ML20095G828

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Rev 0 to Supplemental Reload Licensing Submittal for Duane Arnold Atomic Energy Ctr,Unit 1,Reload 7
ML20095G828
Person / Time
Site: Duane Arnold NextEra Energy icon.png
Issue date: 06/30/1984
From: Lambert P
GENERAL ELECTRIC CO.
To:
Shared Package
ML112230638 List:
References
23A1739, 23A1739-R, 23A1739-R00, NUDOCS 8408280211
Download: ML20095G828 (22)
v • d • e


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ETs'si JUNE 1984 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR DUANE ARNOLD ATOMIC ENERGY CENTER UNIT 1, RELOAD 7 rN' SSEoES8Sjj,5 GENER AL $ ELECTRIC y

I 23A1739 Revision 0 Class I June 1984 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR DUANE ARNOLD ATOMIC ENERGY CENTER UNIT 1, RELOAD 7 Prepared an 64 -

'P. A. Lambert Verified: /

W. A. r is Approved: - 7M/Fh J. f. Charnj/ty, kahager Fuel Licensing kUCLEAR ENERGY BUSINESS OPERATIONS e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL $ ELEChRIC 1/2

23A1739 R;v. 0 IMPORTANT NOTICE RECARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Iowa Electric Light and Power Company (IELP) for IELP's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending IELP's operating license of the Duane Arnold Energy Center 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 information in this document are contained in the contract between Iowa Electric Light and Power Company and General Electric Company for nuclear fuel and related ser-vices for the nuclear system for Duane Arnold Energy Center Unit 1, dated February 8,1968, and nothing contained in this document shall 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 may result from such use of such information.

3/4

e 23A1739 Rev. 0 E

1. PLANT UNIQUE ITEM (1.0)* m 3:

A Appendix A: Transient Analysis Input Paramiters d 1-Appendix B: Feedwater Controller Failure _

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

Fuel Type Cycle Loaded Number Number Drilled Irradiated -

P8DPB289 5 36 36 i

P8DPB289 6 84 84 -

P8DRB299 7 40 40 -

=

P8DRB284H 7 88 88 i

New t BP8DRB299 8 64 64 i BP8DRB30lL 8 56 56 .

3(8 Total 368

3. REFERENCE CORE LOADING PATTERN (3.3.1) c--

Nominal previous cycle core average exposure at end of cycle: 17954 MWD /ST -

Minimum previous cycle core average exposure at end =

of cycle from cold shutdown considerations: 17754 MWD /ST y Assumed reload cycle core average exposure at end _r of cycle: 17499 MWD /ST Core loading pattern: Figure 1 5

  • ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-6, April 1983. A letter "S" preceding the -5 number refers to the appropriate section in the United States Supplement, NEDE-240ll-P-A-6-US, April 1983. t 5

N

~

23A1739 Rev. O

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 eff Uncontrolled 1.112 Fully Controlled 0.960 Strongest Control Rod Out 0.987 R, Maximuu Increase in Cold Core Reactivity 0.003 with Exposure into Cycle, Ak

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (A) gpm (20*C, Xenon Free) 600 0.030

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

(COLD WATER INJECTION EVENTS ONLY)

Void Fraction (%) 41.8 Average Fuel Temperature (*F) 1280 Void Coefficient N/A* (c/% Rg) -9.11/-11.39 Doppler Coefficient N/A (C/*F) -0.225/-0.214 Scram Worth N/A (S) 7.

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

Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposure: BOC to EOC BP/P8x8R 1.20 1.47 1.40 1.051 6.493 112.9 1.23

  • N = Nuclear Input Data, A = Used in Tran;ient Analysis
    • Ceneric exposure dependent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-6, April 1983.

6

23A1739 Rev. 0

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: Yes Rod Withdrawal Limiter: No Thermal Power Monitor: No Improved Scram Time: Yes (0DYN Option B)

Exposure Dependent Limits: No Exposure Points Analyzed: EOC

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single Loop Operation: Yes Load Line Limit: Yes Extended Load Line Limit: Yes Increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No

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

Exposure: BOC to EOC

~

Flux Q/A ACPR Transient (% NBR) (% NBR) BP/P8x8R Figure Load Rejection Without Bypass sa8 113 0.16 2 Loss of Feedwater Heater 121 117 0.14 3 Feedwater Controller Failure

  • 289 111 0.12 4
  • See Appendix B 7

23A1739 Rev. 0

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

SUMMARY

(S.2.2.1)

(Generic Bounding Analysis Results)

ACPR Rod Block Reading (All Fuel Types) 104 0.13 105 0.16 106 0.19 107 0.22 108 0.28 109 0.32 110 0.36 Setpoint Selected: 105

12. CYCLE MCPR VALUES (S.2.2.)

Non-Pressurization Events Exposure Range: BOC to EOC BP/P8x8R Loss of Feedwater Heater 1.21 Fuel Loading Error 1.26*

Rod Withdrawal Error 1.23 Pressurization Events Exposure Range: BOC to EOC Option A Option B BP/P8x8R BP/P8x8R Load Rejection Without Bypass 1.28 1.20 Feedwater Controller Failure 1.24 1.21

  • Includes a 0.02 penalty due to variable water gap R-factor uncertainty.

8

23A1739 Rev. 0

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) si v Transient (psig) (psig) Plant Response MSIV Closure 1244 1275 Figure 5 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: Extrapolated Rod Block Line Decay Ratio: Figure 6 Reactor Core Stability Decay Ratio, x2 /x0 : 0.84 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 Channel Type BP/P8x8R 0.31

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Cap Misoriented Bundle Analysis: Yes Event Initial MCPR Resulting MCPR Misoriented 1.23 1.06

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

Bounding Analysis Results:

Doppler Reactivity Coef ficient Figure 7 Accident Reactivity Shape Functions: Figures 8 and 9 Scram Reactivity Functions: Figures 10 and 11 Plant Specific Analysis Results:

Parameter (s) not Bounded, Cold: None Resultant Peak Enthalpy, Cold: N/A Parameter (s) not Bounded, HSB: Accident Reactivity Resultant Peak Enthalpy, HSB: 274.8 9

23A1739 Rev. 0

17. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)

See " Loss-of-Coolant Accident Analysis Report for Duane Arnold Energy Center (Lead Plant)," June 1984 (NEDO-21080-03-1A).

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23A1739 Rev. 0

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1 3 5 7 91113151719212325272931333537394143 FUdL TYPE A = P8DRB299 D = BP8DRB30lt ' '

B = P8DRB284H E = P8DPB289 [": ff.;,

C = BP8DRB299 F = P8DPB289 i:177-i Figure 1. Reference Core Loading Pattern [h' ,

Y.Al 11 fN

==

s., -

23A1749 Rev. 0 1 NEUTRON FLU ( l VESSEL PRESS RISE (PSI) 2 AVE SURFACE G T FLUX 2 SAFETY V ALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 150.0 300.0 s eve a s? v a tt--E cr ew 100.0  :

J i 200.0 a

- : ~

5 50.0 W 100.0 N

-0 0 0 0 0 0

0. 0 . .
0. 0 ,,,, ,, ,, ,, , ,
0. 0 2.0 4.0 6. 0 0.0 2.0 4. 0 6. 0 TIME (SECONOS) TIME (SECONDS) 1LEVELCINCH-REF-SEP-SKRT) 1 VOID REACTIVITY 2 VESSEL STEA1 FLOW 2 00PPLER REACTIVITY 3 TURBINE STE kMFLOW 3 SCRA,M REACTIvlTY 20 t. 0 ' cEE0unTEo e_gy 1.0 A i 797 t og_a_rvgug7v 1;i.e .4 i

5 0. 0 N, - [

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(0

. .- 9, -s . 0

-100.0 -2.0

3. 3 2.0 4.0 C.0 0.0 2.0 4. 0 6. 0 TIME (SECONDS) TIME (SECONOS)

Figure 2. Plant Response to Generator Load Rejection Without Bypass 12

23A1739 Rev. 0 150.0 l

1 idEUfRON FLUX 1 VESEEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 PEl lEF VALVE FLOW 3 COR E INLET FLOW 3 BYPLSS VALVE FLOW 150.0

  • cc= : r u_ r v sue 100.0 8 0 0 0 0 I

{u 100.0 ,  ; ;

b v g 50.0 M

e W

50.0

0. 0 , , ,, , ,, ,,

. . , . , ~ . . .. .. .

c. e  ;  ; M t ,

0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECONOSI TIME (SECONOSI 1 LEV ELCINCH-REF-SEP-SKRT) i V01 ) REACTIVITY 2 VESEEL STEAMFLOW 2 DOP)LER REACTIVITY 31UR 3tNE STEAMFLOW 3, vgvo_

SCALMer REA.CilVITY le.t . e e tre patro en gu 1.0 ceruvvy d 2 0. 0 . g_.

J a c. u <- .,. -3 ._

ih 8

b 5 t. e -1.0

.a d

C0 ,- ., .,,.,th! -2.0

-, ~ ..

0.0 100.0 203.0 0. 0 180.0 200.0 TIME (SEC0f0S) f!NE (SECONDS)

Figure 3. Plant Response to Loss of 100*F Feedwater Heating 13

23A1739 Rev. 0 158.0 1 NEUTRON FLO ( 1 VESSEL PRESS RISE PSI) 2 AVE SURF ACE HEA l' LUX 2 SAFETY VALVE low 3 CORE INLET 7 LOW 3 RELIEF VALVE LOW 150.0 ' tecE 'u_Et  :'" 4 BYPASS VALVE LOW i....

. 4 5 , i g

l y 100.0 ,'; ,  ; '( y- . ; '

i ac

  • 5s. 0 1

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" $N4 50.0 L3

0. 0 . .. .

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-g  ;  ;  ;

0. 0 0.0 20.0 40.0 0. 0 20.0 40.0 TIME (SECONOSI TIME (SECONOSI f

1 LEVEL (INCE REF-SEP-SKRT) 1 V010 REACTIvlTY 2 VESSEL STEAMFLOW 2 DOPPLER REACTIVITr 3 TURBINE STE ANFLOW 3 SCRAM REACTIVITY 3.8 a. v. nt. a i_ or_ n e_ r v u. t iv.

35 0.6 a c. r_ e n_ w a rr_ o r. i_ nu.

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,\ . -2.0 0.0 20.0 40.0 0. 0 20.0 43.0

  • TIME (SECCNDS) TIME (SECONOS)

Figure 4. Plant Response to Feedwater Cont' roller Failure 14

23A1739 Rev. 0 1 NEUTRON F UX 1 VESSEL PRESS RISE (PSI) 2 AVE SURFA:E HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 15 0. 0 .

300.0 e evonee ua !uE rLOu 190.0 a s - -@- 200.0 h

z ri oc X'

~

50.0 100.0 0 0 0 0 0 00

0. 0 5.0 0.0 5.0 TIME (SECOPOSI TIME (SECDNOS) 1 LEVEL (INC4-REF-SEP-SKR T ) REACTIVITY 2 VESSEL STEAMFLOW 2 00PPL REACTIVITY 3 TURBINE STEAMFLOW 32 SCR AM, A:TIVITY 200.0 m c. r e_ n_ u a r. r_ri_e.n_ u l.0 e. nt. _ e_a- t v. u. t. t. v.

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1 s ?. 0 5 0. 8

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h 08 , , , ,, -1.0 a . . - - a a y

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-I t c. 0 -2.8 C. 3 5.0 0.0 5.0 TIME (SECONDS) TIME (SECONDS)

Figure 5. Plant Response to MSIV Closure (Flux Scram) 15

23A1739 Rev. O A NATURAL CIRCULATI DN B 100 PERCENT R0D L INE C ULT. PERFORMANCE _IMIT

1. 0 9. C -

n.

< .7d

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N Cl X

5 V-

.50

< \

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O

.25 N

C.C3 O. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 6. Reactor Core Decay Ratio 16

23A1739 Rev. 0

0. 0

-5. O to

-10.0 ,

jM o

f n -

'r b

-15.0 f

b 8 j g-20.0 /

Y es ca

-25.0 4/

A CALCULATED VALU'E-COLD CALCULATED VALU'i-HSB C BOUND VAL 280 C ;L/G COLD l D BOUND VAL 280 CAL /G HSB

-30.0

0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.

Figure 7. Fuel Doppler Coefficient in 1/A*C 17

23A1739 Rev. 0 20.0 '

A ACCID $NT FUNC TION 6 BOUNDING VALU E 280 CAL /G 17.5 15.0 2

i W 12.5 z

"000 , 9 o

10.0 -

[

b 7.5 M

E 5. 0 m

2.5

0. 0
0. 0 5. 0 10.0 15.'0 20.

ROD POSITION, FEET GUT Figure 8. Accident Reactivity Shape Function (Cold Startup) 18

23A1739 Rev. 0 20.O A ACCIDENT FUNC T I O'N 8 BOUNDING VALU E 280 CAL /G 17.5 15.0 .,_ a m

(CD l

W 12.5 /

"00 0 l-gj 10.0 a

J

'~

. 7.5 r

o s 5. 0 2.5

0. 0
0. 0 5. 0 10.0 15.0 20.0 ROD POSITION, FEET OUT Figure 9. Accident Reactivity Shape Function (Hot Startup) 19

23A1739 Rev. O s

A SCRAM FUNCTION B BCUNDI NG VALUE 280 CAL /G i

35.0 /

g 30.0 i ,

25.0 /

d a

m -20.0 n 8

5 p l15. 0 t~

B

$ -10. 0 i /

5. 0
0. 0 J

,, m -

0. 0 1.0 2. 0 3. 0 4.0 5.0 6. 0' ELAPSED TIME, SECONDS Figure 10. Scram Reactivity Function (Cold Startup) 20

23Al?39 Rev. 0 50,0 .

A SCRAM FUNCTION 8 BOUNDI NG VALUE 280 CAL /G

,0. 0 E 3 0. 0 J

J 3

D J

C 20.0 m 4

.)

C d 10.0 /

0. 0 ,

,, m

0. 0 1.0 2. 0 3. 0 4.0 5.0 6.0 ELAPSED TIME, SECONDS Figure 11. Scram Reactivity Function (Hot Startup) 21/22 l - - - - -

23A1739 Rev. O APPENDIX A TRANSIENT ANALYSIS INPUT PARAMETERS The values listed below were used as inputs to the licensing analyses rather than the values provided in Reference A-1, in order to reflect actual plant operating parameters.

Analysis Value Dome Pressure (psig) 1026 Turbine Pressure (psig) 975 Reactor Pressure (psia) 1055 Inlet Enthalpy (BTU /lb) 528.6 S/RV Lowest Setpoint (psig) 1110+1%

Re.f erence A-1. NEDE-24011-P-A-6-US, " General Electric Standard Application for Reactor Fuel (U.S. Supplement)", April 1983.

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23/24

23A1739 Rev. O APPENDIX B FEEDWATER CONTROLLER FAILURE The Feedwater Controller Failure (FWCF) event was analyzed at the 100%

power /87% flow point for the Extended Load Line Limit Analysis, since this point was found to be more conservative than the 100% power /100% flow point.

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