ML19296B337

Supplemental Reload Licensing Submittal for Reload 5.

ML19296B337
Person / Time
Site:	Cooper
Issue date:	12/31/1979
From:	Engel R, Hilf C GENERAL ELECTRIC CO.
To:
Shared Package
ML19296B333	List: ML19296B332 ML19296B334 ML19296B337
References
79NED412, NEDO-24230, NUDOCS 8002200429
Download: ML19296B337 (26)

Text

?Y "9E DECEMBER 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR COOPER NUCLEAR STATION UNIT 1 RELOAD 5 Oh h 8

O GENER AL h ELECTRIC eoo22 - y2c.

NEDO-24230 79NED412 Class I December 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR COOPER NUCLEAR STATION, UNIT 1 3

RELOAD 5 Prepared:

C. L. Hi f, Licensing Engineer Reload Fuel Licensing Approved:

R. E. Enge , Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 95125 GENERAL $ ELECTRIC

NED0-24230 IMPORTANT NOTICE RECARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report uae prepared by General Electric solely for Nebraska Public Pouer District (NPPD) for NPPD's use with the U.S. Nuclear Regulatory Comission (USNRC) for amending NPPD's operating license of the Cooper Nuclear Station. The infonnation contained in this report is believed by General Electric to be an accurate and true representation of the facts knoun, obtained or prov':ded 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 Nebraska Public Power District and General Electric Company for nuclear fuct and related services for the nuclear system for Cooper Nuclear Station and nothing contained in this docierent ehali be construed as changing said contract. The use of this information except as defined by said contract, or for any pur-pose other than that for which it is intended, is not authori::ed; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributore to this document makes any representation or warranty (express or implied) as to the completenese, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately ovned rights; nor do they assume any responsibility for liability or damage of any kind which may ,

result from auch use of such information.

6

NEDO-24230

1. PLANT-UNIQUE ITEMS (1.0)*

Bundic 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 cnd 4.0)

Fuel Type Number Number Drilled Irradiated 7DB250 (Initial Core) 80 80 Irradiated 8DB250 (R1) 32 32 Irradiated 8DB274L (R1&2) 112 112 Irradiated 8DB274L (R3) 24 24 Irradiated 8DRB283 (R3) 76 76 Irradiated 8DRB283 (R4) 112 112 New P8DRB283 (RS) 80 80 New P8DRB265L (RS) 32 32 Total 548 548

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure: 16,590 mwd /t Assumed reload cycle exposure: 17,110 mwd /t Core 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 1.102 Fully Controlled 0.947 Strongest Control Rod Out 0.987 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, ok 0.000

• ( ) refers to areas of discussion in " Generic Reload Fuel Application,"

NEDE-24011-P-A-1, August 1979.

1

NED0-24230

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (ak) pygg (20*C, Xenon Free) 600 0.043

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

EOC6 Void Coefficient N/A* (c/% Rg) -7.467/-9.33 Vold Fraction (%) 40.03 Dogpler Coefficient N/A (c/%*F) -0.225/-0.214 Average Fuel Temperature (*F) 1357 Scram Worth N/A ($) -37.86/-30.29 Scram Reactivity vs Time Figure 2

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 7x7 8x8/8x8R/P8x8R Exposure EOC6 EOC6 Peaking factors (local, (1.24, 1.385, 1.4) (1.20, 1.537, 1.4) radial and axial)

R-Factor 1.08 1.054 Bundle Power (MWt) 5.893 6.527 Bundle Flow 118.9 110.4 (103 lb/hr)

Initial MCPR 1.19 1.24

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

None

• N = Nuclear Input Data A = Used in Transient Analysis 2

NEDO-24230

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1) 0 Core h Q/A p p Power Flow (% (% s1 v 8x8R/ Plant Transient Exposure (%) (%) NBR) NBR) (psig) (psig) 7x7 8x8 P8x8R Response Load BOC-EOC6 104.3 100 240 114 1194 1215 0.12 0.17 0.18 Figure 3 Rej ec tion Without Bypass Loss of BOC-EOC6 104.3 100 124 122 1022 1069 0.12 0.14 0.14 Figure 4 100*F Feedwater Heating Feedwater BOC-EOC6 104.3 100 171 115 1125 1167 0.09 0.14 0.14 Figure 5 Con-troller Failure

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

SUMMARY

(5.2.1)*

Rod Rod Block 0 " ( "!' } Limiting (Feet Reading (%) Withdrawn) 8x8 8x8R/P8x8R 8x8 8x8R/P8x8R Rod Pattern 104 3.5 0.09 0.08 14.5 15.6 Figure 6 105 4.0 0.11 0.09 15.0 16.6 Figure 6 106 4.5 0.12 0.10 15.2 16.9 Figure 6 107** 6.0 0.18 0.12 15.9 17.8 Figure 6 108 7.0 0.20 0.13 15.1 17.3 Figure 6 109 9.0 0.20 0.15 13.6 15.8 Figure 6 110 12.0 0.19 0.18 13.0 15.1 Figure 6

• Since the 7x7 fuel is on or near the periphery of the core far removed from the error rod, the Rod Withdrawal Error has no impact on this fuel.

• • Indicates setpoint selected.

• • • Includes densification penalty of 2.2%.

3

NEDO-24230

11. OPERATING MCPR LIMIT (5.2)

BOC TO EOC6 8x8R/

7x7 8x8 P8x8R 1.20 1.25 1.25 ,

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3)

Power Core Flow s1 y Plant Transient (%) (%) (psig) (psig) Response "3 1270 1295 104 100 Figure 7 F1 x a

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability:

Decay Ratio, x2 /*0 '

(105% Rod Line - Natural Circulation Power) 0.78 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 (105% Rod Line - Naturni Circulation Power) 8x8R/P8x8R channel 0.28 8x8 channel 0.37 7x7 channel 0.22 4

NEDO-24230

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

P8DRB265L Exposure MAPLIIGR P. C. T. Local Oxidation

. (mwd /t) (kw/ft) ('F) Fraction 200 11.6 2101 0.023 1,000 11.6 2106 0.023 5,000 12.1 2147 0.026 10,000 12.1 2137 0.025 15,000 12.1 2145 0.026 7

20,000 11.9 2127 0.024 25,000 11.3 2054 0.019 10,000 10.7 1970 0.014 P8DRB283 Exposure MAPLilGR P. C. T. Local Oxidation (mwd /t) (kw/ft) (*F) Fraction 200 11.2 2079 0.022 1,000 11.2 2073 0.021 5,000 11.8 2139 0.026

- 10,000 12.0 2152 0.027 15,000 12.1 2171 0.028 20,000 11.8 2155 0.027 25,000 11.3 2089 0.022 30,000 11.1 2064 0.020 5

NEDO-24230

15. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Doppler Reactivity Coefficient Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 Plant specific analysis results Parameter not bounded: None

16. LOADING ERROR RESULTS (5.5.4)

Limiting Event: Rotated Bundle (8DRB283)

MCPR: 1.07 6

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= o888888880 ll IilIIIII 01 03 05 07 09 ,11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 FUEL TYPE A= INITIAL COR ti, 7D8250 E = RELOAD 3, 80RB283 8 - RELOAD 1, 808250 F = RELOAD 4, 80RB283 C= RELOAD 1 AND 2, 8DB274L G= RELOAD 5, P80RB283 0= R ELOAD 3, BDB27dL H= RELOADS, P80RB265L Figure 1. Reference Core Loading Pattern 7

NEDO-24230 100 45 C - 67d CRD IN PERCENT 1 - NOMINAL SCHAM CUPVE IN (-8)

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NEDO-24230 02 06 10 14 18 22 26 51 47 10 10 43 12 8 39 8 35 14 34 31 8 0 Notes: 1. Rod Pattern Is 1/4 Core Mirror Symmetric Upper Left Quadrant Shown on Map

2. Numbers Indicate Number of Notches Withdrawn out of 48. Blank Is a Withd:cyn Rod

3. Error Rod is (26,31)

Figure 6. Limiting RWE Rod Pattern 12

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NED0-24230 1.2 ULTIMATE STABILITY LIMIT 1.0 - - - = = = = - - - - - - = = = = = - - - -

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NEDO-24230

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Figure 9. Doppler Reactivity Coef ficient Coraparison for RDA 15

NEDO-24230 20 O SOUNDING VALUE FOR 280 cal /g O CALCULATED VALUE 16 -

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NEDO-24230 so O BOUNDING VALUE FOR 290 cal /g 40 -

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NED0-24230 100 O BOUNDING VALUE FOR 280 cal /g O CALCULATED VALUE 80 -

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NEDO-24230 APPENDIX A BUNDLE LOADING ACCIDENT The loading error accidents (mislocated bundle and rotated bundle) for Cooper Cycle 6 have been analyzed using the " revised methods" (Reference A-1). These analysis methods have been previously applied to Cooper (References A-2 and A-3) and these applications approved by the NRC (References A-4 and A-5).

The loading errors will not influence the operating CPR limit. The most severe MLHGR predicted for a misloaded bundle is 16.1 kW/f t. This vclue includes an allowance of 2.2% for power spiking due to fuel densification (see Appendix C).

REFERENCES A-1 Letter, D.G. 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 (NPPD), May 2,1978.

A-5 Letter, T.A. Ippolito (NRC) to J. Pilant (NPPD), August 25, 1978.

A-1/A-2

ndDU-24230 APPENDIX B GETAB INITIAL CONDITIONS Table 5-8 of Reference B-1 states the "Nonvarying Plant GETAB Analysis Initial Conditions". The Cooper core pressure is given as 1045 psia. A value of 1035 psia, which more nearly reflects actual plant data, was assumed for this submittal.

Reference B-1 has been amended to eliminate this discrepancy.

REFERENCES B-1 Licensing Topical Report, " General Electric Boiling Water Reactor, Generic Reload Fuel Application", NEDE-240ll-P-A-1, August 1979.

B-l/B-2

NED0-24230 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 th plant Technical Specifi-cation for those operating BWR's for which it can be shawn that the predicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGR's".

The Cooper Reload-5 submittal contains the required information to remove the power spiking penalty from the Cooper Technical Specifications. Section 10, Rod Withdrawal Erro'r, and Appendix A (Bundle Loading Accident) include the den-sification effect in the calculated LHCR of the 8x8 fuels.

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

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I C-1/C-2

9 b ,

l GEN ER AL h ELECTRIC