reduce THERMAL POWER to less than 25% of RATED' THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

DRESDEN - UNITS 2 & 3 3/4.11-1 Amendment Nos. iso a us

POWER DISTRIBUTION LIMITS SLHGR 3/4.11.0 3.11 LIMITING CONDITIONS FOR OPERATION 4.11 SURVEILLANCE REQUIREMENTS D. STEADY STATE LINEAR HEAT D. STEADY STATE LINEAR HEAT GENERATION RATE bgg 7f GENERATION RATE The LIN5AR HEAT GENEBATlRN RATE Y The SLHGR shall be determined to be equal (LH Rjjor each type of fuel as a function 3 to or less than the limit:

of AVERAGE PLANAR EXPOSUREAhall not xceed the STEADY STATE LINEAR HEAT 1. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , GENERATION RATE (SLHGR) limits specified in the CORE OPERATING UMITS 2. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a REPORT. THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and APPLICABILITY: 3. Initially end at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a OPERATIONAL MODE 1, when THERMAL UMITING CONTROL ROD PATTERN for POWER is greater than or equal to 25% of SLHGR.

' RATED THERMAL POWER.

4. The provisions of Specification 4.0.D are not applicable.

With an LHGR exceeding the SLHGR limits i

specified in the CORE OPERATING LIMITS I

REPORT:

1. Initiate corrective ACTION within 15 minutes, and

2. Restore the LHGR to within the SLHGR limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

DRESDEN - UNITS 2 & 3 3/4.11 4 Amendment Nos. tse a 145

REACTOR CORE 5.3 5.0 DESIGN FEATURES L3 REACTOR CORE Fuel Annamblian gg 5.3.A The reactor core shall contain 724 fuel assemblie Each assembly consists of a l matrix of Zircoloy clad fuel rods with an i-itial composition of natural or slightly enriched uranium dioxide as fuel material. The assemblies may contain water rods or a water box. Limited substitutions of Zircaloy or ZlRLO or stainless steel filler rods for fuel rods, in accordance with NRC approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes a ethods, and shown by tests or analyses to comply with all fuel safety design base

JA limited number of lead test assemblies that have not completed representative testing may be placed in non -

limiting core regions.

Control Rod Assemblies 5.3.8 The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (BeC) and/or hafnium metal. The control rod assembly shall have a nominal axial absorber length of 143 inches.

t h6Ic.kd.

1. ATRIUM 98 fuel with exception of leed test eesemblies is ordy enowed in the reactor core in Operational Modes 3,4 and i 5 and vnth no more then one control rod withdrawn, for urut 2 only. I 2 operation in oil modes with ATMluM-98 fuel is allowed for Dreeden. Unit 3. Cycle 1s, on;y.

3 The design beoes applicable to ATRIUM 98 fuel are those which are applicable to operational Modes 3,4, and 5, for Urut 2 only. ~~ ...____..

DRESDEN - UNITS 2 & 3 55 Amendment Nos. 160 8 155 u I

ADMINISTRATIVE CONTROLS

b. The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of topical reports:

(1) ANF 1125(P)(A), ' Critical Power Correlation - ANFB."

(2) ANF 524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors."

(3) XN NFh9 71(P)(A), Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors.*

(4) XN NF 8019(P)(A), " Exxon Nuclear Methodology for Boiling Water Reactors."

(5) XN NF 85 67(P)(A), " Generic Mechanical Design for Exxon Nuclear Jet Pump Boiling Water Reactors Reload Fuel."

(6) ANF 913(P)(A), *CONTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis.*

(7) XN NF 82 06(P)(A), Qualification of Exxon Nuclear Fuel for Exterded Burnup

- Supplement 1 Extended Bumup Qualification of ENC 9x9 BWR Fuel, Supplement 1, Revision 2, Advanced Nuclear Fuels Corporation, May 1988.

(8) ANF 8914(P)(A), Advancsd Nuclear Fuels Corporation Generic Mechanical Design for Advance Nuclear Fuels Corporation 9x9 IX and 9x9 9X BWR Reload Fuel, rievision 1 and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, October 1991. . .

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

. Revision 1 and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995. .

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

(11) Commonwealth Edison Company Topical Report NFSR 0091, " Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods", and associated Supplements on Neutronics Licensing Analyses (Supplement 1) and 1.a Salle County Unit 2 Benchmarking (Supplement 2).

W jyasert 0 DRESDEN UNITS 2 & 3 6 15 Amendment Nos. 160 & 155

INSERT B DRESDEN Section 6.9.A 6.b Technical Specification Insert (12) ANF-1125(P)(A), ANFB Critical Power Correlation Uncertainty For Limited Data Sets, Supplement 1, Appendix D, Siemens Power Corporation. (DATE TO BE DETERMINED).

I

Attachment E Marked Up Pages and Inserts for LaSalle Unit 1 Technical Specifications 43

_____~

%aET7Eh I DEFINITI 5 h"

SECTIO [

1.0 DEFINITIONS ffftE 1.1 ACTI0N........................................................... 1-1 1.2 AV E RAG E PLANAR EX PO SUR E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .~ . . . . 1 - l' dek 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE....................... 1-1 1.4 CHANNEL CALIBRATION.............................................. 1-1 1.5 CHANNEL CHECK.................................................... 1-1 1.6 CHANNEL FUNCTIONAL TEST.......................................... 1-1 1.7 CORE ALTERATION.................................................. 1-2 1.8 CORE OPERATING LIMITS REP 0RT..................................... 1-2 1.9 CR IT I CAL POWER RAT I 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.10 DOSE EQUIVALENT I-131............................................ 1-2 1.11 E-AVERAGE DI SINTEGRATION ENERGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME............... 1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME........ 1-2 l

1.14 FRACTION OF LIMITING POWER DENSITY............................... 1-3 1.15 FRACTION OF RATED THERMAL P0WER.................................. 1-3 1.16 FREQUENCY N0TATION............................................... 1-3 1.17 GASEOUS RADWASTE TREATMENT SYSTEM................................ 1-3 1.18 IDENTI F I ED LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.19 ISOLATION SYSTEM RESPONSE TIME................................... 1-3 1.20 DELETED.......................................................... 1-4 1.21 LIMITING CONTROL ROD PA1 TERN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.22 LINEAR HEAT GENERATION RATE...................................... 1-4 1.23 LOGIC SYSTEM FUNCTIONAL TEST..................................... 1-4 1.24 MAXIMUM FRACTION OF LIMITING POWER DENSITY....................... 1-4 1.25 MEMBER (S) 0F THE PUBLIC.......................................... 1-4 1.26 MINIMUM CRITICAL POWER RATI0....................................., 1-4 LA SALLE - UNIT 1 I Amendment No.110 1

c. b

4.0 DEFINITIONS following terms are defined so that uniform interpretation of these speci-

/ .u gb ' ations may be achieved. The defined terms appear in capitalized type and all be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

[VERAGE PLANAR EXPOSURE)PbELE"if -

Tyzk L 1.2 kAVERAGE PLANAR EXPOSURE shall be applicable to a specific planar i

heightandisequaltothesumoftheexposureofallthefuelrodsin\

hthespecifiedbundleatthespecifiedheightdividedbythenumberof )

Uuel rods in the fuel bundle. "

AVERAGE PLANAR LINEAR HEAT GENERATION RATE

1. 3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK

1. 5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST

1. 6 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a situlated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

LA SALLE - UNIT 1 1-1

L

{ SAFETY LIMITS BASES 2.1.2 THERMAL POWER. Hlah Pressure and Hiah 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 damaga could occur. Although it is recognized that a departure fros' nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the 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 is determined using the ANF Critical Power Methodology for boiling water reactors (Reference 1) which is a statistical model that 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 date d using the SPC-developed ANFB critical power correlation, ghcq The bases for the un rtain ies in system-related parameters are presented in NEDO-20340, Reference found in References 1, 3 The bases for the fuel-related uncertainties are 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 Methodoloey for Boiling Water Reactors: Methodolog Assembly Channel Bowing Effects /NRC Correspondence,y for Analysis of IN-NF-524(P)(A)-

Revision 2 and Supplement 1 Revision 2, Supplement 2. Advanced Nuclear Fuels Corporation, November 1990.

-2. Process Computer Performance Evaluation Accuracy, NED0-20340 and Amendment 1, General Electric Company, June 1974 and December 1974, respectively.

3. ANFB Critical Power Correlation, ANF-ll25(P)(A), and Supplements 1 and 2, 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,-Advanced Nuclear Fuels Corporation, November 1990.

5. 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, March 1983.

-+

LA SALLE - UNIT 1 B 2-2 Amendment No.116 h6W~

o

}

INSERT C LASALLE UNIT 1 Bases Section 2.1.2 Technical Specifications insert

6. ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1125(P)(A), Supplement 1 Appendix C, Siemens Power Corporation, August 1997.

7_.

ANFB Critical Power Correlation Uncertainty for Limited Data Sets, ANF-1125(P)(A), Supplement 1, Appendix D, Siemens Power Corporation,-(DATE TO BE DETERMINED).

3/4.2 POWER MSTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CON 0' TION FOR OPERATION 3.2LA.11 AVERAGE PLANAR LINEAREAT GENERATION RATES (APLHG95)ff

~

Qg (of fuel as a function of AVERAGE PLXNKR UF03URE7ih~alrnot exceed the limits specifica in the CORE OPERATIETIMITFREPORT ' l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits specified in the 001E OPERATING LIMITS !

REPORT, initiate corrective action within 15 minutes anc restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

I SURVEILLANCE REQUIREMENTS 4.2.1 All APL%Rs shall be verified to be eaual to or less than the limits -

specified in the CORE OPERATING LIMITS REPORT.

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

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

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

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

BASES 3/4.2.4 LINEAR HEAT GENERATION RATE GE Fuel The specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated. The effects of fuel densification are discussed in the General Electric Standard Application for Reactor Fuel (GESTAR), NEDE-240ll-P-A. The GESTAR discusses the methods used to ensure LHGR remains below the design limit.

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 l during normal operation or anticipated operational occurrences (A00s).

Operation above the LHGR limit followed by the occurrence of an A00 could i 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 A00s 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 term 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 LHGR factor (LHGRFAC ) corresponding to the existing,) core flow and power. The LHGRFAC multiplierI 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 c, ore flow transients. The applicable LHGRFAC 4HGRFA(mult4 pliers-afs apectffed-in-th

^

on d. Det Pum YeV'5IDrs -for- 2ELA

$3M Alu F-Q l- 04S(JXM ,p Mode l&ppkme,d /, Semenc hw po.wMn R' "" c ' '

't E To e be reRmINEb. a ,j i

1. Advanced cle Fue s orat ogy or Boiling Water Reactors EXEM BWR ECCS Evaluation iodel, ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation, January 1993.

2. 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, March 1983.

3. Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, XN-NF-80-19(P)(A), Volume 3 Revision 2, Exxon Nuclear Company, January 1987.

LA SALLE - llNIT 1 B 3/4 2-5 Amendment No.116

ADttlNISTRATIVE CONTROLS Core Operatino timits Report (Continued)

(16) Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-19-71(P)( A), Revision 2 Supplements 1, 2, and 3. Exxon Nuclear Company, March 1986.

(17) Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)j:A), Revision 1 and Revision 1 Supplement 1 Advanced Nuc mar Fuels Corporation, May 1995.

(18) NEDE-24011-P-A, " General Electric Standard Application for Reactor fuel," (latest approved revision).

(19) Comonwealth Edison Topical Report NFSR-0085, " Benchmark of BWR Nuclear Design Methods," (latest approved revision).

(20) Comonwealth Edison Topical Report NFSR-00B5, Supplement 1,

" Benchmark of BWR Nuclear Design Methods - Quad Cities Gama Scan Comparisons," (latest approved revision).

(21) Comonwealth Edison Topical Report NFSR-0085, Supplement 2,

" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

(22) " Benchmark of Comonwealth Edison Topical Report NFSR-0091, Revision D, Supplements 1 and 2, December CASM0/MICR0 19BWR l, March Nuclear 1992, respectively; SERBURN Desihn Methods "

and May j letter dated March 22, Y

Insu+ D LA SALLE UNIT 1 6-25b Amendment No. 116

INSERT D l LASALLE UNIT 1 Section 6.6.A.6.b Technical Specifications insert (23) BWR Jet Pump Model Revision for RE! AX, ANF-91-048(P)(A), Supplement 1 Siemens Power Corporation, (DATE TO BE DETERMINED).

l (24)

ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1125(P)(A), Supplement 1, Appendix C, Siemens Power Corporation, August 1997.

(25) ANFB Critical Power Correlation Uncertainty for Limited Data Sets, ANF-1125(P)(A), Supplement 1, Appendix D, S',emens Power Corporation, (DATE TO BE DETERMINED).

Attachment F Marked up Pages and Inserts for LaSalle Unit 2 Technical Specifications 44

.. 1

l DELET ED /

MDEX

\

l DEFINITIO

f SECTION[

1.0 0[FINITIONS PJfd 1.1 ACTI0N........................................................... 1-1

~-

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

y 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE....................... 1-1 1.4 CHANNEL CALIBRAT10N.............................................. 1-1 1.5 CHANNEL CHECK.................................................... 1-1 1.6 CHANNEL FUNCTIONAL TEST.......................................... 1-1 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.................................. 1-2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME............... 1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME........ 1-2 1.14 0ELETED.......................................................... 1-3 1.15 FRACTION OF RATED THERMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.16 FREQUENCY NOTAT!0N............................................... 1-3 1.17 GASEOUS RADWASTE TREATHENT SYSTEM................................ 1-3 1.18 IDENTIFIED LEAKAGE...........................'.................... 1-3 1.19 ISOLATION SYSTEM RESPONSE TIME................................... 1-3 1.20 DELETED.......................................................... 1-3 1.21 LIMITING CONTROL ROD PATTERN..................................... 1-4 1.22 LINEAR HEAT GENERATION RATE...................................... 1-4 1.23 LOGIC SYSTEM FUNCTIONAL TEST..................................... 1-4 1.24 DELETED.......................................................... 1-4 1.25 MEMBER (S) 0F THE PUBLIC.......................................... 1-4 1.26 MINIMUM CRITICAL POWER RATI0..................................... 1-4 LA SALLE - UNIT 2 I Amendment No. 101

h 1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these speci-fications may be achieved. The defined terms appear in capitalized type and l shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

hRAGE PLANAR EXPOSURE [b6LET6 g 1.2khe AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in (thespecifiedbundleatthespecifiedheightdividedbythenumberof fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The~ AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION

1. 4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the e channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channti steps such that the entire channel is calibrated.

CHANNEL CHECK

1. 5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips,

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL 5EST eay be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is tested.

LA SALLE - UNIT 2 1-1 11 L ~

SAFETY LIMITS BASES

' 2.1.2 THERMAL POWERI Minh Pressure and Hiah 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 t'anage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical-power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.95 of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using the ANF Critical Power Methodology for boiling water reactors (Reference 1) which is a statistical model that combines al of the uncertainties in operation parameters and the procedures used to calculate critical power. The probability of the occurrence ci boiling transition is deterni m using the SPC-developed ANFB critical power correlation. r g The bases for the un rtai in system-related parameters are presented in NED0-20340, Reference The bases for the fuel-related uncertainties are foundinReferences1,3-/ The uncertainties used in the analyses are

, provided in the cycle-specific transient analysis parameters document.

1. Myanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors /Mvanced 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, and Supplement 1 Revision 2 Supplement 2, Mwanced Ndclear Fuels Corporation, November 1990. .

2. Process Computer Performance Evaluation Accuracy, NEDO-20340 and Amendment 1 General Electric Company, June 1974 and December 1974, respectively.

3. ANF8 Critical Power Correlation, ANF-ll25 (P)(A), and Supplements 1 and 2, Myanced Nuclear Fuels Corporation, April 1990.

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

5. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, IN-NF-80-19(P1(A) Volume 1 and Supplements 1 and 2 Exxon Nuclear Company,~ Marc h7983'.

6. " Application of the ANFB Critical Power Correlation to Coresident GE Fuel' for LaSalle Unit 2 Cycle 8," EMF-96-021(P), Revision 1, Siemens Power g Corporation, February 1996; hRC SER letter dated September 26, 1996. )

4 ./~

U h60.Y $

LA SALLE - UNIT 2 B 2-2 Amendment No. 101 l

4 .

INSERT E LASALLE UNIT 2 Bases Section 2.1.2 Technical Specifications insert

6. ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1125(P)(A), Supplement 1. Appendix C, Siemens Power Corporation, August 1997.

7. ANFB Critical Power Correlation Uncertainty for Limited Data Sets, ANF-1125(P)(A), Supplement 1, Appendix D, Siemens Power Corporation, (DATE TO BE DETERMINED).

l

, 3/4.2 POWER DISTRIBUTION LIM!'5 '

3/4.2.1 AVERAGE PLANAR LINEAR 4 EAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.21 All AVERAGE PLANARJ IMEAR_ HEAT GENERAL 10MJtAIEUAPLHGRs)

Delp'H. (of fuel as a function of AVERAGE PLANAR EXPOSURE /shall not exceed the limits 7peciried in T.no (,uRMPERATIKIIMIT57EPOR1. i APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMA' POWER. .

ACTION: ,-

With an APLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . .

i SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits specified in the CORE OPERATING LIMITS REPORT.

\

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

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

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

LA SALLE - UNIT 2 3/4 2-1 Amendment No. 54

314.2 POWER DISTRIBUTION LIMITS RASES 3/4.2.4 LINEAR HEAT GENERATION RATE (Continued) 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 durinfi continuous operation with LHGRs up to the limit defined in the CORE OPERA"ING LIMITS REPORT. The analysis also includes allowances for short tem 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 durinti limiting transients. At reduced power and flow conditions, the LHGR imit is reduced (multiplied) using the smaller of either the flow dependent LHGR factor LHGRFAC to the e(xisting,) core flow and power.or The the power-dependent LHGRFAC m LHGR factor (LHGRFAC ) corre protect the core during slow flow runout transienks.ultiplier,s The LHGRFAC are used to multipliers are used to protect the core during plant transients other than cIre flow transients. The applicable LHGRFAC, and LHGRFAC, multipliers are specified in the CORE OPERATING LIMITS REPORT.

References:

f 1. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors ,

EXEM BWR ECCS Evaluation Model, ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation, January 1993. j

." 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, March 1983.

3. Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Themal Limits Methodology Summary Description, XN-NF-80-19(P)(A), Volume 3 Revision 2. Exxon Nuclear Campany, Jammary 1987.

4. Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-71(P)(A) Revision 2 Supplements 1, 2, and 3 Exxon Nuclear Company, March 1986.

5. COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.

l \

mg --

C LA SAllt - UNIT 2 B 3/4 2-5 Amendment No. 101 9 &nd. BWR Te+ bP / N el Feui.sio~ b 2EA46 ,

Ap)F- Q\-O% CP)LQ Spplaned I , S&* s W W P "*I'o~i (

Lb ATE 77> BE hETEJEntrNfb). y

s. _ - y _ ______ - j-J

ADMIN 15T8tATIVE CONTROLS

. Core Onaratina Limits annart (Continued)

(9) Generic Mechanical Desion for Exxon Nuclear Jet Pump BWR Reload Fuel IN-NF-85-67 Company, tamber 1986.(P)(A) Revision 1, Exxon Nuclear (10) Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9xD-sK OWR Reload Fuel ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2. Detober 1991.

(11) Volume 1 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain Volume 2 - STAIF - A

~ .

Camputer Program for SWR Stability Analysis in the F c EMF-CC-07 Af, Domain,CodeQualificationReportSiemensPouerCorporation,Ju (12) REDEIZ Fuel Red Themal-Mechanical Response Evaluation IN-NF-81-58 Model, Exxon Nuclear Campa(P)(A), Revisiun 2 Supplements 1 and 2, ny, March 1984.

,<.; . : w .wrc w*.n n (13) XCD8RA-T: A er Code for BWR Trans ont Themal-raulicCoreglysis,IN-NF-84-105 Volume 1 and Volume Advanced Nuclear 1 Supplements Fuels Corporation,1 February and 1 Supp 2;1987 Volume an (P) A)lement 1988, respectively.

... g..t. . w . o. .

(14) Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A),

- ep&. .$ ,"""

. **E.. Fuels Corporation, January 1993.

. (15),

Exxon Nuclear Methodology for toiling Water Resctors -

A Neutronic Methods for Design and Analysis,

'5 XN-if-80-19(P)(A)Richland, Nuclear .Campany, WA 99352, March 1983. Volume 1 and Supple a 1.. , . ,

(16) Exxon Nuclear Plant Transient Methodology for Boilina Water Reactors, IN-NF-79-71(P)(A), Revision 2 D plements I, 2

.A and 3. Exxon Nuclear Company, March 1986.

. h%. .,'d,$hr$i%.a. . .

(17) Generic Mechanical Design Criteria for BWR Fuel Designs ANF-89-98(P)fA), Revision 1andRevision1 Supplement 1, ,

2.w Advanced Nuc lear Fuels Corporation, May 1995.

. w g ., ~ . y (18)

,o,..Reactor NEDE-24011-P-A,(latest Fuel," approved revision)." General Electric Standa (19) Commonusalth Edison Tosical Report NFSR-0085, ' Benchmark of f; ; g(Nuclear, Design Mettods,' (latest approved revision).

(5)' CommNnSealth Edison Topical Report NFSR-0085, Supplement 1,

. ". " Benchmark of BWR Nuclear Design Methods - Quad Cities

.. Gamma Scan Comparisons," (latest approved revision).

(21)_ Commanusalth Edison Topical Report NFSR-0085, Supplement 2,

' Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

(22) Commonusalth Edison Topical Report NFSR-0091, " Benchmark of CASMD/MICRDBURN BWR Nuclear Desi Supplements 1 and 2 December 19b, March Methods" Revision 0 1992,andMay

. f,7 , 1992, respectively:,SER letter dated March 22, 1993.

3 n

3^ns e r+ F LA SALLE UNIT 2 6-25 a Amendment No.101

INSERT F LASALLE UNIT 2 Section 6.6.A.6.b Technical Specifications Insert (23) BWR Jet Pump Model Revision for RELAX, ANF 91-048(P)(A), Supplement 1 Siemens Power Corporation, (DATE TO BE DETERMINED).

(24) ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1125(P)(A), Supplement 1, Appendix C, Siemens Power Corporation, August 1997.

(25) ANFB Critical Power Correlation Uncertainty for Limited Data Sets, ANF-1125(P)(A), Supplement 1, Appendix D, Siemens Power Corporation, (DATE TO BE DETERMINED).

ATTACHMENT G EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS G. EVALUATION OF SIGNIFICANT ll A7.ARDS CONSIDERATIONS Adding References 1 and 7 to Technical Speci0 cation Section 6 and applying these methods at Comed BWRs is evaluated for significant hazards consideration in this section. These documents have been submitted to the NRC under separate correspondence. References 1 and 7 are in NRC review, and require approval to be insened into Section 6.

l Comed has evaluated the proposed Technical Specification amendment and determined it does not represent a significant hazards consideration. Based on the criteria for defming a significant hazard consideration established in 10CFR50.92(c), operation of Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1 and 2, in accordance with the proposed amendments, will not represent a significant hazards consideration for the following reasons:

These changes do not:

1. Involve a significant increase in the probability or consequences of an accident previously evalueted.

The piobability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences.

Limits nave been established consistent with NRC approved methods to ensure that fuel perfonnance during normal, transient, and accident conditions is acceptable. These changes do not affect the operability of plant systems, nor do they compromise any fuel performance limits.

Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)

The Reference i methodology to be added to the Technical Specifications is used as part

- of the LOCA analysis and does not introduce physical changes to the plant. The Reference I revised jet pump model changes the calculational behavior of the jet pump under reversed drive flow conditions. The revised jet pump model methodology makes the LOCA model behave more realistically and calculates small break LOCA PCTs that are comparable to the large break LOCA results. Therefore, this change only affects the methodology for analyzing the LOCA event and determining the protective APLHGR limits. The Technical Specification requirements for monitoring APLHGR are not affected by this change. The revised method will result in higher APLHGR limits, thus the SPC fuel will be allowed to operate at higher nodal powers. The approved methodology, however, still protects the fuel performance limits specified by 10CFR50.46. Therefore, the probability or consequences of an accident previously evaluated will not change.

46

ATTACHMENT G EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Addition of SPC Generie Methodology for Application of ANFH Critical Power Correlation to Non SPC Fuel (Quad Cities Units 1 and 2 and LaSalle Units I and 2)

The probability or consequences of a previously evaluated accident are not increased by adding Reference 3 to Section 6.9.A.6.b of the Quad Cities Technical Specifications and Bases Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical Specifications.

Reference 3 determines the additive constants and the associated uncertainty for application of the ANFB correlation to the coresident GE fuel. Therefore, it provides data that is used in the determination of the MCPR Safety Limit. This approved methodology for applying the ANFB critical power correlation to the GE fuel will protect the fuel from boiling transition. Operational MCPR limits will also be applied to ensure that the MCPR Safety Limit is protected during all modes of operation and anticipated operational occurrences. Because Reference 3 contains conservative methods and calculations and because the operability of plant systems designed to mitigate any l consequences of accidents have not changed, the probability or consequences of an i

accident previously evaluated will not increase.

Addition of SPC Topical for Revised ANFH Correlation Uncertainty (Quad Cities Units 1 I and 2, Dresden Units 2 and 3, and LaSalle Units 1 and 2)

The probability or consequences of a previously evaluated accident is not increased by adding Reference 7 to Section 6.9.A.6.b of the Quad Cities and Dresden Technical Specifications and Bases Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical Specifications. Reference 7 documents the additive constant uncertainty for SPC (

ATRIUM-9B fuel design with an internal water channel. This methodology is used to I determine an input to the MCPR Safety Limit calculations, which ensures that more than 99.9% of the fuel rods avoid transition boiling during nonnal operation as well as anticipated operational occurrences. This change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. This methodology for determining the ATRIUM-9B -additive constant uncertainty for the MCPR Safety Limit calculation will continue to support protecting the fuel from boiling transition. Operational MCPR limits will be applied to ensure the MCPR Safety Limit is not violated during all modes of operation and anticipated operational occurrences. Therefore, no individual precursors of an accident are affected and the operability of plant systems designed to mitigate the probability of consequences of an accident previously evaluated are not affected by these changes.

47 i

ATTACHMENT G EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Change to Minimum Critical Power Ratio Safety Limit (Quad Cities Units 1 and 2 and Dresden Units 2 and 3)

Changing the htCPR Safety Limit at Quad Cities Units 1 and 2 and Dresden Units 2 and 3 will not increase the probability of an accident previouQ evaluated. This change implements the htCPR Safety Limits resulting from the SPC ANFB critical power i correlation methodolc gy using a revised additive constant unentainty from Reference 7.

The hiCPR Safety Limit of 1.09 that is proposed for Quad Cities Units I and 2 and Dresden Units 2 and 3 is anticipated to be conservative and acceptable for future cycles.

Cycle specific htCPR Safety Limit calculations will be performed, consistent with SPC's approved methodology, to confirm the appropriateness of the h1CPR Safety Limit.

Additionally, operational $1CPR limits will be applied that will ensure the htCPR Safety Limit is not violated during all modes of operation and anticipated operational occurrences. Changing the htCPR Safety 1.imit will net alter any physical systems or operating procedures The h1CPR Safety Limit is set to 1.09, which is the CPR value where less than 0.1% of the rods in the coic are expected to experience boiling transition.

This safety limit is expected to be applicable for future cycles of ATRIUht 9B at Dresden and Quad Cities. Therefore the probability or consequences of an accident will not increase.

Removal of Footnotes Limiting Operation with ATRIUM 911 Fuel Reloads (Quad Cities Unit 2 and Dresden Unit 3)

The removal of footnotes from the Quad Cities and Dresden Technical Specifications does not involve any significant increase in the probability or consequences of an accident previously evaluated. The footnotes were added to clarify that cycle specific methods were used until the generic methodology was approved by the NRC, Since the NRC has approsed SPC's generic methodology for application of the ANFB correlation to the coresident GE fuel (Reference 3) and SPC has addressed the conceros regarding the database used to calculate the ATRIUht 9B additive constant uncertainties (Reference 7),

the footnotes are no longer necessary. The removal of the Unit 2 specific "a" pages,2 la and B2 3a, in the Quad Cities Technical Specifications is justified by the removal of the footnotes. Therefore, removing these footnotes and "a" pages does not require any physical plant modifications, nor does it physically affect any plant components or entail changes in plant operation. Therefore, the probability or consequences of an accident previously evaluated is not expected to increase.

48

ATTACHMENT G EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Revision to Thermal 1.imit Descriptions (Quad Cities Units I and 2. Dresden Units 2 and 3, and LaSalle Units I and 2)

The revision to the Section 3 Technical Specification description of the APLilGR limits has no implications on accident analysis or plant operations. The purpose of the revision is to allow Ocxibility for the MAPLilGR limits and their exposure basis to be specined in the COLR and to establish consistency with approved methodologies currently utilized by Siemens Power Corporation, which calculates MA?LilGR limits based on bundle or planar average exposures. This revision also provides for consistency in the APLilGR l limit Technical Specification wording between the Comed DWRs. The revision to the 3.11.D SLilGR Technical Specification for Dresden also has no implications on accident l analysis or plant operations. The purpose of this revision is to allow nexibility for the LilGR limits and their exposure basis to be specined in the COLR. This revision makes the Dresden LilGR dennition consistent with NUREG 1433/1434 wording. The l

deunition of the Average Planar Exposure is deleted, becau the exposure basis of the APLilGR is being removed. Therefore, no plant equipment or processes are affected by this change. Thus, there is no alteration in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated:

Creation of the possibility of a new or difTerent kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications to the plant configuration, including changes in allowable modes of operation. His Technical Specification submittal does not involve any modifica' ions to the plant configuration or allowable modes of operation. No new precursors of an accident are created and no new or different kinds of accidents are created. Therefore, the proposed changes do not create the possibility of a new or ditTerent kind of accident from any accident previously evaluated.

Addition of SPC Hevised Jet Pump Methodology (LaSalle Units 1 and 2)

The revised jet pump model methodology will be used to analyze the LOCA fbr LaSalle Units I and 2, and does not introduce any physical changes to the plant or the processes used to operate the plant. This change only alTects the methods used to analyze the LOCA event and detennine the MAPLilGR limits. Therefore, the possibility of a new or diD'erent kind of accident is not created.

49

ATTACHMENT G EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Addition of SPC Generie Methodology for Application of ANFil Critical Power Correlation to Non SPC Fuel (Ouad Cities Units 1 and 2 and LaSalle Units I and 2)

Addition of the generic methodology for the application of the ANFB critical power correlation to GE fuel in Section 6.9.A.6.b of the Quad Cities Technical Specifications and Bases Section 2.1.2 and Section 6.6.A.6.b of the I.aSalle Technical Specifications does not introduce any physical changes to the plant, the processes used to operate the plant, or e.llowable modes of operation. This change only involves adding an NRC approved methodology, which is used to determine the additive constants and additive constant uncertainty for GE feel, to Section 6 of the Technical Specifications. Therefore, no new precursors of an accident are created and no new or different kinds of accidents are created.

Addition of SPC Topleal for Revised ANFil Correlation Uncertainty (Quad Cities Units I and 2. Dresden Units 2 and 3 and LaSalle Units I and 2)

Addition of the Reference 7 methodology to Section 6.9.A.6.b of the Quad Cities and Dresden Technical Specifications and Bases Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical Specifications will not create the possibility of a new or different kind of accident from any accident previously evaluated. This methodology describes the calculation of an input to the htCPR Safety Limit the ATRIUM 9B additive constant uncertainty. Therefore, no new precursors of an accident are created and no new or ditTerent kinds of accidents are created.

Change to Minimum Critical Power Ratio Safety Limit (Quad Cities Units I and 2 and Dresden Units 2 and 3)

Changing the MCPR Safety Limit will not create the possibility of a new accident from an accident previously evaluated. This change will not alter or add any new equipment or change modes of operation. The MCPR Safety Limit is established to ensure that 99.9%

of the rods avoid boiling transition.

The MCPR Safety Limit is changing for Quad Cities Unit I due to the transition to SPC ATRIUM 90 fuel and SPC methodologies. The MCPR Safety Limit is changing for Quad Cities Unit 2 due to the Reference 7 methodology, which documents a 0.0195 ATRIUM 9B additive constant uncertainty and supports a 1.09 MCPR Safety Limit.

This MCPR Safety Limit is lower than the current MCPR Safety Limit for Quad Cities Unit 2,1.10, which is based on a higher interim conservative additive constant uncertainty of 0.029. The lower ATRIUM 9B additive constant uncertainty results in the lower MCPR Safety Limit for Quad Cities Unit 2. The new MCPR Safety Limit for Dresden Units 2 and 3,1.09, is greater than the current value at Dresden Units 2 and 3 50

ATTACHMENT C l

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS and is being increased now in anticipation of bounding future reloads of ATRIUM.911.

Therefore, no new accidents are created that are different from any accident previously evaluated.

Removal of Footnotes Limiting Operation with ATRIUM 911 Fuel Reloads (Quad Cities Unit 2 and Dresden Unit 3)

The removal of the footnotes from the Quad Cities and Dresden Technical Specifications does not create a new or different kind of accident from any accident previously evaluated. The removal of the footnotes does not affect plant systems or operation. The footnotes were temporarily established to implement a conservative cycle specific MCPR Safety 1.imit until the SPC generic methodology was approved. With the approval of the generic Reference 3 methodology and the anticipated approval of the Reference 7 additive constant uncertainty methodology, these footnotes are no longer applicable. The removal of the Unit 2 specific "a" pages,2.la and ll2 3a, in the Quad Cities Technical Specifications which is justified by the removal of the footnotes, also does not create a new or difTerent kind of accident from any accident previously evaluated.

Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle I and 2)

The revision of the APLilGR and LilGR limit descriptions will not create the possibility of a new or different kind of accident from any accident previously evaluated. This revision will not alter any plant systems, equipment, or physical conditions of the site.

This revision allows the flexibility of the APLilGR and the LilGR limits to be specified in the COLR and to maintain consistency with the calculated results of nothodologies currently used to determine the APLilGR. The definition of the Average Planar Exposure is deleted, because it is being remosed from LilGR and APLilGR Technical Specifications.

3, involve a significant reduction in the margin of safety for the following reasons:

Addition of SPC Revised Jet Pump Methodology (LaSalle Units I and 2)

The revised jet pump model methodology, and the MAPLilGRs, resulting from the revised jet pump methodology, will continue to ensure fuel design criteria and 10CFR50A6 compliance. The results of LOCA analyses performed with this 51

ATTACHMENT G EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS methodology must continue to comply with the requirements of 10CFR50.46. Therefore.

there is no significant reduction in the margin of safety.

Addition of SPC Generle Methodology for Appileation of ANFH Critical Power Correlation to Non SPC Fuel (Quad Cities Units I and 2 and LaSalle Units 1 and 2)

The margin of safety is not decreased by adding this reference to Section 6.9.A.6.b of the Quad Cities Technical Specincations and Bases Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical Specifications. Siemens Power Corporation methodology for application of the ANFB Critical Power Correlation to coresident GE fuel is approved by i the NRC and is the same methodology used in the cycle specine topical for coresident fuel (Reference 4 and $). The MCPR Safety Limit will continue to ensure that greater than 99.9% of the rods in the core avoid boiling transition. Additionally, operating limits will be established to ensure the MCPR Safety Limit is not violated during all modes of operation.

Addition of SPC Topleal for Revised ANFH Correlation Uncertainty (Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units I and 2)

The MCPR Safety Limit provides a margin of safety by ensuring that less than 0.1% of the rods are expected to be in boiling transition if the MCPR Safety Limit is not violated.

This Technical Speci6 cation amendment proposes to insert the topical report that describes SPC's calculation of the ATRIUM 9B additive constant uncertainty. The new ATRIUM 9B additive constant uncertainty calculation is conservative and is based on a larger database than previous calculations. Because a conservative method is used to calculate the ATRIUM 9B additive constant uncertainty, a decrease in the margin to safety will not occur due to adding this methodology to the Technical Specincations. In addition, operational limits will be established to ensure the MCPR Safety Limit is protected for all modes of operation. This revised methodology will only ensure that the appropriate level of fuel protection is being employed.

Change to Minimum Critical P ,wer Ratio Safety Limit (Quad Cities Unit I and 2 and Dresden Units 2 and 3)

Changing the MCPR Safety Limit for Quad Cities and Dresden will not involve any reduction in margin of safety. The MCPR Safety Limit provides a margin of safety by ensuring that less than 0.1% of the rods are expected to be in boiling transition if the MCPR Safety Limit is not violated. The proposed Technical Specincation amendment reflects the MCPR Safety Limit results from conservative evaluations by SPC using the l

52

ATTACHMENT O EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS ANFil critical power correlation with the new 0.0195 ATRIUhi 9B additive constant uneenainty documented in Reference 7.

Because a conservative method is used to apply the ATRIUhi 9B additive constant uncertainty in the hiCPR Safety Limit calculation a decrease in the margin to safety will not occur due to changing the htCPR Safety Limit. The revised h1CPR Safety Limit will ensure the appropriate level of fuel protection. Additionally, operational limits will be ertablished based on the proposed hiCPR Safety Limit to ensure that the hiCPR Safety Limit is not violated during all modes of operation including anticipated operation occurrences. This will ensure that the fuel design safety criterion of more than 99.9% of the fuel rods avoiding transition boiling during nonnal operation as well as during an anticipated operational occurrence is met.

Removal of Footnotes Limiting Operation with ATRIUM 9H Fuel Reloads (Quad Cities Unit 2 and Dresden Unit 3)

The removal of the cycle specific footnotes in Quad Cities and Dresden Technical Specifications does not impose a change in the margin of safety. These footnotes were added due to concems regarding the calculation of the additive constant uncenainty for the ATRIUht 9B fuel and the cycle specific application of the ANFB critical power correlation to coresident GE fuel in Quad Cities Unit 2 Cycle 15. Because the generic ANFB application to coresident GE fuel htCPR methodology (Reference 3) has received NRC approval and the topical report describing the increased database used to calculate the additive constant uncertainties for ATRIUht 9B (Reference 7) have been submitted to the NRC and both are pronosed to be added to the Technical Specifications in this amendment, there is no reason for the footnotes to remain. Removal of the Unit 2 specific "a" pages, 2-la and B2 3a, in the Quad Cities Technical Specifications is justified by the removal of the footnotes. Therefore, the removal of the "a" pages,2 la and B2 3a, also does not impose a change in the margin of safety.

Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1 and 2)

The revision to the APLiiGR and LliGR limit descriptions will not involve a reduction in the margin of safety. The methodology used to calculate the APLilGR must comply with the guidelines of Appendix K of 10 CFR Part 50, and the APLilGR and LilGR will still be required to be maintained within the limits specified in the COLR. The surveillance requirements for these two thermal limits remain unchanged. Thus, there will be no reduction in the margin of safety.

53

ATTACHMENT G EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS 1his proposed amendment does not involve a significant relaxation of the criteria used to establish the safety limits, a significant relaxation of the bases for the limiting safety system settings, or a significant relaxation of the bases for the limiting conditions for operations.

l Therefore, based on the guidance provided in 10CFR50.92(c), the proposed change does not constitute a significant hazards consideration.

54

Attachment II Environmental Assessment Applicability Review 1

l 55

ATTACHMENT H ENVIRONMENTAL ASSESSMENT APPLICABILITY REVIEW H. ENVIRONMENTAL ASSESSMENT APPLICAlllLITY REVIEW ComlId has evaluated the proposed amendment against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with 10CFit$1.21. It has been determined that the proposed changes meet the criteria for categorical exclusion as provided for under 10 CFit $1.22(c)(9). This conclusion has been determined because the changes requested do not pose signincant hazards considerations and do not involve a signincant 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 I or cumulative occupational radiation exposure.

56

N Attaclariierit I References I

i 57

l ATTACHMENT l REFERENCES l

1. REFERENCES

1. ANF 91048(P), Supplement 1,"BWR Jet Pump Model Revision for RELAX",

Submitted to the NRC by SPC letter,"ANF 91048(P), Supplement 1 and ANF-91048(NP),

Supplement 1,"BWR Jet Pump Model Revision fbr RELAX," RAC:96 042 R.A. Copeland to US NRC, May 6,1996.

2. XN NF 8019(P)," Exxon Nuclear Methodology for Boiling Water Reactors -- Volume 2A, RELAX: A RELAP4 Based Computer Code for Calculating Blowdown Phenomena" June 1981.

3. EMF-1125(P)(A), Supplement 1 Appendix C. "ANFB Critical Power Correlation Application for Coresident Fuel", August 1997, and NRC SER," Acceptance for Referencing of 1.icensing Tepical Report EMF-1125(P), Supplement 1 Appendix C, 'ANFB Critical Power Conelation Application for Co. Resident Fuel", J. E. Lyons to R. A. Copeland, May 9, 1997.

4. EMF 96 021(P), Revision 1 " Application of the ANFB Critical Power Correlation to Coresident GE fuel for LaSalle Unit 2 Cycle 8", February 1996, and NRC SER, " Safety Evaluation for Topical Report EMF--95 021 (P), Revision 1, ' Application of the ANFB Critical Power Conclation to Coresident GE Fuel for LaSalle Unit 2 Cycle 8' (TAC No.

M94964)", D.M. Skay to 1. Johnson, September 26,1996.

5. EMF-96 051(P)," Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15", May,1996, and NRC SER," Approval of Topical Report EMF 96-051(P) - Quad Cities, Unit 2 (TAC NO. M96213)", R. Pulsifer to 1. Johnson, May 16,1997.

6. ANF-1125(P)(A) Supplements I and 2 "ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation", April 1990.

7. ANF-1125(P), Supplement 1, Appendix D,"ANFB Critical Power Correlation Uncertainty For Limited Data Sets", Submitted to the NRC by SPC letter," Request for Review of ANFB Critical Power Correlation Uncertainty for Limited Data Sets, ANF ll25(P), Supplement 1, Appendix D", llDC:97:032,11. D. Curet to Document Control Desk, April 18,1997.

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

9. "Dresden Nuclear Power Stations Units 2 and 3 Application for Amendment of Facility Operating License DPR 19 and DPR 25 Technical Specifications, NRC Docket Nos. 050-237 and 050 249," J.S. Perry to US NRC, June 20,1996.

58

ATTACHMENT I REFERENCES

10. "Dresden Nuclear Power Station Units 2 and 3 Supplement to Application for Amendment of Facility Operating i icenses DPR 19 and DPR 25 Technical Specifications", J.S. Perry to US NRC, December 30,1996.

I1. "Dresden Nuclear Power Station Units 2 and 3 Supplement to Application for Amendment of Facility Operating 1.icense DPR 19 and DPR 25 Technical Specifications", J.S. Perry to US NRC, h1 arch 5,1997.

12."LaSalle County Nuclear Power Station Units I and 2 Application for Amendment Request to Facility Operating Licenses NPF ll and NPF 18. Technical Specifications Changes for Siemens Power Corporation Fuel Transition Docket Numbers 050 373 and 050 374", R.E.

l Querio to US NRC, April 8,1996.

l 13."LaSalle County Nuclear Power Station Units I and 2 Supplement to Application for Amendment of Facility Operating Licenses NPF il and NPF-18, Appendix A, Technical Specification Changes for Siemens Power Corporation Fuel Transition", W.T. Subalusky to U.S. NRC, October 14,1996.

14. " Quad Cities Nuclear Power Stations Units 1 and 2, Application for Amendment Request to Facility Operating Licenses DPR 29 and DPR 30. Technical Specification Changes for Siemens Power Corporation (SPC) Fuel Transition, Docket Nos. 50 254 and 50-265", E.S.

Kraft, to USNRC, June 10,1996.

15." Quad Cities Nuclear Power Stations Units I and 2 Supplement to Application for Amendment of Facility Operating License DPR 29 and DPR 30 Technical Specifications",

E.S. Kran to US NRC, February 17,1997, 16." Quad Cities Nuclear Power Station Units 1 and 2 Exigent Application for Amendment Request to I acility Operating Licenses Pursuant to 10CFR50.91(a)(6), DPR 29 and DPR 30.

Technical Specification Changes for Revised hiinimum Critical Power Ratio Safety Limit for Quad Cities Unit 2 Cycle 15, Docket Nos. 50-254 and 50 265", E.S. Kraf, Jr. to USNRC, April 21,1997, 17," Quad Cities Nuclear Power Station Units 1 and 2, Emergency Application for Amendment to Facility Operating Licenses Pursuant to 10CFR50.91, DPR 29 and DPR-30, Operation with ATRIUht 9B Fuel in hiodes 3,4, and 5", E.S. Kraft. Jr. to USNRC, April 29,1997.

i 59

ML20216C638

Text

References:

Navigation menu

Search