ML20129B798

Proposed Tech Specs 2.1.2 Re Thermal Power,High Pressure & High flow,5.3 Re Reactor core,4.0 Re Design features,3/4.1.3 Re Control Rods & 3/4.2.4 Re Linear Heat Generation Rate

ML20129B798
Person / Time
Site:	LaSalle
Issue date:	10/14/1996
From:	COMMONWEALTH EDISON CO.
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ML20129B773	List: ML20129B768 ML20129B798
References
NUDOCS 9610230114
Download: ML20129B798 (23)

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, REVISED TECHNICAL SPECIFICATION PAGES FOR LASALLE UNITS 1 AND 2 LASALLE UNIT 1 LASALLE UNIT 2 S 2-2 B 2-2 .

'r. sert 2 Insert 2 )

5-4 5-4 )

Insed 10 Insed 10 4 6-25 6-25

) Insed 9 Insed 9 8 3/4 1-4 B 3/4 1-4 8 3/4 2-6 B 3/4 2-6 Insed 6 Insed 6 Insed 7 Insed 7 I

1 l

l l

l 9610230114 961014 PDR ADOCK 05000373 P PDR

4 j SAFETY UMITS d

i SASES 1 '

l 2.1.2 THERMAL POWER. Ninh Pressure and High Flow -

t i 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 j which result in fuel demage are not directly observable during reactor operation.

the thermal and hydraulic conditions resulting in a departure from nucleate l boiling have been used to aerk 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 1 in the procedures used to calculate the critica power result in an uncertainty j in the value of the critical power. Therefore, the fuel cladding integrity i 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 i transition considering the power distribution within the core and all

, uncertainties.

l The Safety Limit l

lysis Basis, GETAB,MCPR is determined using the General Electric Thermal, w i uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is

' detaruined using the General Electric Critical Quality (X) Boiling Length (L),

GEXL correlation.

l The bgses for the uncertainties in the core parameters are given in NED0-20340 and in NED0-10958-A,the . The power basis for the uncertainty distribution is based in on the GEXL a typical correlation 764 assembly is given 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 any fuel cycle would not be as severe 4

as the distrhtion used in the analysis.

\

! a. " General Electric BWR Thi. mal Analysis Bases (GETAB) Data, Correlation j and Design Application," NED0-10958-A.

! b. General Electric " Process Computer Performance Evaluation Accuracy"

NEDD-20340 and Admenchment 1, NED0-20340-1 dated June 1974 and December 1974, respectively.

l l _

\i I

%ed #?  %]

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

4

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 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 determined using the SPC-descloped ANFB critical power correlation.

4 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 and Supplement 1 Revision 2, Supplement 2, Advanced Nuclear 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. ANFB Critical Power Correlation, ANF-1125 (P) (A) and Supplements I 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, Advaro:d Nuclear Fuels Corporation, l November 1990. l l

5. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and )

Analysis, XN-NF-80-19(P)(A) Volume 1 and Supplements I and 2, Exxon Nuclear Company, March 1983.

I

'DE5!dNFEATURES l

5.3 REACTOR CORE .

g /,cs eJ pg N f ,w /

FUEL ASSEMBLIES .

5.3. he reactor core shall contain 764 fuel assemblies. Each assemely onsists of a matrix of 7.irca11oy clad fuel rods with an initial composition of slightly enriched uratiium dioxide, UD . Fuel assemblies shall be limited I to those fuel designs approved for use i$ SWR's. d l-k -

CONTROL ROD ASSEM LIES l 5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemolies. i The control satorial shall be boron carbide power (8 C) and/or hafnium metal. .-

The control rod assemoIy shall have a nominal axial , absorber length of 143 inches.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance.for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of:

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. ,

l VOLUME i

1 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. l 5.5 METEOROLOGICAL TOWER LOCATION .

5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

LA SALLE - UNIT 1 .

5-4 Amendment No. ?0

J Ossign Featuras 4.0 3 '

! 4.0 DESI A$ er l W /0 TWLES f

. Si tion [T t ion]

4.2.1 uel Ass ==ihlien 76 l The reactor shall contain fuel asses 411es. Each assambly

! shall consist of a matriz of irtalloy = !!" a A l rods with i an initial composition of natural or slightly enr;iched uranium j dioxide  %) as fuel materiabd rf -2r =] . Limitad suhetituttaae af -t;::-t;; :1. ~ er stainless steel filler rods EI@ y for feel rods, is accordance vi approved applications of fuel

! rod configurations, may be used. Fuel assemblies shall be limited Or to those fuel designs that have analyzed with applicable 1stC i staff approved UAas and authods shoun by tests er analyses to

!. @lgt.O comply with all safety design bases. A limited member of lead

' test assemblies that have set camp 1 representative testing may-be placed la analisiting core regions.

! 7/ 0, t 4.2.2 drol Red Ass ==h11as

~

nkaA./4 ,.ds .<-sea As** r.

The re core shal n 133 crucifors sh=w N1 rod

as, iss. i teriai slin ne careide, hafn l .s 1] as by the IRC. ,

i

/ ,

/' / '

! 4. Fuel Stora

' / ,.

4.3.1 iticality

'/ 4.3.1.1 The fuel storage maintained with:

are design and shall be '

, j l /

/

a. Fme/ l assemblies ving a marium -infinity efil.31] in Ahe normal re core confi ici. at co1Jgonditions]

l .

/ [ average U- enrichment of .5] weight ];

/ b. Q g 0. if fully f1 with unbora unter, ubich inclad an allomance uncertainti as described la

/ [ Sect 4an 9.1 of the  ; /

/ / /

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% inued)

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l Sint/6 STS '

4.0-1 1,04/07/95 j / .

, 4DMINISTRaTTVE CONTROLS

} Semiannual Radienetive Effluent Release Renert (Continued) j Any changes to the OFFEITE 005E CALCULATION MANUAL shall be abeitted with the Monthly Operating Re '

1 change vs. port within 90 days In addition, in which a report of any theajor i changesto(s) was made effect' the 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 l Investigative Function.

, E. tore Deeratine timits Renert -

s

a. Core ting limits shall be established and documented in the

CDRE O

TING LIMIT 3 REPORT before each reload cycle or any remaining part of a reload' cycle for the following

l T

Eff=* ef I (1) Technical specification 3.L1The Avercge Planar Linear Heat Ge j

on/yse/ . , M-D i

,~

U (2) The minians Critica matte 1

.p s. r. 1CpR)its n'.  !

'r ' k, and.f.power and l #"

flowdeendent36 ICR lim  !

tat-for Technical $pecification f ,Y.h/e/) I*I 3 .

! (3) The Linear Heat3.2.4.

Specification Generation Rate (LHGR) for Technical )

(4) The Rod Block Monitor Upscale Instrissentation setpoists for

' Technical specification Table 3.3.5-2. 1 i b. The analytica1' methods used to determine the core operaties y limits shall be those previousl J^5*#

l

1. 4 j ""'i[MisNN5$'y reviewed and approveH by thes. '

unit 1, the topical reports are:

! L i

1 h(tg/NEDE-24011-p-A, Reactor Fuel," ( latest

" General Electric standardapproved Applicatism for rwision i

- M Commonwealth Edison ToHeal Report, a'5R-0085, "Senchmark of . '

[g[ BWR Nuclear Design ;;n.4,* (latest approved revision).

M Coun.nnwealth Edison Topicai NFSR-0085, Supp~lement L

• Renchmark of SWR Nuclear Desi Methods - Quad C1 ties IN, Gamma scan Comparisons.* (la approved revision).

,,pr Commonwealth Edison Topical RepoM N' FSR-0085, Supp1mmest 1, pf; Licensing Analyses,' (latest approved revision).*Renchm .

\

(p) c . e.N( sd:ss. 7& rce/ Repet AffSA* **

..g,,dag , f cdsno/hicroftdAt Skf ) tacky l l

i h;to "

rfee 3l lII{ $C IWh LA SALLE - UNIT 1 6-25 Amendment Es. 103

{ .

1. 0tch l9931 and IO s Wsfecfiyefyj$ck l

-ldler de fe lh arcis 12, 1993,

)

Insert #9

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

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 Water Reactors / Advanced Nuclear Fuels Corporation Critical Power Methodology for Beiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing EfTects/NRC Correspondance, XN-NF-524(P)(A) Revision 2 and Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.

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

5. HUXY A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option, ANF-CC-33(P)(A) Supplement 1 Revision 1; and Supplement 2, Advanced Nuclear Fuel Corporation, August 1986 and January 1991, respectively.

6. Advanced Nuclear Fuel 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.

7. Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon Nuclear Company, June 1986.

8. 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.

9. Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(P)(A) Revision 1, Exxon Nuclear Company, September 1986.

10. Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel, ANF-89-014(P)(A), Revision I and Supplements 1 and 2 October 1991,

11. Volume 1 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain, Volume 2 - STAIF - A Computer Program for BWR Stability Analy sis in the Frequency Domain, Code Qualification Report, EMF-CC-074(P)(A), Siemens Power Corporation, July 1994.

i Insert #9 (continued) i 4

12. RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model,  ;

XN-NF-8158(PXA), Revision 2 Supplements 1 and 2, Exxon Nuclear Company, March 1984.

13. XCOBRA-T: A Computer Code for BWR Transient Thermal-H)draulic Core Analysis, XN NF 105(PXA), Volume I and Volume 1 Supplements I and 2; Volume 1 Supplement 4, Advanced i Nuclear Fuels Corporation, February 1987 and June 1988, respectively. I

14. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91048(P)(A), Advanced Nuclear Fuels Corporation, January 1993. l l

15. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and i Analysis, XN NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Company, l Richland, WA 99352, March 1983. l

16. Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN NF-79 71(PXA), Resision 2 Supplements 1,2, and 3. Exxon Nuclear Company, March 1986.

17 Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(PXA), Resision 1 and Revision 1 Supplement 1. Advanced Nuclear Fuels Corporation, May 1995.

)

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i-i BATET ,

t 3/4.h.3C6HTR6iR66,( Castinued) ,

' (see Table In addition, 1.2.1-1 the antaustic CID charging water header law pressure scras bility to insert initistas well before any accumulator loses its full capa-a control red. With this added automatic scram feature -

3 the surveillance of each individual accumulator check valve is no longer necessar

• j action. y to demonstrate adequata stored energy is available for normal scram 1

i analysis of the rod drop accident le the FIALControl rod coupling in 1he overtravel posities feature provides the only positive means of determining that a red is properly c i

i . and therefore this check must be perfomed prior to achieving criticaJity a

! 'co eting ing integrity. CORE ALTERATIONS that osuld have affected the control red drive i

I initial demonstration.The subsequent check is performed as a %== to the 1

j therefore that other parameters ass withis their limits, position indicati'on systam mest be OPERA 8LL s

{ .

The control rod housing i

i control red to less thae 3.s_ support restricts the outuard sevement of a I amount of rod reactivity which coult' he added by this small amo -

i j withdrawai is less than a samal withdrawal increment and to any damage to the primary coolart system.

j housing.there is no pressure to act as a driving, force to rapidly sject a

! The required su[veillance intarvals b adagneta to datamine that the j rods are OPERABLE and not se n p.a as to cause excessive usar en the components.

• k 3/4.1.4 CDNTROL ROD PRoena_g um,noig i -

that the maximum insequence individual control red er

to result control rod in a peak fusi enthalpy greater than 28D cal /g t

i homogeneous drop accident.

! scattered patterms of centrol red withdrawalThe specified s=fa is greater than 105 of RATE THEBlv. POWER, themssible is aorod po.worth When THERMAL pol which -

if dropped at the design rate of the velocitylimitar, could result is a peak enthalpy of 280 cal /gs. Thus requiring the Rift to be OPERABLE uhes .

THERMAL adequata control. POWER is less than or equal to 105 of RATB THERMAL POWE '

will not be withdrawn or insertad.The Rial provide automatic sapentston to a

. The analysis of the rod drop accident is pmsamtad in Section 15.9 of OX'4)

.the FSAR t .s and gy the techniques e.

-a-------

- of Xu-the analysis are prese hlse he&Urg fv BAAyVd*'An'd"J " N*d t "' #p-p/r,&

hedh Dery, a.) Ad,lrinf vdurcl ad SupW*fs I ~d d Hard Iff3 i LA SALLE 'tD(IT 1 33/41-4 -

Amendment No. 29 m__ _ . _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l ,

POWER DISTRIBUT10N SYSTD6 ,

BA5Es 3/4.2.4' LINEAR NEAT CENERATION RATE

m .

! Tie specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in I

any rod is less than the desian linear haat eeneranion even if fuel cellet .

r th

densifd entina is noctulate)d 't ift is ed

! al "s p en un ett 3. . of i repo 73

une 1 arl nere v ati n 1 s e t and pa ass a ence at no an ne j ~),o exc s des L1 ON d to r ik l

j

References:

d (b5*d H 4sre]

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

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

i

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

! Boiling Water Reactors,' General Doctric Company Licensing Topical

{ Report NEDD 24154 Vols. I and II and NEDE-24154 vol. III as sup-

! plemented by letter dated September 5,1980, from R. H. Buchholz l (GE) to P. $. Check (NRC).

! 3. "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-

! Coolant Accident Analysis," General Doctric Company Report

! NEDC-32258P, October 1993.

s

! 4. " General Doctric Standard Applic'ation for Reactor Fuel,'

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

5. " Extended Operati Domain' and Equipment Out-of-servicegfor LaSalle County Nuclear St ion Units 1 and 2,* NEDC-31435, November 1987.

6.

• ARTS Improvement Program Analysis for: LaSalle County Units 1 and 2,*J General Doctric Company Report NEDC-31531P, December 1993. j lzsarf #7 dere]

4,. e.pfeAr af A/ de,r.-4; cd.~ , are bsc um./*

r., the Go,eal E/ecfric &l'd d# bc'N' f,, pachc fue/ (Gssr#s), JEff- 29M~M

& ((SMg l S(4/sef fem .r USel N .

g,,9,e lllSg remi r SeN flv brf b* W-LA SALLE - UNIT 1 B 3/4 2-6 Amendment No. 103

' ~

insert #6 i

4 i SPC Fuel 1 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 j axial location. LHGR limits are specified to ensure that fuel integrity 1

limits are not exceeded during normal operation or anticipated operational occurrences (AOOs). Operation above the LHGR limit followed by the occurrence of an AOO could potentially result in fuel damage and subnquent release of radioactive material. Sustained

operation in excess of the LHGR limit could also result in exceeding i the fuel design limits. The failure mechanism prevented by the

! LHGR limit that could cause fuel damage during AOOs is rupture of l the fuel rod cladding caused by strain from the expansion of ths fuel l pellet. One percent plastic strain of the fuel cladding has been 4

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 l 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 lie reduced to ensure adherence to the fuel mechanical design bases during limiting transients. At reduced power and flow conditions, i the LHGR limit is reduced (multiplied) using the smaller of either the j 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 l 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 specified in the CORE OPERATING LIMITS REPORT.

i 4

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

k Insert #7

1. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR ECCS Evaluation Model, ANF 91048(P)(A), Advanced Nuclear Fuels Corporation, January 1993.

l i 2. Enon Nuclear Methodology for Boiling Water Reactors. Neutronic Methods for Design and l- Analysis, XN-NF 80-19 (P)(A), Volume I and Supplements 1 and 2, Enon Nuclear Company, March 1983.

l t .

3. Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, XN-NF-8019 (P)(A), Volume 3 Revision 2 Enon Nuclear Company, 1 January 1987.

i

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

Revision 2 Supplements 1,2, and 3 Enon Nuclear Company, March 1986.

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

$ 6. XCOBRA T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, XN NF i L 105(P)(A) Volume I and Volume 1 Supplements I and 2. Enon Nuclear Company, February 1987.

7. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF 89 98(P)(A) Revision 1, and l Revision i Supplement 1 Advanced Nuclear Fuels Corporation. May 1995. ,

l .

i 8. LaSalle County Station Units I and 2 SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis, j NEDC-32258P, General Electric Company, October 1993.

9. ARTS Improvement Program analysis for LaSalle County Station Units I and 2, NEDC-31531P, l General Electric Company, December 1993.

I i

h 1

i i

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,, , __ . . . , _ - - - _ = . , _ . . - -_ .

SAFETY LIMITS l

l l j BASE 5 2.1.2 THERMAL POWER. High Pressure and,High Flowl damage l

The fuel cladding integrity SafetySince Limit is set such that the parameters during reactor operation.

l is which calculated result in fuel to damageoccur ifare thenotlimit is not l

directly t e from observab violated.

nucleate e

fuel damage i the themal and hydraulic conditions resulting hin a depar ur l

boiling have been used to markd the beginning the critical power at of th could occur. not necessarily result in damage to 8WR dfuel pted as a convenient l

would ro s, t and which limit.

boiling transition is calculated to lt in an occur has b uncertainty in the procedures u6ed to calculate the critical power l for which resuTherefore in the value of the critical power. d to avoid boilingfuel assem l Safety Limit is defined as the CPR in the limiting e and all

more than 99.9% of the fuel rods in the cote are uncertainties. -

determined using the General Electric ll of theThemal 1

! Analysis The Safety LimitGETA8,MCPR Basis, . which is a statistical is model ed thatto combines calculate a

uncertainties critical power.

in operating parametersBoiling andLangth the procedu (L),

determined using the General. Electric critical Quality (X) 4 j

GEXL correlation.

r--

,N5 tf f h S We )

l

" General Electric BWR Thermal Analysis Bases (G a.

and Design Application," NEDD-10958-A.

(

Amendment No. 41 8 2-2 LASALLE - UNIT 2

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

i k

l Insert #2 b

l 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 j boiling transition is determined using the SPC-developed 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 in References 1,3-6. 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 and Supplement i Revision 2, Supplement 2, Advanced Nuclear 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.- ANFB Critical Power Correlation. ANF-1125 (P)(A) and Supplements I 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.

a_

l 5. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A) Volume 1 and Supplements I and 2, Exxon Nuclear Company, March 1983.

6. " Application of the ANFB Critical Power Correlation to Coresident GE Fuel for LaSalle Unit 2 Cycle 8," EMF 96-021 (P), Revision I, Siemens Power Corporation, February 19%; NRC SER letter dated September 26,19%.

l d

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.=

i .

.0ESIGN' FEATURES i

5.3 REACTOR CORE d ## #N FUEL ASSEMBLIES M I4JF' I!O b [ eactor core shall contain 764 fuel assemblies. Each assemely i rconsists of a matrix of Zircalloy clad fuel rods with an initial composition

! of slightly enriched uranium dioxide UDs. Fuel assemblies shall be lisi pthosefueldesignsapprovedforuseinSWR's.

CONTROL R00 ASSDSLIES

! 5.3.2 The reactor core shall contain 185 cruicfore shaped control rod i assemolies. The control material shall be boron carbide powder (8 C) anvor j hafnium metal. The control rod assembly shall have a nominal axial absorter length of 143 inches. I j 5.4 REACTOR COOLANT SYSTEM

! DESIGN PRESSURE AND TEMPERATURE I

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

i . .

l a. In accordance with the code requirements specified in Section 5.2 i

of the FSAR, with allowance for normal degradation pursuant to the j applicable Surveillance Requirements, a

j b. For a pressure of: *

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

l 2. 1650 psig free the recirculation pump discharge to the outlet j side of the discharge shutoff valve.

4

3. 1500psigfromthedischargeshutoffvalvetothejetpumps.

j c. For a temperature of 575'F.

VOLUME i

5.4.2 Tha totai water and steam volume of the reactor vessel and reci.rculation j system is

• 21,000 cubic feet at a nominal 7,,,of 533*F.

l 5. 5 METEOROLOGICAL TOWER LOCATION i

5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

6 1

}

i i \

LA SALLE - UNIT 2 5-4 Amendment No.54 i

9 1

Design Features 4.0

, 4.0 DE3! TM f A$trl h/0 1

l . si tion [T tion of : tion) 1 -

,. . /

nv i 4.2.1 uel Asaggtligg 4

76 i The mactor shall contain feel assemblies. Each assembly i shall consist of a astriz of ircalley "i" -T.e1 rods with

] an initial composition of natural er slightly enr;iched uranium i dioxide (US,) as fuel estaria'd. rf r t - c ). Limited suhetitartiame af_ ';Zt f,' ~ WP stainless steel filler rods

• for feel rods, in accordance vi approved applications of fuel i

fiNallQj l rod configurations, may be used. Fuel assemblies shall be limited

! Or to those fuel designs that have analyzed with applicable K l staff approved codes and methods shown by tests or analyses to

$1RLO cowly with all esfety ensism bases. A iinited esiber of 1.as

! test assamblies that have set coup 1 representative testing may-be placed ta aca11 siting core regions.

'. 7Ao An/As A S.

4.2.2 M Red Assa=hlies _

! The core shal a control rod 4

ass tas. I material abcrucifomil be sh carbide, hafn

/

u- = th. = . .

1

/- / -

4. Fuel Stora

.d

/

4.3.1 itiemlity

/ 4.3.1.1 The fuel storage are desi and shall be maintained with: -

, f

/

3~

/

a. Fe/ el asseu611es ng a marians ' -infinity efl1.31] in normal re core confi lim.atce1Jgooditions]

[averageU-enrichment of .5] weight ];

/ b. 5 0. if fully fl with unbora ustar which ine ud an allowance encartainti asdescrIbedla

./ [Sectiong.1ofthe  ;

l ,

,/ ,s ,/

i / /

j

/

% inued)

! 8WR/6 STS 4.0-1 ' 1,04/07/95 1 ,

/

Core Onoratine limits Renert (Continued) l l

3 (1) The Average planar Linear Heat Generation Rate (APUffR) for j gggg Technical specification 3.2.1.g, g I

, aw/gss' (2) The minime Critical Ratio (MCPR 'ishdh; M

'" spendent MCPR lim)its, and power and i g jg [.g g/fv;ce 4- xrs tir, dent fl t MPit sitst for Technical specification l 3y de (ore 7.d /dj (3) The Linear Heat Generation Rate (LHER) for Techn .

$ pacification 3.2.4.

(4) The Rod Block Monitor Upscale Instrumentation Satpoints for i Technical specification Table 3.3.5-2.

b. The analytical methods used to datemine.the com operati )

i limits s_ hall be .those previousl,y reviewed and approved.15 e 1 L Serf ,I

____---->--2__ _ _ _ -,-_ .

.q l

_n ; f" ~31 Z l'

• U " E E C 'For

~

,asa11e'dunty 5ta U o~n j N 'l Unit 2, the topi31 reports art:

a A l NEDE-24011-p-A ' General Electric standard Application for

/ Reactor Fuel,',(latest ' approved revision).

c 4

.J2t Commonwealth Edisse Tasical Report NF3R-0085, "Benihmark of l l (g) IWR Nuclear Design Metuds," (latest approved revision).

.J37 Commonwealth Edison Topical Report NF5R-0085, Susplement 1,

/,ap *Senchmark of EWR Nuclear Design Method - Guad cities l Samma scan Comparisons," (latest approved revision).

- J47 Commonwealth Edison Topical Report NFSR-0085, Supplement 2, "Sanchmark of BWR Nuclear Design Methods - Neutronic l , (g) Licensing Analyses," (latest approved revision (. '

i c.-

i 11ie applicable corelimits operating (limits'shall e.g., fuel be determined thermal-ecchani' N that all limits, core l

nuclear limits such as I j

thermal-hydraulic shutdown liar.ts. and margin, and transient ECCSacc Limits,ident analysis limits) o  !

the safety analysis are met.  !

1

d. The CDRE OPERATING LIMITS REPORT, including any mid-cycle -

revisions or supplements thereto, shall be provided issuance, for each reload cycle to the U .3. Nuclear platory -

Commission Document Control DesE with copies to the <onal Administrater and Resident laspector.

3. Deleted. ,_

(3) (*~~~.~* /M Eak TopW foprf NFCht%

j i #Bedart a f C4tho//rletosw' BW #<6-pgyi,fjeNolC , i ve t ' grarve'ramioO a

. gg, ton (s4&eds Iand w , resee% kc'*W <*p

}"I's im' a gg nherte dded flotsh 1% iffL \

LA SALLE - UNTT 2 6-25 Amendment No. 88

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

i' ,

j .

j .

,e i

i

Insert #9

1. ANFB Critical Power Correlation, ANF-1125(P)(A) and Supplements 1 W 2. Advanced Nuclear i Fuels Corporation, April 1990,

. 2. Letter, Ashok C. Thadani (NRC) to R. A. Copeland (SPC)," Acceptance for Referencing of ULTRAFLOW Spacer on 9x9-IX/X BWR Fuel Design,"

July 28,1993.

3. 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 Correspondance, XN NF-524(P)(A) Revision 2 and Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.

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

5. HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option, ANF-CC-33(P)(A) Supplement 1 Revision 1; and Supplement 2, Advanced Nuclear Fuel Corporation, August 1986 and January 1991, respectively.

6. Advanced Nuclear Fuel Methodology for Boiling Water Reactors, XN NF-8019(P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.

7. Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, XN NF-80-19(P)(A), Volume 4 Revision 1, Exxon Nuclear Company, June 1986.

8. Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description, XN-NF-8019(P)(A), Volume 3, Revision 2. Exxon Nuclear Company, January 1987.

9. Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF 85-67(P)(A) Resision 1, Exxon Nuclear Company, September 1986.

10. Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9 9X BWR Reload Fuel, ANF-89-014(P)(A), Revision I and Supplements I and 2, October 1991.

11. Volume 1 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain, Volume 2 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain, Code Qualification Report, EMF-CC-074(P)(A), Siemens Power Corporation, July 1994.

Insert #9 (continued)

12. RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model, XN NF-8158(P)(A), Revision 2 Supplements I and 2, Enon Nuclear Company, March 1984.

13. XCOBRA T, A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, XN-NF 105(P)(A), Volume I and Volume i Supplements I and 2; Volume 1 Supplement 4, Advanced Nuclear Fuels Corporation, February 1987 and June 1988, respectively.

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

15. Euon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN NF-80-19(P)(A) Volume I and Supplements I and 2, Exxon Nuclear Company, Richland, WA 99352, March 1983.

16. 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.

l

17. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A), Revision 1 and  ;

Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995. I l

1 l

l l

l l

REACTIVITY CONTROL TYSTD5 f

f -

BA5Es .

i 3/4.1.3 CONTROL RODS (Continued) l f In addition, the automatic CRD charylag water bander low pressure scram "

l (see Table 1.1.1-1) isitistas well before any accumulator loses its es11 capa-l i

bility te insert the control red. With this added automatic scram feature, the surveillance of each individual accessister check valve is no longer l

necessary to demonstrate adequate stored energy is available for moraal scras action.

j Control rod coupling integrity is required to ensure compliance with.the i

walysis of the red drop accident la the FSAR. The overtravel positiae feature provides the only positive means of detenrining that a red is properly coupled and therefore this check must be perforund prior to achieving criticality after -

completing CORE ALTERATIONS that could have afPected the control red drive coupling.sntegrity. The subsequent check is perfonned as a backup to the l

initial demonstration. -

In order to ensure that the control ved patterns can be followed and i

therefore that other parameters are withis their limits, the control red position indication system sust be OPERABLE.

l -

The control rod housing support restricts the estuarif sovement of a l

control rod to less than 3.E5 inches is the event of a housing failurs. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdraal increment and .-i, iswill notact costribute required when to any damage to the primary coolant systas. The ei j

there is no pressure to act as a driving forts to rapidly eject a drive

housing. ,

The required surveillance intervals are adequate to determine that the .

I rnds are OPERABLE and not so frequent.as to cause escassive usar en the system components. .

~

3/4.1.4 CONTROL ROD PROGRAM Ginikult '

j control rod withdrawal and inserties sequences are estabTished to assure l

. that the maximun insequence individual centrol rod or control rod segments

' which are withdrahn at any time during the feel cycle could not be worth enoug'h

• to result in a peak fuel enthalpy greater than 280 cal /gs ta the event of a j

control rod drop accident. Om specified sequences are characterized by haugeneous, scattered patterns of control rod'witi:dramat. When THERMAL PolC1t l

is greater than 105 of RATED THERMAL PERER, there is as possible rod morth

which, if dropped at the design rate of the velocity liatter, could result. la a peak enthalpy of 280 cal /ga. 1 bus requiring the Rlfi to be OPERABLE when ~ '

THERMAL POWER is less than er equal to 105 of RATED 1HERMAL poler provides adequate control.

l i

5 j The Rlet provide automatic supervision to assure that eut-of-sequence rods

will not be withdream er insertad. g/

i The analysis of the rod drop accident is presented in Section 15.4 of *

--d the FSAR and the techniques of the analysis are presented 5 : + ;y%w, l -

.. . :, = =; ' --" . Y z *feufur

" % w-- Np - vo-tr,dd /s 191y Mw Fedes.u i

Neeleer tidL../:&:/%g for pecn m

• m Volw I b{r ad A$ y e.ts/a,,dDit%d /9s'3 l .

'~ --

'"W Acen6nt No. 74 1

y _ . _ _ . _ _ _ _ . . _ _ _ _ .. _ . _ _ _ _ _ .._ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _

j . .

POWER DISTRIBUTION .SYSTDt3 l BASES l l

3 /4.2.4 tfNEAR MEAT RENERATION RATE i & F h/ -

l Thi specification assures that the LINEAR HEAT GENERATION RATE (LNGR) in any rod is less than the desion linear heat omaan+ia- ==a if fuel pellet i densification is santulatad e rs pen ys fios nas on Ana a ysis ont n i .2.1 the top al rep 073 i

(5 etwo ese core , s a 11 arly as var tion axia aps I on top as a9 confi ce no re t one caeds e de gn LI GEN ON du r

- piki . , a Referenced f4JN MI *

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

Analysis in Accordance with 10 CFR 50, Appendix K, NED0-20566A',

September 1986.

l

2. ' Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," General Electric Co. Licensing Topical I Report NED0 24154 Vols. I and II and NEDE-24154 Vol. ILI as sup- I'

, plemented by letter dated September 5, 1980, from R. H. Buchholz (GE) to P. 5. Check (NRC). ,

3. "Lasalle County Station Units 1 and 2 SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis,' General Electric Co. Report NEDC-32258P, l October 1993. g

4. " General Electric Standard Application for Reactor Fuel,'

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

I

5. ' Extended operating Domain-and Equipment out-of-service for LaSalle County Nuclear Station Units 1 and 2 " NEDC-31455, November 1987.

,, 1. .

6. ' ARTS Improvement Program Analysis for LaSalle County Station Units 1 and 2," General Electric Co. Report NEDC-31531P, December 1993, h*SYY5'Ihere]

Tk [Pfeeft

• F S'*/ d*' *Tf ' "'* "' W

g, g,g,.af g/ecfrie Shd"l Nff bN*' S E**Y g,j (gggyg),]yEDE-290//~/4. gge A, entwe L#6A reme-r Y$e de/o-SSMA 1We YW'"*

ges,,Qlin*Ib*

Tys LA SALLE - UNIT 2 8 3/4 2-6 Amendment No. 88 l

l

l l Insert #6 J

! SPC Fuel j 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 j operational occurrences (AOOs). Operation above the LHGR limit i followed by the occurrence of an AOO could potentially result in fuel i damage and subsequent release of radioactive material. Sustained l operation in excess of the LHGR limit could also result in exceeding j the fuel design limits. The failure mechanism prevented by the l 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 l pellet. One percent plastic strain of the fuel cladding has .been ,

l defined as the limit below which fuel damage caused by i overstraining of the fuel cladding is not expected to occur. Fuel ,

j design evaluations are performed to demonstrate that the  !

! mechanical design limits are not exceeded during continuous l l operation with LHGRs up to the limit defined in the CORE i 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 j during limiting transients. At reduced power and flow conditions,

the LHGR limit is reduced (multiplied) using the smaller of either the ;

j flow-dependent LHGR factor (LHGRFAC,) or the power-dependent j LHGR factor (LHGRFAC,) corresponding to the existing core flow j and power. The LHGRFAC, multipliers are used to protect the core during slow flow runout transients. The LHGRFAC, multipliers are i 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 l

1 i

j ..

l 1

1

..-.- .-- - - - -. .. . - . - . . - - . - - . . - - . - - . . ~

o

)

4 Insert #7

1. Advsnced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR ECCS Evaluation Model, ANF 91048(P)(A), Advanced Nuc! car 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 I and Supplements I and 2, Exxon Nuclear Company, March 1983.

l

3. Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology l

Summary Description, XN-NF-80-19 (P)(A), Volume 3 Resision 2, Exxon Nuclear Company, 1 January 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.

1

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

1

6. XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, XN-NF  !

105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2. Exxon Nuclear Company, February 1987.

7. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89 98(P)(A) Revision 1, and Resision i Supplement 1, Advanced Nuclear Fuels Corporation, May 1995.

8. LaSalle County Station Units I and 2 SAFER /GESTR - LOCA Loss-of Coolant Accident Analysis.

- NEDC 32258P, General E!cetric Company, October 1993.

i

9. ARTS Improvement Program analysis for LaSalle County Station Units I and 2, NEDC-31531P, i General Electric Company, December 1993. '

1 4

i i

,e A

I i

. . . _ _. . _ .