""fa':DECEMGER 1978

SUPPLEMENTAL RELOAD LICENSINGSUBMITTAL FOR COOPER NUCLEAR

STATION UNIT 1 RELOAD 4

,

79020963 C 6 GENER AL h ELECTRIC

NED0-2417078NED402Class I

Decer::ber 1978

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL

FOR

COOPER NUCLEAR STATION

UNIT 1 RELOAD 4

?'Prepared'

J. L. Rash

V|'

M'"7.Y' ' ?Approved: CdR. E. Engel, ManagerOperating Licenses I

NUCLEAR ENE RGY PROJECTS DIVISION * GENER AL ELECTRIC COMPANYSAN JOSE, CALIFORNI A 93125

G EN ER AL h ELECTRIC

NED0-24170

IMPORTANT NOTICE REGARDING

CONTENTS OF THIS REPORT

PLEASE READ :AREFULLY

This report was prepared by General Electric solely for Nebraska

Public Pouer District (NPPD) for NPPD's use with the U.S. NuclearRegulatory Comission (USNRC) for amending NPPD's operating license

of the Cooper Nuclear Station. The information contained in thisreport is believed by General Electric to be an accurate and truerepresentation of the facts knoun, obtained or provided to GeneralTlectric at the time this report uas prepared.

The only undertakings of the General Electric Company respecting

information in this document are contained in the contract betweenNebraska Public Pouer District and General Electric Company for

nuclear fuel and related services for the nuclear eystem forCooper Nuclear Station and nothing contained in this documentshall be construed as changing said contract. The use of thisinformation except as defined by said contract, or for any pur-pose other than that for which it is intended, is not authorized;and with respect to any such unauthori::ed use, neither GeneralElectric Company nor any of the contributors to this documentmakes any representation or carranty (express or implied) asto the completeness, accuracy or usefulness of the informationcontained in this documcnt or that such use of such informationmay not infringe privately ouned rights; nor do they assume anyresponsibility for liability or damage of any kind uhich mayresult from such use of such information.

NEDO-24170

1. PLANT-UNIQUE ITEMS (1.0)*

Bundle Loading Error Analysis - Appendix A

GETAB Analysis Initial Conditions - Appendix B

Densification Power Spiking - Appendix C

2. RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0)

Fuel Type Number Number Drilled

Irradiated Initial Core 7DB250 152 152

Irradiated 8DB250 (R1) 72 72Irradiated 8DB274L (R162) 112 112Irradiated 8DB274L (R3) 24 24Irradiated 8DRB283 (R3) 76 76

New 8DRB283 (R4) 112 112

Total 548 548.

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure: 15,753 mwd /tAssumed reload cycle exposure: 16,230 mwd /tCore loading pattern: Figure 1

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

BOC k gg

Uncontrolled 1.110Fully Controlled 0.952Strongest Control Rod Out 0.986

R, Maximum Increase in Code Core Reactivitywith Exposure Into Cycle, ak 0.000

*( ) refers to areas of discussion in " Generic Reload Fuel Application,"NEDE-240ll-P-A, Revision 0, August 1978,

1

NT90-24170

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (Ak)1331 (20'C, Xenc, Free)

600 0.05o

6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)

EOC5

Void Coefficient N/A* (c/% Rg) -8.15/-10.19Void Fraction (%) 40.03Doppler Coefficient N/A (c/%'F) -0.226/-0.217Average Fuel Temperature (*F) 1352

Scram Worth N/A ($) -38.78/-31.02Scram Reactivity vs Time Figure 2

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INI11AL CONDITION PARAMETERS (5.2)

7x7 8x8/8x8RExposure EOC5 EOC5

Peaking factors (local, (1.24, 1,378, 1.4) (1.19, 1.578, 1.4)

radial and axial)R-Factor 1.08 1.054

Bundle Power (MWt) 5.864 6.703

Bundle Flow 119.85 109.86(103 lb/hr)Initial MCPR 1.20 1.22

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

None

*N = Nuclear Input DataA = Used in Transient Analysis

2

NEDO-24170

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

A

Core $ Q/A ACPRp pPower Flow (% (% sl v 8x8/ Plant

Transient Exposure _ _ %) (%) NBR) NBR) (psig) (psig) 7x7 8x8R Response(

Turbine BOC-EOCS 104 100 226 107 1170 1194 0.09 0.14 Figure 3TripWithoutBypass

Load BOC-EOC5 104 100 248 108 1172 1196 0.10 0.15 Figure 4RejectionWithoutBypass

Loss of BOC-EOC5 104 100 124 117 1023 1069 0.13 0.14 Figure 5100'FFeedwaterHeating

Feedwater BOC-EOC5 104 100 168 110 1123 1165 0.09 0.13 Figure 6ControllerFailure

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

SUMMARY (5.2.1)

RodPosition

Rod Block (Feet ACPR MLHGR (Kw/ft) LimitingReading Withdrawn) 7x7 8x8/8x8R [x7 8x8/8x8R Rod Pattern

105* 3.5 0.16 0.10 17.3 16.1 Figurc 7106 4.0 0.23 0.12 17.5 16.4 Figure 7107 4.5 0.31 0.13 17.5 16.4 Figure 7108 5.0 0.33 0.14 17.3 16.2 Figure 7109 5.5 0.33 0.15 17.2 16.2 Figure 7110 7.0 0.32 0.19 16.7 15.5 Figure 7

* Indicates setpoint selected

3

NED0-24170

11. OPERATING MCPR LIMIT (5.2)

BOC to EOC5

1.23 (8x8/8x8R fuel)1.23 (7x7 fuel)

12. OVERPRESSURIZATION ANALYSIS SLWiARY (5.3)

Power Core Flow sl v PlantTransient (%) (%) (psig) (psig) Response

pf gsu e

104 100 1237 1276 Figure 8,

13. STABILITY ANALYSIS RESULTS (5.4.)

Decay Ratio: Figure 9

Reactor Core Stability:

Decay Ratio, x /*02(105% Rod Line - NaturalCirculation Power) 0.79

Channel Hydrodynamic Performance Decay Ratio, x /*02(105% Rod Line - NaturalCirculation Power)

8x8/8x8R channel 0.37

7x7 channel 0.23

14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)

No new bundle types are added; no ECCS Tech Spec changes will be

necessitated by this reload.

4

NED0-24170

15. LOADING ERROR RESULTS (5.5.4)

Limiting Event: Rotated Bundle (SDRB283)

MCPR: 1.07

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Doppler Reactivity Coefficient: Figure 10

Accident Reactivity Shape Functions: Figures 11 and 12Scram Reactivity Functions: Figures 13 and 14Plant specific analysis results

Parameter not bounded: Accident ReactivityShape Function,Cold Startup

Resultant peak enthalpy: 214 cal /gm

5

NED0-24170

DIED El.00 8 8 ODiB ED 0s2 Y8 00'8 0; 3D B B 8'ED 8 ED 8so@@ @@ 0D 8 8@ 8 8 ED 8 00:00 ED ED @@a

E0 8 8 00 8 0 8 8 8 E8 8 8 E0 8'8 8 00 8 8 EDa0 8 8 00 8 8 800 8 8 ED 8 8iB ED 8 0 8 8 8 8 0a Y YED B 59 @@ 8 ED @@ B 8 8'8 8 8 8 8 8 OD ED 8428 0 8@ 8 @ SiGD 8_@ 8_ 8 @ 8 B i B B B @@ 55a

ED E+D ED 8ED o ED 8'ED 00 8 8 8 0 0 8'UD 8 ED ED 0 8 ED ED>=as -O GD DiGD BiED EiDD SiED 8 9 BiGD 85 ED 8 8 ED E_D5 5 5 00 0Y Y YY24 -ED B 818 00's BD'ED 8'ED ED ED ED'ED EDED ED E B ED 8 00 B ED ED B22 -EDiB 8_@ 8i00 8 8 85 8,8 Bi8 8 8 bib OD 8 8 8 58 8 EDYso -00'ED 8 8 8'UD ED ED ED 8 8'ED 8'8 8 8 8 OD ED ED 8 ED 8 ED ED ED28 -BiED BiED ED 8 8 GD ED 8 8 8 8 8 8 8 8 ED ED:8 8 ED 8 ED 0+0 8YYYY Y Y28 -00'00 8'8 B B S ED ED ED B ED E ED ED 8 ED 00 8'8 @@ ED 8 C 0D24 -ED 8 58 @ @ bib UD ED @@ 8_@ $@ $5 SiED SS B E0 8 8Y Y YY22 -ED B B B B B 8'ED B B B B B B B B B B E]'8 OD 8 ED O B ED2o -EDiED 8 8 8 ED BiED @i8 ED E 8 8 8 ED 8 8 EDi8 8 ED 8 ED ED EDYY YY Y Yis -O'8 B O B B 8800 8'8 B ED T B @@ B B 8'8 B 8 5@ 00 C

BD EDiO 8_ ED 8 ED 095 5@ S@ bib B B ED 8 8 00 BiED EDis Y Y| |; EDiED 88 EDiED Op888 BpD E0 8 8 E0

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FUEL TYPE

A = NOT USED E * RELOAD 1 AND 2 8082748 INITIAL CORE 7D8250 F * RELOAD 3 808274C = INITIAL CORE 708250 G = RELOAD 3 80RB283D = RELOAD 1808250 H = RELOAD 4 80R B283

Figure 1. Reference Core Loading Pattern

6

NED0-24170

100 4r.,

9040

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35

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30

60

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40

15

30

1020 NOMINAL SCRAM CURVE IN (-5)

5

10

SCRAM CURVE USED IN ANALYSIS

0 00 1 2 3 4

TIME (sec)

Figure 2. Scrara Reactivi y and Control Rod Drive Specificationst

7

NED0-24170

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Figure 4. Plant Response to Generator Load Rejection, without Bypass

[ t EUTFDJ FLUX 1 VESSEL PfES RISE (PSI)2 HVE SUHFfCE T RT FLUX 2 K LIEF VFLVE FL N/ 3 COME IM_E T FLOW 3 BYPRSS VFLVE FLOW150* ID*Q'CNE IfUT SUB---~ 45 5

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2. I-0. I40. 80.-

* *

TIME ISEC1 TIIE (SEC)

Figure 5. Plant Response to Loss of 100 F Feedwater IIeating, MFC

I E11THtN FLIDC 1 VESSEL PtES HISE (PSI)

150* u--} 2 HVE StiffCE iffli Ft1DC 2 SAFETT VFLVE FLfM

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{" ' ?''0.

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--

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Figure 6. Plant Response to Feedwater Failure, Maximum Demand, with liigh LevelTurbine Trip

NEDO-24170

02 06 10 14 18 22 26 30

51

47 10 10

43 30 18

39 6 10 6

35 30 40 42

31 14 0 18

27 40 42 42

23

NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC UPPER LEFT QUADR ANT SHOWNON MAP.

2. NUMBERS INDICATE NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS AWITHDRAWN ROD.

3. EhROR ROD iS 118.31).

Figure 7. Limiting RWE Rod Pattern

12

1

1 EUTTUI FLLIX 1 WSEL PFES RIE (PSI)2 RVE Sl WFCE ERT FLUX 2 SEETY VfLVE FL N

150. 3 COE ItuT FL m 3 ELIEF VCLVE Fim300.g q-~BNFL~NS S

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Figure 8. Plant Response to MSIV Closure, Flux Scram

NEDO-24170

12

ULTIMATE STABILITY LIMIT_ ,, _

0.8

^OX

_NO NATURAL

Q O.6 CIRCULATIONe>$wO 105% ROD LINE

0.4

02

00 20 40 60 80 100

PE RCE NT POWE R

Figure 9. Decay Ratio

14

NEDO-24170

0

IN 1/ DELTA degrees C (MILLIONTHS)

-5

-10

/

$w - 152i8O

$$o ~200

O SOUNDING VALUE FOR 280 cal /g COLD

A BOUNDING VALUE FOR 280 cal /g HS8

- 25 O CALCULATED VALUE - COLD

h CALCULATED VALUE - HSB

-30

V

i 10 400 800 1200 1600 2000 2400

FUEL TEMPERATURE (OC)

Figure 10. Fuel Doppler Reactivity Coefficient Comparison for RDA

15

NED0-24170

20

16

Gh12 - -S5,

d OI-

N>-b28 -U3m

i

O BOUNDING VALUE FOR 280 cal /g

O CALCULATED VALUE

4 -

i l I lg0 4 8 12 16 20

ROD POSITION (ft OUT)

Figure 11. RDA Reactivity Shape Function at 20*C

16

NED0-24170

20

16

G[ BOUNDING VALUE FOR 280 cal /g

S12 O

$OE_

v''.t5h8<E

CALCULATED VALUE3__ _

4

0[_0 4 8 12 16 20

ROD POSITION (ft OUT)

Figure 12. RDA Reactivity Shape Function at 286*C

17

NEDO-24170

70

60 -

s0

G?OZ

$ 4085

N'.

5mPNa

20 -

CALCULATE D yALUE

BOUNDING VALUE FOR 280 cal /g

10

OC ^ k0 4 8

ELAPSED TIME (sec)

Figure 13. RDA Scram Reactivity Function at 20*C

18

NED0-24170

100

80

5s

!=2

NI

i22G#

CALCULATED V ALUE

BOUNDING VALUE FOR 280 cal /g

20

0(2 2 20 2 4 6 8 10

ELAPSED TIME bec)

Figure 14. RDA Scram Reactivity Function at 286*C

19/20

NEDO-24170

APPENDIX A

BUNDLE LOADING ACCIDENT

The loading error accidents nislocated bundle and rotated bundle) for Cooper

Cycle 5 have been analyzed using the " revised methods" (Reference A-1). These

analysis methods have been previously applied to Cooper (References A-2 andA-3) and these applications approved by the NRC (References A-4 and A-5) .

The loading errors will not influence the opera t: ng CPR limit for 7x7 fuels;however the analysis of a rotated 8DRB281 bundle indicates more severe conse-quences than any abnormal operational transient. The operating CPR limit for

8x8 and 8x8R bundles will therefore be established by the rotated bundleaccident. This limit includes a bias of 0.02 for R factor uncertainties asrequired by the NRC (Reference A-1). The most severe MLHGR predicted for a

misloaded bundle is 16.4 kW/ft. This value includes an allowance of 2.2% forpower spiking due to fuel densification (see Appendix C) .

REFERENCES

A-1 Letter, D.C. Eisenhut (NRC) to R.E. Engel (GE), transmitting Safety Eval-uation Report on "new calculational procedures. .for the fuel bundle.loading error analyses", May 8, 1978.

A-2 Letter, J. Pilant (NPPD) to George Lear (NRC), April 14, 1978.

A-3 Letter, J. Pilant (NPPD) to T.A. Ippolito (NRC), August 16, 1978.

A-4 Letter, George Lear (NRC) to J.M. Pilant (NPP D) , May 2, 1978.

A-5 Letter, T.A. Ippolito (NRC) to J. Pilant (NPPD) , Augus t 25, 19/6.

A-1/A-2

NFDO-24170

APPENDIX B

GETAB INITIAL CONDITIONS

Table 5-8 of Reference B-1 states the "Nonvarying Plant GETAB Analysis InitialConditions". The Cooper core pressure is given as 1045 psia. A value of1035 psia, which more nearly reflects actual plant data, was assumed for thiss ub mit tal .

Reference B-1 wi?1 be revised to eliminate this discrepancy.

REFERENCES

B-1 Licensing Topical Report, " General Electric Boiling Water Reactor, GenericReload Fuel Application", NEDE-24011-P-A, May 1977.

B-1/B-2

NED0-24170

APPENDIX C

DENSIFICATION POWER SPIKING

Reference C-1 documents the NRC staff position that ". .it (is) acceptable to.remove the 8x8 and 8x8R spiking penalty factor from the plant Technical Specifi-cation for those operating BWR's for which it can be shown that the predictedworst case maximum transient LHGR's, when augmented by the power spike penalty,do not violate the exposure-dependent safety limit LHCR's".

The Cooper Reload-4 submittal contains the required information to remove thepower spiking penalty from the Cooper Technical Specifications. Section 10,Rod Withdrawal Erro , and Appendix A (Bundle Loading Accident) include the den-

sification effect in the calculated LHGR of the 8x8 fuels.

REFERENCES

C-1 " Safety Evaluation of the General Electric Methods for the Considerationof Power Spiking Due to Densification Ef fects in BWR 8x8 Fuel Design andPerformance", Reactor Safety Branch, DOR, May 1978.

C-l/C-2

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