ML19345F249

Supplemental Reload Licensing Submittal for Facility Reload 8 (Cycle 9), Class 1

ML19345F249
Person / Time
Site:	Monticello
Issue date:	12/31/1980
From:	Engel R, Hilf C GENERAL ELECTRIC CO.
To:
Shared Package
ML112970812	List: ML112970811 ML112970813 ML19345F249 ML19354C423
References
Y1003J01A16, Y1003J1A16, NUDOCS 8102170015
Download: ML19345F249 (24)

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Y1003J01A16 Rev. O Class I December 1980 l

l SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR MONTICELLO NUCLEAR GENERATING PLANT RELOAD 8 (CYCLE 9) d.c C. L. Hilf Reload Fuel Licensing Approved:

R. E. Eage , Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CAllFORNi A 95125 GENER AL h ELECTRIC L

%/6 2- rio od

I I

Y1003J01A16 Rev. O I IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY I

l This report was prepared by General Electric solely for Northern States Power Company (NSP) for NSP's use with 'he 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 f

j 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 resc ecting information I

l in this document are contained in the contract between Northern States Power l 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

}

j 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 f General Electric Company nor any of the contributors to this document makes I

i any representation or warranty (2xpress or implied) as to the complettaess,

! accuracy or usefulness of the information contained in this document or thc' j

such use of such information may not infringe privately owned rights; nor do i they assume any responsibility for liability or damage of any kind which may result from such use of such information.

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G Y1003J01A16 Rev. O I

k 1. PLANT-UNIQUE ITEMS (1.0)*

i Plant Parameter Changes l

l Srfety/ Relief Valves - see Appendix A

! GETAB Initial Conditions - see Appendix A Initial MCPR - see Appendix A l

Loading Error LHGR - see Appendix A channels - see Appendix A RWM Operability Limit - see Appendix A ODYN Code for Transient Analyses - see Appendix B i Extended Exposure Fuel - see Appendix C

l

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

Fuel Type Number Number Drilled tradiated, 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, Reinad 6 8DRB282 60 60 Irradiated, Reload 7 P8DRB282 56 56 Irradiated, Reload 7 P8DRB265L 44 44 j

New P8DRB265L 40 40 New P8DRB284LB** 44 -

44 I

Total 484 296 l'

l 3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of. cycle: 17.04 GWd/st.

Assumed reload cycle core average exposure at end of cycle: 16.87 GWd/st.

Core loading pattern: Figure 1.

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

NEDE-240ll-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 9", November 17, 1980 _

1

Y1003J01A16 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)

BOC k eff Uncontrolled 1.110 Fully Controlled 0.954 Strongest Control Rod Out 0.989 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, Ak 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 Coef ficient N/A** (c/% Rg) -6.55/-8.19 Void Fraction (%) 37.14 Doppler Coefficient N/A (C/ F) -0.214/-0.203 Average Fuel Temperature (*F) 1171 Scram Worth N/A (S) -37.05/-29.64 Scram Reactivity versus Time See " Reactivity Components" l

on Figures 3, 4, and 5 l

l

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

l l

i **N = Nuclear Input Data A = Used in Transient Analysis l 2

Y1003J01A16 Rev. 0

7. RELOAD-UNIQUE GETAB TRA"SIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

EOC9 Exposure 8x8 8x8R P8x8R Peaking factor 1.22, 1.53, 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) 5.157 5.642 5.496 Bundle Flow (10 lb/hr) 104.1 101.1 102.0 Initial MCPR 1.40 1.40 1.44

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

None

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Core p p Nominal t.CPR Power Flow t Q/A SL v Plant Transtent Exposure (%) (I) (1 NBR) (% NRR) (psig) (psig) 8x8/8x8R/P8x8R Response Generator lead EOC9 100 100 689.3 123.9 1206 1230 0.33/0.33/0.37 Figure 3 Rejection without Bypass Loss of 100'F --

100 100 118.1 116.5 1023 1067 0.15/0.16/0.16 Figure 4 Feedwater Heater Feedwater Con- EOC9 100 100 547.4 123.9 1169 1201 0.32/0.32/0.35 Figure _

troller Failure

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

SLW1ARY (5.2.1) l ACPR LHGR l

l Rod Block Rod Position 8x8/8x8R 8x8/8x8R Limiting Reading (Feet Withdrawn) and P8x8R and P8x8R Rod Pattern 104 3.0 0.13/0.15 13.5/15.8 105 3.0 0.15/0.15 13.5/15.8 Figure 6 106 3.5 0.20/0.19 14.6/17.2 107 3.5 0.20/0.19 14.6/17.2 108* 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.

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Y1003J01A16 Rev. 0

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

Option A 1.49 1.49 1.53 Option B 1.41 1.41 1.45 Feedwater Controller Failure (EOC9)

Option A 1.47 1.47 1.51 Option B 1.36 1.36 1.39 Loss of Feedwater Heating (BOC-EOC9) 1.22 1.23 1.23 Rod Withdrawal Error (BOC-EOC9) 1.34 1.30 1.30 Fuel Loading Error (BOC-EOC9) 1.48 1.48 1,48

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5. 3)

Power Core Flow Psl Pv Plant Transient _1%) (%) __ (psig) (psig) Response MSIV Closure 100 100 1222 1243 Figure 7 (Flux Scrain)

13. STABILITY ANALYr.IS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability:

Decay Ratio, x2/* o 0.50 (Natural Circulation-100% Rod Line)

Channel Hydrodynamic Performance Decay Ratio (Natural Circulation-100% Rod 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

a - -

Y1003J01A16 Rev. 0

15. LOADING ERROR RESULTS (5.5.4)

Limiting event: Mislocated Bundle MCPR: 1.07

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Maximum incremental control rod worth: 0.52% ak a

5

Y1003J01A16 Rev. 0 52 3 M[0]T 5o @

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MMM IIiilI 1 3 3 7 9111315171921232527293133353739414345474951 FUEL TYPE B B ,RE O 4 G 82 R L AD 6 D= 8 RE AD = 8DRB L REL AD 6 E = P80RB282, RELOAD 7 J= P80R B284 LB, N EW Figure 1. Reference Core Loading Pattern 6

Y1003J01A16 Rev. 0

• his figure has been deleted because of the tse of the ODYN code.

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

i l

l Figur. . Scram Reactivity and Control Rod Drive Specifications 7

I NEUTRON FLUX 1 vf "," f t orts niv fesli I 2 Sf 4F i i r Vr_ vt fi t a

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

1. NLMBER INDICATES NUMBER OF NOTCHES WITHDRAW';

OUT OF 48. BLANK IS A WITHDRAWN ROD.

2. ERROR ROD IS (10,27).

Figure 6. Limiting RWE Rod Pattern 11

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Figure /. Plant Response to MSIV Closure < o

f Y1003J01A16 Rev. 0 1.2 ULTIMATE PERFORMANCE LIMIT 1.0 - - - " -"----""""""" 0.8 -

                    ~o E

_E 9 E 0.6 - x S O 0.4 - N ATUR A L CIRCULATION 100% ROD LINE 0.2 - I ' ' 0 O 20 40 60 80 100 PERCENT POWER Figure 8. Decay Ratio 13/14

[ i l l Y1003J01A16 Rev. 0 l Appendix A PLANT PARAMETER CHANGES 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 i The initial MCPR for the 8x8R and P8x8R fuel was less than the , operating limit MCPR. This is discussed on pp. B-114 and B-115 of the

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

, Loading Error Results (5.5.4, Table 5-8, page 5-66) l LHGR: 21.2 kW/ft All Channels not supplied by GE - At the direction of Northerm States Power Company, the analyses have assumed that performance characteristics of channels not supplied by GE are identical to the characteristics of channels supplied by GE. RWM Operability L!riit (1 tem 16) - 10% l 15/16 k

i 11003J01A16 Rev. O 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 ACPRs by employing the conversion method described in Reference B-2. These adjustments are based on conservatism factors applied to the ratio ACPR/ICPR. The resulting MCPR is calculated by adding the ACPR to che safety 4 limit. The MCPRs resulting f rom 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 cr load rejection without bypass based on ODYN;

2. feedwater controller failure avent 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 operate under Option A.

17

Y1003J01A16 Rev. 0

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 enoosing toeoperate 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, R. P. Denise (NRC) to G. G. Sherwood (GE), January 23, 1980 B-2 Letter (with attachment), R. H. Buchholz (CE) to P.S. Check (NRC),

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

September 5, 1980. B-3 " Generic Reload Fuel Application," NEDE-240ll-P-A-1, August 1979. i I i Y e e O 18 '

Y1003J01A16 Rev. O APPENDIX C EXTENDED EXPOSURE FUEL This extended exposure program description and inft.ial 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 Proposed Program The peak pellet exposure is calculated to be. 52,300 FNd/ST at EOC9. 2.1 Fuel .".od 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 l 750 F, respectively. l l 2.2.1 Cladding Creep Collapse { l The calculation demonstrated that cladding creep collapse is not expected to occur in the event of a maximum overpressure transient for an exposure l of 52,300 mwd /ST. l l 2.2.2 Stress Evaluations Fuel rod stress analyses of the extended exposure bundles were perfonned with the model documented in Reference 1 (to NED0-24202) and nominal input values (Reference C-2) for operation through Cycle 9. Fuel design ratios were shown to be well below 1.0 l 19 I L_--__________-_-_____-_-____-_

Y1003J01A16 Rev. 0 2.2.4 Fatigue Evaluation The cumulative fatigue damage is calculated to be less than the allowable fatigue limit for Cycle 9. 2.3.3 Fretting Wear and Corrosion The fuel b indles 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 Peaking Factors Calculated maximum local peaking factor for the extended exposure fuel bundles increases by approximately 0.04 from 45,000 mwd /ST to 55,00 mwd /SI 3.3.5 Accident Evaluations The new MAPLHCR values and associated peak cladding tera-:atures and oxi-( . tion fractions are given in Reference C-3.

4. References C-1 "Monticello Nuclear Generating Plant Extended Exposure Fuel Program",

NE90-24202, dated July 1979. C-2 R. L. Tedesco (NRC) letter to R. E. Engel (GE), " Acceptance for Licen-sing Reference of Changes to Topical Report Number NEDE-240ll-P-A-1,

               ' Generic Reload Fuel Appil_otion', dated August 1979", dated November 7, 1980.

C-3 " Loss-of-Coolant Accident Analysis Report for Monticello Nuclear Gen-erating Plant", NED0-24050-1, dated December 1980. 20 (Final)

EXHIBIT D License Amendment Request dated February 6,1981 9ccket No. 50-263 License No. DPR-22 Exhibit D is the 7 - sf-Coolant Accident Analysis Report for Monticello Nuclear Generating Plant.}}