ML20198B931

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Rev 0 to Supplemental Reload Licensing Submittal for LaSalle County Station,Unit 1,Reload 1 (Cycle 2)
ML20198B931
Person / Time
Site: LaSalle Constellation icon.png
Issue date: 06/30/1985
From: Charnley J, Elliott P, Verbryke P
GENERAL ELECTRIC CO.
To:
Shared Package
ML19344C148 List:
References
23A1843, 23A1843-R, 23A1843-R00, NUDOCS 8511070370
Download: ML20198B931 (28)
v • d • e


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A v .c y.~ LICENSING SUBMITTAL 3..FORM

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23A1843 1 Revision 0 Class I June 1985 r

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR LA SALLE COUNTY STATION UNIT 1, RELOAD 1 (CYCLE 2)

Frepared: .

bc P. E. Elliott Verified: b I b/7 5

  • P. A. bryke

.. m , y

/ '

/  !

Approved. C ,

J 'S dffdrnley, Man 'e r Licensing ,

i NUCLEAR ENCAGY BUSINESS OPERATIONS GENERAL ELECTRIC COMPANY ,

SAN JOSE, CALIFORNIA 95125 l

GENERAL $ ELECTRIC 1/2 j l

23A1843 Rev. 0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General- Electric solely for Common' wealth Edison Company (CECO) for Ceco's use with the United States Nuclear

, Regulatory Commission (USNRC) for amending CECO's operating license of the laSalle County Station, Unit 1. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting infor-mation in this document are contained in the contract between Commonwealth Edison Company and General Electric Company for Nuclear fuel and related ,

services for the nuclear system for LaSalle 1 and nothing contained in this '

document shn11 be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that-for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty' (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which nay result from such use of such information. .

l l

3/4 l s di P!

r 23A1843 Rev. 0

1. PLANT UNIQUE ITEMS (1.0)*

' Transient Analysis Basis: Appendix A Control Rod Drop Analysis: Appendix B Plant Parameters: Appendix C

2. RELOAD FUEL' BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

. Fuel Type Cycle Loaded Number Number Drilled Irradiated 8 CRB176 *

  • 1 100 100 8CRB219 ** 1 144 144 8 CRB219 ** 1 288 288 New:

BP8CRB299L 2 232 232 Total 764 764

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 10,504 mwd /ST

!!inimum previous cycle core average exposure at end _,

of cycle from cold shutdown considerations: 10,303 mwd /ST Assumed reload cycle core average exposure at end of cycle: 14,654 mwd /ST Core loading pattern: Figure 1

( ) . Refers to area of discussion in " General Electric Standard Application for Reactor Fuel", NEDE-240ll-P-A-6, dated April 1983. A letter "S" preced-ing the number refers to the appropriate country-specific supplement.

    • Included in " Final Safet'y Analysis Report for LaSalle County Station,"

dated January 1981.

5

.s

23A1843 Rev. 0

4. , CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)

Beginning of Cycle, k,f f Uncontrolled 1.121.

Fully Controlled 0.952 Strongest Centrol Rod Out 0.981

  • - R, Maximum Increase in Cold Core Reactivity 0.004 with Exposure into Cycle, ok
5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (4k)

(20*C, Xenon Free)

J82E 660 0.038

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)

Values normally reported for this section are REDY. inputs. There were no transients analyzed using REDY.

l I

6

23A1843 Rev. 0

7. REIDAD UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)

Exposure: BOC 2 to EOC 2 Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Fac tor (MWt) (1000 lb/hr) MCPR BP8x8K' 1.20- 1.47 1.40 1.051 6.274 116.3 1.29 8x8R 1.20 1.50 1.40 1.051 6.414 115.3 1.26

8. SELECTED MARCIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: Yes Rod Withdrawal Limiter: No Thermal Power Monitor: Yes Improved Scram Time: No Exposure Dependent Limits: No Exposure Points Analyzed: 1 .

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3 )

Single-Loop Operation: Yes Load Line Limit: No _,

Extended Load Line Limit: No Increased Core Flow: No Flow Point Analyzed: *N/A Feedwater Temperature Reduction: No 7

I

23A1843 Rev. 0

10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Exposure: BOC 2 to EOC 2 ACPR Flux Q/A Transient (% NBR) (% NBR) BP8x8R 8x8R Figur e Load Rejection Without Bypass 474 122 0.22 0.20 2

. Loss of !eedwater-Heater 125 125 0.09 0.09

11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(S.2.2.1)

Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: No ACPR LHGR (kW/ft)

Rod Block Rod Position Reading (feet withdrawn) BP8x8R 8x8R 8x8R/BP8x8R 104 3.5 0.11 0.11 16.67' 105 4.0 0.14 0.14 16.93 106 4.5 0.17 0.17 16.93 .

107 5.0 0.19 0.19 16.93

~"

108 5.5 0.21 0.21 16.93 '

109 6.0 0.22 0.22 16.93  !

110 7.5 0.26 0.26 16.93

{

Setpoint Selected: 107 I

  • See Appendix A 8

s .

23A1843 Rev. 0

12. CYCLE HCPR VALUES (S.2.2)

Non-Pressurization Events Exposure Range: BOC 2 to EOC 2 '

BP8x8R 8x8R Loss of Feedwater Heater 1.16 1.16 Fuel Loading Error *

  • Rod Withdrawal Error 1.26 1.26 Pressurization Events Exposure Range: BCC 2 to EOC 2 Option A Option B BP8x8R 8x8R- BP8x8R 8x8R Load Rejection Without Bypass 1.35 1.32 1.25 1.23 Feedwater Controller Failure 1.29 1.28 1.26 1.24

- 13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) s1 v Transient __psig)

( (psig) Plant Response MSIV Closure 1234 1269 Figure 4 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed:

  • 105%

Decay Ratio: " Figure 6 Reactor Core Stability Decay Ratio, x2 /*0: 0.60 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 Channel Type BP8x8R/8x8R 0.47

  • Fuel loading error not applicable for C-Lattice plants.

9

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23A1843 Rev. O t.

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15. ' LOADING ERROR RES1'LTS (S.2.5.4)

I Not applicable for C-Lattice plants.

l I

f 16. - CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1) .

L

. 'See Appendix B.

17. LO,SS-OF-COOLANT ACCIDENT RESULTS (S.2.5.2)

Fuel Type: PP8CRB299L Exposure MAPLHGR PCT Local 0xidation (kW/ft) (*F) (Fraction)

.CWD/MT) 0.22 10.80 1828 0.007 l 1.1 11.00 1831 0.007 -

5,. 5 11.80 1880 0.008 11.0 12.30 1909 0.009 17.0 12.40 1932 0.009 '

22.0 12.30 1945 0.009 28.0 11.80 1897 0.008 39.0 10.70 1768 0.005 49.0 9.20 1595 0.002 -*

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23A1843 Rev. 0-

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FUEL TYPE-A = 8CRB176 C = 8CRB219 B = 8CRB219 D = BP8CRB'z 99L Figure 1. Reference Core leading Pattern 11

23A1843 Rev. 0 1 NEUTRON FLt.HC 1 VESSEL PRES'S RISECPSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE Flow 3 CORE IPLET TLOV 3 RELIEF VALVE FLOV

'150.0 380.0

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Figure '2. - Plant Response to Generator Ioad Rejection Without Bypass 12

23A1843 Rev. 0 -

150.0 1 NEddTRON FLU 1 VESSEL PRE RISE (PSI) 2 AV 3 SUR. ACE HEAT FLUX 2 SAFETY VAL FLOV 3 CC $ If T LOW 3 RELIEF VALVE FLOW 13;. ; *C" E 'NL ? '* 4 BYPASS VALVI FLOV a

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TIME (SECONDS) TIME (SECONDS) 1 LEVEL (INCH-REF-SEP-SKRT) i VOID REACT!9 TTY-2 VESSEL STEA1FLOV 2 DCPPLER REAUTIVITY 3 TUR9IhE STE AMFLOV 3, S,C,p.H , erREA.CTjIV,ITY c

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Figure 3. Plant Response to Feedwater Controller Failure 13

i 23A1843 Rev. O i

~

I 1 Nhi GN kUX 1 VESSEL PRESS RISE : PSI) 2 AVE URFA E HEAT l' LUX 2 SAFETY VALVE FLOW 13 3 ; 3 C0f INLE FLOW 3 RELIEF VALVE FLOW 320.0 . e_v. o i c_ e. " e ' r e e: r_ 2 100.0 - - O 2 200.0 7 1 tz  %

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e TUR8IN.E STE,AMFLOW 200.0 c. e_ r_ n_m_.

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Figure 4. Plant Response to MSIV Closure, Flux Scram 14

23A1843 Rev. 0 -

i i

NOTES: l'.

-Rod pattern is 1/4-core mirror symnetric.

2. No. indicates number. of notches withdrawn out of 48.

Blank is a withdrawn rod..

3. -Error rod is (30,43). ,

4 2 6 10 14 18 22 26 30 39 6 6 55 -40 36 ,

51 6 6 8 47' 36- 44 43 2 6 8 0

-39 32 40 36- 44 ~"

35 2. ,6 6 8 31 32 32 40 36 I 1

i i

Figure 5. Limiting RWE Rod Pattern 15

i-23A1843 Re v . O Ab ATURAL C RCULATIO N 81 05 PERCENT ROD L1 1E CL LTIMATE $TABILITY LINE 1.00 C C

.75 N

A b

" .50 '

7 U

LtJ Q

.25 0.00 "

0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 6. Reactor Core Decay Ratio versus Power 16

23A1843 Rev. 0 APPENDIX A TRANSIENT ANALYSIS BASIS The Loss of Feedwater Heating event was analyzed using the BWR Simulator Code (keference A-1). The us'e of this code is permitted in GESTAR II

. (Reference A-2).

The Loss of Feedwater Heating Event plot normally reported in bection 10 is not an output 'of- the BWR Simulator Code. Therefore, this plot is not

- included in this document. I l

REFERENCES A-1. "Three Dimensional BWR Core Simulator," NED0-20953A, January 1977.

. A-2.- " General. Electric Standard Application for Reactor Fuel,"

t

' {. NEDE-240ll-P-A-6, dated April 1983.

,1 l,

o y.

1 4

.E 17/18

23A1843 Rev. 0 APPENDIX B CONTROL ROD DROP ANALYSIS The cycle-specific control rod drop accident analysis has been discon-tinued for banked position withdrawal sequence (BPWS) plants based on the fact that in all cases the peak fuel enthalpy from a control rod drop acci-dent would be much less than the 280 cal /gm limit. This change in procedures

. was reported and justified in Reference B-1.. Reference B-2 indicates this change is acceptable to the NRC.

REFERENCES l B-1. Letter, R. E. Engel (GE) to D. G. Vassallo (NRC), " Control Rod Drop Accident," February 24, 1982.-

B-2. NRC Memo, L. S. Rubenstein to G. C. Lainas, " Changes in GE Analysis of the Control Rod Drop Accident for Plant Reloads (TACS-48058),"

February 15, 1983.

4 l.

d' e

19/20

23A1843 R:v. O APPENDIX C PLANT PARAMETERS These values differ from the values reported in NED0-240ll-P-6-US.

l l

Dome Pressure 1029 psig l Turbine Pressure 980 psig Rated Steamflow 6 14.90 x 10 lb/hr Dual Mode Safety / Relief Valves 18 Relief Mode Low Setpoint 1087 psig Safety Mode Low Setpoint 1162 psig i

f

  • 21/22 (FINAL)

4 ADDITIONAL INFORMATION REGARDING THE SUPPLEMENTAL RELOAD LICENSING SUENITTAL FOR IaSalle 1 REIAAD 1 / CYCLE 2 (Page 1 of 2]

Section-3. REFERENCE CORE LOADING PATTERN Cycle N-1 Incremental Exposure 10,504 mwd /t

_ Cycle N Exposure Increment 6,791 mwd /t Cycle N Full Power Capability (if different from above) 6,791 mwd / t i Section 4. CALCUIATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH (NO VOIDS, 20*C)

Cycle Incremental Exposure Corresponding to Minimum Shutdown Margin R-Value 6,000- mwd / t

. Section 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (REDY EVENTS) i 3 EOC Void Fraction (Haling)

EOC Bypass Flow Fraction e

Delayed Neutron Fraction (BOC/EOC) > No REDY Analyses were performed .

Void Coefficient x 103 (B0C/EOC) i-

[ Units are (&/k)/( A% voids)] ,

i- --

Section 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS P,1l

~

P yl P vl Limiting Power Flow Flux. Q/A (Ede) (Bottom)

Exposure Transient (% NBR) (% NBR) (% NBR) (% NBR) (psig) (psig) (psig)

EOC TTNBP 103.94- 100.0 393.7 120.4 1170 1168 1196 EOC IRNBP 103.94 100.0 474.3 122.2 1170 1169 1198 EOC MSIVF 103.94 100.0 662.3 127.1 1234 1238 1269 Not Analyzed 'MSIVD BOC LFWH 103.94 100.0 124.7 124.7

  • 1026
  • EOC FWCF. 103.94 100.0 312.4 119.9 1143 1142 1174
  • Numbers not available. PANACEA analysis was used.

ADDITIONAL INFORMATION REGARDING THE SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR IaSalle 1 RELOAD 1 / CYCLE 2 [Page 2 of 2]-

Section 10.- CORE-WIDE TRANSIENT ANALYSIS RESULTS (continued) )

Were resolved OPL-3 values used for safety and relief valve characteristics? Yes Assumed MSIV Closure Characteristics:

Time (sec) MSIV Area (per unit) 0.0 1.0 (fully opened) 0.6 1.0 1.7 0.01 3.0 0.0 (fully closed) As used.

Section 14. ROD BLOCK LINE EQUATION RB < 0.58WD + 50 Not applicable since no LLLA performed.

Section 15. LOADING ERROR RESULTS "

Bundle Type for Limiting Misorientation Not Analyzed 1

9

r- _ . _ _ . - . _ _ _ _ _ _ _ - - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___

GEN ER AL O iticTaic NUCt. EAR ENERGY BuslNESs OPULATIONs GENERAL ELECTRIC COMPANY

  • 175 (URTNER AVENUE o SAN Jose, CAUFORNIA 9s195 Mail code 174 June 26, 1985 cc: H. E. Bliss REP:85-115 P. R. Mattson Mr. J. L. Anderson Fuel Buyer

. COPt10NWEALTH EDIS0N COMPANY Fuel Department, 234 E -

P. O. Box 767 Chicago, IL 60690

SUBJECT:

Responses to Edison Questions on the LaSalle Unit 1 Reload 1 Licensing Submittal

REFERENCE:

Letter, J. L. Anderson to R. E. Parr, "LaSalle County Unit 1 Cycle 2 Draft Reload Licensing Submittal",

JLA:85-176, June 11, 1985.

Dear Mr. Anderson:

The following are responses to the questions submitted to GE by Edison (Reference) concerning the LaSalle Unit 1 Reload 1 Licensing Submittal:

1. %y is fuct type BCRB219 divided into 2 groups?

E An artificial division of the 8CRB219 bundles into two groups

! has been maintained in Cycle 2 to be consistent with the initial core

data books. The bundles are all the same.

l

2. Why is LRGR constant with rod motion?

A plant, cycle-specific RWE was analyzed. The LHGR is not constant with rod motion. What is shown is the maximum LHGR that occurs anywhere between the beginning of the event (rod full in) and the current rod position.

3. Confirm that there are no NRC reporting requirements for the Puel Loading Error (Mislocation or Misorientation).

From NEDE-24011-P-A-6-US Licensing Topical Report Section S.2.5.4.1: " Analysis of the mislocated bundle accident is perfonned for initial cores only. Mislocated bundle analyses are not performed for reload cores because, based on an analysis of data available from past reloads, the probability that a mislocated fuel bundle loading error will result in a CPR less than the safety limit is sufficiently

- small that plant / cycle-specific analyses are not necessary (Reference S.2-71). Interim approval of this is given in Reference S.2-72.

,' GENERAL $ ELECTRIC Mr. J. L. Anderson June 26, 1985 4 The Option B MCPR values for FWCF are not consistent. Rich values are correct?

Option B MCPR values are consistent with the E0C adjustment factor for FWCF (no MOC analysis was perfonned).

l I

Basic Equation

  • af+SL aV+SL + ADj.

Solving

5. ny are the decay ratios less than the initial cycle?

The core decay ratio decreased from 0.70 to 0.60 for Reload 1 for the following reasons:

(1) The stability analysis methods have changed in the nine l years since the FSAR analysis in the area of inputting the

(

nuclear void coefficient:

a. Method of converting void coefficient inputs to the reactivity coefficients needed by the stability
l. analysis code.

I

( b. Method of allowance for mid-cycle effects.

^'

(2) The thennal-hydraulic inputs to the stability analysis were a little more conservative in the FSAR. This effect also shows up in the channel hydrodynamic decay ratio change from 0.49 in the FSAR to 0.47 in the Reload 1 analysis.

The methods currently in use for stability analysis have not changed in four years, so that many reload analyses have been accepted by the NRC since then. If there is a question by the NRC, the LaSalle Unit 1 Reload I results can be compared with other plant analyses done in recent years.

6. %y doesn't reactor vater level vary ao in the original FSAR analysis?

A REDY analysis was used for the initial core whereas an ODYN analysis was used for Cycle 2 transients.

GENERAL O ELECTRIC Mr. J. L. Anderson June 26, 1985

7. Appendix C valucc for dome pressure and rated steam flou do not match resolved OPL-3 values.

The difference in the dome pressures (1045 psia in OPL-3 and 1044 in ODYN) is within acceptable gesign tolerance. Thedifferenge in the steam flow rates (14.97 x 10 lb/hr in OPL-3 and 14.90 x 10 in ODYN) will have a negligible effect on MCPR; i.e., a.01 aCPR.

Seventy copies of the final Reload Licensing Submittal are included

, with this letter.

Please call me if you have any questions.

Very truly yours, dM &

. E. Parr Senior Fuel Project Manager Edison Projects i

, M/C 174; (408) 925-6526 I rem Encl.

ATTACHMENT E GENER AL $ ELECTRIC NUCLEAR ENEAGY OUsiNESs OPDtADONs CENERAL ELECTRIC COMPAW

  • 175 CURTNER AWNUE o SN4 JOSE, CAUFORNIA 95125 Mall Code 174 May 17, 1985 cc: H. E. Bliss REP:85-094 P. R. Mattson Mr. J. L. Anderson Fuel Buyer COMMONWEALTH EDIS0N COMPANY Fuel Department, 234 E P. O. Box 767  !

Chicago, IL 60690 l l

SUBJECT:

LaSalle Fuel Bundle BP8CRB299L i i

Dear Mr. Anderson:

The subject fuel bundle is not currently included in GESTAR II (NEDE-24011-P-A) and cannot be incorporated until the next amendment to that document, currently expected to be submitted in late June, 1985.

Therefore, in order to license this bundle, Edison should send the attached bundle information to the NRC when they submit the Supplemental Reload License Submittal for review.

I have attached a draft letter that Edison may use as the basis for their letter to the NRC, including the infor1 nation required to license a ,

fuel bundle. Note that some of the information is considered proprietary.

Very.trul yours, i v e

'R. E. Parr l Senior Fuel Project Manager Edison Projects l M/C 174; (408) 925-6526 i

rem .

Attach.

, . es en e. . e s e e.

Attention: LaSalle Project Manager

SUBJECT:

NEW BUNDLE ENRICHMENT i

l

Reference:

J. F. Quirk (GE) to 0. D. Parr (NRC), " General Electric '

Licensing Topical Report NEDE-24011-P-A, Generic Reload Fuel Application," 9/11/78 A fuel bundle designated "BP8CRB299L" is to be loaded into LaSalle County Unit I for use during Cycle 2. This fuel bundle is not currently included in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel,"

although it will be incorporated in the next amendment to that document. That amendment is presently scheduled to be made prior to the start of Cycle 2, but we are submitting the information required to license this fuel bundle in advance. Table 1 contains the type of information included in NEDE-24011-P-A for each fuel bundle, and Figures 1, 2, and 3 provide a physical description of the bundle. This bundle has been analyzed with the approved methods and complies with the approved limits described in NEDE-24011-P-A. Because the BP8x8R fuel bundle design has been approved by the NRC, there are no unreviewed safety questions, and the licensing of a new bundle enrichment is considered a non-safety related change to NEDE-24011-P-A.

The information in Figures 1, 2 and 3 is of the type which the General Electric Company customarily maintains in confidence and withholds from public disclosure.

The information has been handled and classified as proprietary to the General Electric Company as indicated in the affidavit attached to the reference letter. We hereby request that this information be withheld from public disclosure in accordance with the provisions of 10CFR2.790.

cc: General Electric Company _ _ .

l I

l 1

PAV:csc/IO5087-2

Table 1 BP8CRB299L Fuel Bundle Information Weight Weight Exposure at Enrichment of UO of U Maximum Max. K-inf. )

Lattice (wt %) (kg)2 (kg) K-inf. (GWd/st)

BP8CRL319 2.99 206.9 182.3 1.266 7.0 l

I l

l PAV:csc/IO5087-3

ATTACHMENT F Attention: LaSalle Project Manager

SUBJECT:

NEW SUNOLE ENRICHMENT

Reference:

J. F. Quirk (GE) to 0. D. Parr (NRC), " General Electric Licensing Topical Report NEDE-24011-P-A, Ganeric Reload Fuel Application," 9/11/78 A fuel bundle designated "BP8CRB299L" is to be loaded into LaSalle Cour.ty Unit I for use during Cycle 2. This fuel bundle is not currently included in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel,"

although it will be incorporated in the next amendment to that document. That amendment is presently scheduled to be made prior to the start of Cycle 2, but we are submitting the information required to license this fuel bundle in advance. Table 1 contains the type of information included in NEDE-24011-P-A for each fuel bundle, and Figures 1, 2, and 3 provide a physical description of the bundle. This bundle has been analyzed with the approved methods and complies with the approved limits described in NEDE-24011-P-A. Because tne BP8x8R fuel bundle design has been approved by the NRC, there are no unreviewed safety questions, and the licensing of a new bundle enrichment is considered a non-safety related change to NEDE-24011-P-A.

The information in Figures 1, 2 and 3 is of the type which the General Electric Company customarily maintains in confidence and withholds from public disclosure.

The information has been handled and classified as proprietary to the General Electric Company as indicated in the affidavit attached to the reference letter. We hereby request that this information be withheld from public disclosure in accordance with the provisions of 10CFR2.790.

cc: General Electric Company _,

t PAV:csc/105087-2

Table 1 BP8CRB299L Fuel Bundle Information Weight Weight Exposure at Enrichment of UO of U Maximum Max. K-inf.

Lattice (wt %) (ko)2 (ko) K-inf. (GWd/st)

BP8CRL319 2.99 206.9 182.3 1.266 7.0 l

PAV:csc/IO5087-3 ,

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