ML20003E570

Supplemental Reload Licensing Submittal for Monticello Nuclear Generating Plant Reload 8 (Cycle 9).

ML20003E570
Person / Time
Site:	Monticello
Issue date:	03/31/1981
From:	Engel R, Hilf C WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML112970784	List: ML112970783 ML112970817 ML112970818 ML20003E570
References
Y1003J01A16, Y1003J01A16-R01, Y1003J1A16, Y1003J1A16-R1, NUDOCS 8104060322
Download: ML20003E570 (24)

Text

. - - . --

Y1003J01A16 REV.1 l

9 CLASSI MARCH 1981 i

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR MONTICELLO NUCLEAR GENERATING PLANT l RELOAD 8 (CYCLE 9) l 1

4 l

l GENER AL h ELECTRIC

- 810'406 0 3W

i 4

1 i

Y1003J01A16

! Rev. 1 Class I a March 1981 -

I SUPPLEMENTAL RELOAD LICENSING SUBMITTAL i FOR MONTICELLO NUCLEAR GENERATING PLANT RELOAD 8 (CYCLE 9)  ;

4 i,

t i

d4 C. L. Hilf s

1 Reload Fuel Licensing i

i 1

4 Approved:

i R. E. Enge Manager Reload Fuel-Licensing i

1, NUCLEAR POWE3 SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 95125 GENER AL $ ELECTRIC i

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Y1003J01A16 Rev. 1 IMPORTANT NOTICE RECARDING CONTENTS OF T11IS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Northern States Power Company (NSP) for NSP's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending NSP's operating license of the Monticello Nuclear Generating Plant. 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 Northern States Power Company and General Electric Company for nuclear fuel and related services for the nuclear system for Monticello Nuclear Generating Plant, dated December 4,1967, 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.

11

Y1003J01A16 Rev. 1

1. PLANT-UNIQUE ITEMS (1.0)*

Plant Parameter Changes - see Appendix A Safety / Relief Valves - see Appendix A GETAB Initial Conditions- - see Appendix A Initial MCPR - see Appendix A Loading Error LHGR - see Appendix A Channels - see Appendix A l

RWM Operability Limit - see Appendix A l ODYN Code for Transient Analyses - see Appendix B .

Extended Exposure Fuel - see Appendix C

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

Fuel Type Number Number Drilled Irradiated, Reload 2 8DB262 12 None Irradiated, Reload 3 8DB250 28 None Irradiated, Reload 4 8DB219L 40 None Irradiated, Reload 5 8DB262 108 None Irradiated, Reload 6 8DRB265L 52 52 Irradiated, Reload 6 BDRB282 60 60 Irradiated, Reload 7 P8DRB282 56 56 Irradiated, Reload 7 P8DRB265L 44 44 New P8DRB265L 40 40 New P8DRB284LB** 44 44 Total 484 296

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 16.57 GWd/st.

Assumed reload cycle core average exposure at end of cycle: 16.82 GWd/st.

Core loading pattern: Figure 1.  ;

1

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

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

• • Letter MFN-196-80/REE-077-80, R.E. Engel (GE) to J.S. Berggren (NRC),

" General Electric Co. Licensing Topical Report NEDE-240ll-P-A-1, Generic Reload Fuel Application, Amendment j[", November 17, 1980.

l 1

1

)

Y1003J01A16 Rev. 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 1.1105 Fully Controlled 0.9547 Strongest Control Rod out 0.9899 ,

R, Maximum Increase in Cold Core Reactivity

.with Exposure.Into Cycle, dk 0.000 5 .' STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) ppm Shutdown Margin (ak)

(20*C, Xenon Free) 600* 0.037  ;

6. RELOAD-UNIQUE TRANSIENT' ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)

EOC9 Void Coefficient N/A** - (c/% Rg) -6.57/-8.22

+ Void Fraction (%)- 37.14

, Doppler Coefficient N/A (c/*F) -0.223/-0.212~

~

Average Fuel Temperature (*F) 1171 <

Scram Worth N/A ($) -37.05/-29.64' Scram Reactivity versus Time- See " Reactivity Components" on Figures 3, 4, and 5 l-i i'

~

i

• Previous documents have indicated a boron concentration of 900 ppm. but ,

that concentration is the concentration injected into the. core to produce _ -

a conservatively calculated concentration of 600 ppm due'to imperfect mixing and dilution from water in the cooldown circuit. The 600 ppm boron concen-tration is sufficient to provide the 3% Ak subcritical condition.

• • N = Nuclear Input Data

.A~=.Used in Transient Analysis

2

Y1003J01A16 Rev. 1

7. RELOAD-t3IQt'E GETAB TTsCSIENT ANALYSIS INITIAL CONDITION PAPJLVETERS (5.2)

EOC9 Exposure 8xS 8x8R P8x3R Peaking factor 1.22, 1.52, 1.40 1.20, 1.67, 1.40 1.20, 1.63, 1.40

~

(local, radial, axial) r.-Factor 1.098 1.052 1.052 Bundle Power (MWt) 5.131 5.618 5.496

}

Bundle Flov (10 lb/hr) 104.2 97.2 97.9 ]

Initial MCPR 1.42 1.42 1.46 }

8. SELECTED MARGIN IMPROVDIENT OPTIONS (5.2.2)

None

9. CORE-VIDE TRANSIDiT ANALYSIS REST 1TS (5.2.1)

Core p Nominal f.CFR 7 SI, Power Flow e Q/A y Plant Transient _Ex N sure (1) (!) ( NBR) {t NERi (psig) (pstr) Sx9/9x5R/F5x91 Response Generator Load E0C9 100 100 $78.6 724.8 1207 1231 0.35/0.35/0.39 Figure 3 Rejection without Bypass-Loss of 100*F -- 100 100 118.1 116.5 1023 1067 0.15/0.16/0.16 Figure &

Teedwater Heater Feedvater Con-trol'er Failure EOC9 100 100 518.2 124.3 1169 1201 0.33!!,.34/0.37 Tigure 5 ]

10. LOCAL ROD WIniDRAWAL ERROR (VITH LIMITING INSTRtHDIT FAILURE) TPXiSIENT SDSIARY (5.2.1)

LCYR I.HGR Rod Block Rod Position 8x8/8xSR 8x8/Sx8R Limiting Reading (Feet Withdrawn) and P8x8R and P8x8R Rod Pattern 104 3.0 0.15/0.15 13.5/15.8 105 3.0 0.15/0.15 13.5/15.8 Figure 6 106 3.5 0.20/0.19 14.6r17.2 107 3.5 0.20/0.19 14.6/17.2 10S* 4.0 0.27/0.23 15.3/18.2 109 4.5 0.32/0.25 15.6/18.6 110 4.5 0.32/0.25 15.6/18.6

• Indicates setpoint selected.

3

Y1003J01A16 Rev. 1

11. OPERATING MCPR LIMIT (5.2, APPENDIX C) 8x8 8x8R P8x8R Load Rejection without Bypass (EOC9)

Option A 1.48 1,48 1.52 )

Option B 1.43 1.43 1.47 ]

Feedwater Controller Failure (EOC9)

Option A 1.46 1.47 1.50 Option B 1.37 1.38 1.41 Loss of Feedwater Heating (BOC-EOC9) 1.22 1.23 1.23 Rod Withdrawal Error (BOC-EOC9) 1.34 1.30 1.30 Fue1 Loading Error (BOC-EOC9) 1.48 1.48 1.48

12. OVERPRESSURIZATION ANM.YSIS

SUMMARY

(5.3)

Power Core Flow Psi Pv Plant Transient (%) (%) fpsig) (psig) Response MSIV Closure 100 100 1223 1244 Figure 7 )

(Flux Scram)

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability:

0.55 Decay Ratio x /*

2 o }

(Natural Circulation-100% Red Line)

Channel Hydrodynamic Performance Decay Ratio (Natural Circulation-100% Pod Line) 8x8 Channel 0.12 P8x8R/8x8R Channel 0.07

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

(See Loss-of-Coolant Analysis, NEDO-24050-1) 4

4 Y1003J01A16 Rev. 1

15. LOADING ERROR RESULTS (5.5.4) i Limiting event: Mislocated Bundle MCPR: 1.07 4

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

^

Maximum' incremental control rod worth: 0.52% Ak 4

f 5

Y1003J01A16 Rev. 1

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MMM lIi11I 1 3 5 7 9111315171921232527293133353739414345474951 FUEL TYPE A= 808262, RELOAD 5 F = P80RB265L, RELOAD 7 B= 8DB219L.RELOA0 4 G= 80RB282, RELOAD 6 C= 8DB262, R El.OAD 2 H= P8DRB265L, NEW D= 809250, R ELO AD 3 I= 80RB265L. RELOAD 6 E = ' P8DRB282, RELOAD 7 J= P80R B284LB, NEW Figure 1. Reference Core Loading Pattern

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i l

Y1003J01A16 Rev. 1 This figure has been deleted because of the use of the ODYN code.

Scram reactivity versus time is indicated on the figures provided for the transients considered (Figures 3, 4, and 5).

Figure 2. Scram Reactivity and Control Rod Drive Specifications 7

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Y1003J01A16 Rev. 1 2 6 10 14 18 22 26 30 34 38 42 46 50 51 4 47 10 0 0 10 43 10 28 10 28 10 39 6 22 6 6 22 6 35 22 20 26 20 22 31 32 8 8 32 27 18 0 20 0 20 0 18 23 32 8 8 32 19 22 20 26 20 22 15 6 22 6 6 22 6 11 10 28 10 28 10 7 10 0 0 10 3 4 NOTES:

1. NL'MBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.

2. ERROR ROD IS (10,27).

Figure 6. Limiting RWE Rod' Pattern 11:

m.

i Y1003J01A16 Rev. 1 i

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1.2 ULTIMATE PERFORMANCE LIMIT 1.0 - - - - - - - ---------

0.8 -

^

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25 9

Q 0.6 -

m S

O NATURAL CIRCULATION 0.4 -

100 % ROD LINE s

0.2 -

0.0 I I I I O 20 40 60 80 100 PERCENT POWER Figure 8. Decay Ratio 13/14

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

Y1003J01A16 Rev. 1 Appendix A PLANT PARAMETER CRWGES Safety / Relief Valve - (Tables 5-4, page 5-62, Operating Plants Pressure Relief Systems) 8 S/R valves installed 7 S/R valves assumed in analysis. Capacity at setpoint = 83.0%

Lowest setpoint = 1108 + 1% psig GETAB Initial Conditions (Table 5-8, page 5-66)

See the revision to Table 5-8 enclosed with the letter, J. F. Quirk (GE) to Olan D. Parr (NRC), " General Electric Co. Licensing Topical Report NEDE-240ll-P-A, Generic Reload Fuel Application, Appendix D,,

Second Submittal," February 28, 1979.

Initial MCPR The initial MCPR for the 8x8R and P8x8R fuel was less than the operating limit MCPR. This is discussed on pp. B-ll4 and B-ll3 of the .

" Generic Reload Fuel Application," NEDE-240ll-P-A-1. .

Loading Error Results (5.5.4, Table 5-8, page 5-66)

LHGR: 21.2 kW/ft All Channels not supplied by C3 - At the direction of Northerm States Power Company, the analyses have ariumed that performance characteristics of channels not supplied by GE a .'4entical to the characteristics of channels supplied by GE.

RWM Opertbility Limit (Item 16) - 10%

15/16

Y1003J01A16 Rev. 1 Appendix B ODYN CODE FOR TRANSIENT ANALYSES All rapid pressurization and overpressure protection events have been analyzed using the ODYN transient code as specified in Reference B-1. The ACPR values given for the pressurization events in Section 9 are the plant-specific deter-ministic values calculated by ODYN based on the initial MCPR given in Item 7 of this submittal. These ACPRs may be adjusted to reflect either Option A or Option B ACPRE by employing the conversion method described in Reference B-2.

These adjustmeats n're based on conservatism factors applied to the ratio ACPR/ICPR. The 'regulting MCPR is calculated by adding the ACPR to the safety limit. The MCPRs resulting from the adjusted ACPRs from pressurization events as well as the MCPRs for nonpressurization events are presented in Section 11.

Code overpressure protection analysis results are deterministic as discussed in Reference B-2.

The operating limit MCPR is the maximum MCPR of the following events:

1. turbine trip or load rejection without bypass based on ODYN;

2. feedwater controller failure event based on ODYN;

3. loss of feedwater heating event;

4. rod withdrawal error event; and

5. bundle loading error accident; where the loss of feedwater heating, rod withdrawal error, and loading error MCPRs are calculated as described in Reference B-3 but the MCPRs for the pressurization events analyzed with ODYN have been adjusted as follows:

1. MCPRs adjusted for Option A (adding 0.044 to ACPR/ICPR) for all plants choosing to c; erate under Option A.

17

I Y1003J01A16 Rev. 1

2. MCPRs adjusted for Option B for all plants choosing to operate under Option B which meet all scram specifications given in Reference B-2.

'3.. MCPRs are determined by a linear interpolation between the Option A MCPR and 'the Option B MCPR for all plants choosing to operate under Option B which do not meet the scram time specification. This interpolation is based on the tested measured scram time and is described in Reference B-2.

REFERENCES

- B-1 ? Letter,RR. P. Denise (NRC) to G. G. Sherwood (CE), January 23, 1980

~ B-2 Lett'er (with attachment), R. H. Buchholz (GE) to P.S. Check (NRC),

" Response to-NRC Request for:Information on ODYN Computer Model,"

September 5, 1980.

. B-3f " Generic Reload ~ Fuel Application," NEDE-240ll-P-A-1, August 1979.

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18

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Y1003J01A16 Rev. 1 APPENDIX C EXTENDED EXPOSURE FUEL This extended exposure program description and initial safety evaluation is presen-ted in Reference C-1. The information contained in this appendix updates the safety evaluation for continued irradiation of the selected bundles for Cycle 9.

The information contained herein is presented in the same format as Reference C-1.

1.0 Propa -

Program The peak pellet exposure is calculated to be 52,300 mwd /ST at EOC9.

2.1 Fuel Rod Thermal Analysis Analyses performed for the extended exposure fuel bundles resulted in values (for 1% plastic strain limit) of 14.7 kW/ft at a peak pellet exposure of 52,300 mwd /ST for UO 2 fuel r ds and 15.5 kW/ft at 44,760 mwd /ST for urania-gadolinia rods.

2.1.1 Fuel Cladding Temperatures The inside, average, and outside cladding temperature during normal oper-ation at the end of Cycle 9 are calculated not to exceed 815 F, 782 F, and 750 F, respectively.

2.2.1 Cladding Creep Collapse The calculation demonstrated that cladding creep collapse is not expected to occur in the event of a maximum overpressure transient for an exposure of 52,300 mwd /ST.

2.2.2 Stress Evaluations Fuel rod stress analyses of the extended exposure bundles were performed with the model documented in Reference 1 (to NED0-24202) and ncminal input values (Reference C-2) for operation through Cycle 9. Fuel design ratios were shown to be well below 1.0 19

Y1003J01A16 Rev. 1 1

2.2.4 Fatigue Evaluation The cumulative fatigue damage is calculated to be less than the allowable fatigue limit for Cycle 9.

I i 2.3.3 Fretting Wear'and Corrosion The fuel bundles which will operate to higher exposures will again be

_ visually examined before loading in Cycle 9.

' 3.1.1 ' Reactivity Hot reactivity of the extended exposure fuel bundles decreases by approx-imately 0.003 Ak, from 45,000 mwd /ST to 55,000 mwd /ST.

'3.1.2' ' Local Peakin'g' Factors Calculated maximum local peaking factor for the extended exposure fuel l bundles increases by approximately 0.04 from 45,000 mwd /ST to 55,000 mwd /ST. .

L3.3.5 Accident Evaluations

)

. The new MAPLHCR values and associated peak cladding temperatures and 4

oxidation fractions are -given .in Reference C-3.

4. References

- C-1 "Monticello Nuclear Generating' Plant Extended Exposure Fuel Program'#,:

.NEDO-24202, dated July.1979.

C-2 R. L..Tedesco.(NRC) letter to R. E..Engle'(GE), " Acceptance for Licensing Reference of Changes to Topical Report Number NEDE-240ll 1 P-A-1,:' Generic Reload Fuel Application', dated August 1979",

' dated. November _7, 1980.

C-3 "L'oss-of-Coolanr Accident Analysis Report for Monticello Nuclear

. Generating[Plant",'NEDO-24050-1, dated December 1960.

20 (Final)

A,____.___1.___.__'_i_-___ ___-._.____._______m5_