ML20154F793

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Rev 0 to J11-03354-10, Supplemental Reload Licensing Rept for CNS Reload 18,Cycle 19
ML20154F793
Person / Time
Site: Cooper Entergy icon.png
Issue date: 08/31/1998
From: Brohaugh T, Watford G
GENERAL ELECTRIC CO.
To:
Shared Package
ML20154F718 List:
References
J11-03354-10, J11-03354-10-R00, J11-3354-10, J11-3354-10-R, NUDOCS 9810090376
Download: ML20154F793 (24)
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GENuclearEnergy J11-03354-10 Revision 0 Class I August 1998 Supplemental Reload Licensing Report for COOPER N'UCLEAR STATION Reload 18 Cycle 19 MA lES U$$$E P PDR, u- - - - -

O GE Nuclacr Energy J11-03354-10 Revision 0 Class I August 1998 - ,

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J11-03354-10, Rev. 0 l

Supplemental Reload Licensing Report for Cooper Nuclear Station Reload 18 Cycle 19 l

1 i

i Approved

// /

Approved

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. A. ford, Manager T. R. Brohaugh Nuclear Fuel Engineering Fuel' Project Manager

COOPER STATION Jil-03354-10 Reload 18 Rev. 0 Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by General Electric Company (GE) solely for Nebraska Public Power District (NPPD) for NPPD's use with the U. S. Nuclear Regulatory Commission (USNRC) to amend NPPD's operating license of the Cooper Nuclear Station. The information contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this report was prepared.

The only undertakings of GE respecting information in this document are contained in the contract between NPPD and GE for nuclear fuel and related services for the nuclear system for Cooper Nuclear Station 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 GE nor any of the contributors to this document makes any representation or warranty (expressed 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.

I Page 2 i.--

COOPER STATION J11-03354-10 Reload 18 Rev. 0 Acknowledgment The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by R.H. Szilard. The Supplemental Reload Licensing Report was prepared by R.H. Szilard. This document has been verified by Carmen Alonso.

Page 3

COOPER STATION Jil-03354-10 {

Reload 18 Rev. O j The basis for this report is Geveral Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A- .

13, August 1996; and the U.S. Supplement, NEDE-24011-P-A-13-US, August 1996. j l

1. Plant-unique Items l

Appendix A: Analysis Conditions

)

Appendix B: Decrease in Core Coolant Temperature Events l

Appendix C: SRV Tolerance Analysis  !

Appendix D; One Turbine Bypass Valve Out of Service

2. Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated: l l

l GE9B-P8DWB348-11 GZ-80M-150-T (GE8x8NB) 16 48 GE9B-P8DWB348-12GZ-80M-150-T (GE8x8NB) 16 24 ,

GE9B-P8DWB348-llGZ-80M 150-T(GE8x8NB) 17 152 GE9B-P8DWB348-1 1 GZ-80M-150-T (G E8x8NB) 18 4 GE98-P8DWB350-10GZ-80U-150 T(GE8x8NB) 18 160 New:

G E9B-P8 DWB350-10GZ l-80U-150-T (GE8x8NB) 19 100 GE9B-P8DWB350-10GZ-80U-150-T (GE8x8NB) 19 60 Total ,

548 i

3. Reference Core Loading Pattern' Nominal previous cycle core average exposure at end of cycle: 26737 mwd /MT

( 24255 mwd /ST)

Minimum previous cycle core average exposure at end of cycle 26406 mwd /MT from cold shutdown considerations: ( 23955 mwd /ST)

Assumed reload cycle core average exposure at beginning of 15794 mwd /MT cycle: ( 14328 mwd /ST)

Assumed reload cycle core average exposure at end of cycle: 27037 mwd /MT

( 24528 mwd /ST)

Reference core loading pattern: Figure 1

' The end of cycle core average exposure reflects the basis for the license work.

Page 4

COOPER STATION J11-03354-10 Reload 18 Rev. 0

4. Calculated Core Effective Multiplication and Control System Worth - No Voids,20*C Beginning of Cycle, k greet e ye Uncontrolled 1.106 Fully controlled 0.967 Strongest control rod out 0.989 R, Maximum increase in cold core reactivity with exposure into cycle, Ak 0.001
5. Standby Liquid Control System Shutdown Capability Boron (ppm) Shutdown Margin (Ak)

(at 20 C) (at 20 C, Xenon Free) 660 0.038

6. Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters s

Exposure: BOC19 to EHFP19-2205 mwd /MT (2000 mwd /ST)

Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

( M W t) (1000 lb/br)

GE8x8NB 1.20 1.72 1.40 1.000 7.297 101.0 1.18 Exposure: EHFP19-2205 mwd /MT (2000 mwd /ST) to EHFP19 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE8x8NB 1.20 1.66 1.40 1.000 7.041 102.5 1.23 Page 5

COOPER STATION J11-03354-10 Reload 18 Rev. 0

7. Selected Margin Improvement Options Recirculation pump trip: No Rod withdrawal limiter: No Thermal power monitor: No Improved scram time: Yes (ODYN Option B)

Measured scram time: No Exposure dependent limits: Yes Exposure points analyzed: 2 (EHFP-2205 mwd /MT, EHFP)

8. Operating Flexibility Options Single-loop on: ration: Yes Load line limit: Yes Extended load line limit: Yes Increased core flow throughout cycle: No Increased core flow at EOC: No

' Feedwater temperature reduction throughout cycle: No Final feedwater temperature reduction: No ARTS Program: Yes Maximum extended operating domain: No Moisture separator reheater out of service: No Turbine bypass system out of service: No One turbine bypass valve out of service: Yes Safety / relief valves out of service: No Feedwater heaters out of service: No ADS out of service: No EOC RPT out of service: No Main steam isolation valves otit of service: No h'

Page 6 l

l.

COOPER STATION J11-03354-10 Reload 18 Rev. 0

9. Core-wide AOO Analysis Results Methods used: GEMINI; GEXL-PLUS Exposure range: BOC19 to EHFP19-2205 mwd /MT (2000 mwd /ST)

Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.

(%NBR) (%NBR)

FW Controller Failure 202 115 0.12 2 Load Reject w/o Bypass 290 115 0.11 3 Turbine Trip w/o Bypass 275 113 0.10 4 Exposure range: EIIFP19-2205 mwd /MT (2000 mwd /ST) to EHFP19 Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.

(%NBR) (%NBR)

FW Controller Failure 293 120 0.17 5 Load Reject w/o Bypass 358 119 0.16 6 Turbine Trip w/o Bypass 335 117 0.15 7

10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary Rod withdrawal error (RWE) is analyzed in GE Licensing Report, Extended Load Line Limit and ARTS Improvement Program Analysesfor Cooper Nuclear Station Cycle 14, NEDC-31892P, Revision 1, May 1991. A cycle-specific analysis was performed for this cycle to verify that the ARTS RWE generic limits in NEDC-31892P remain valid with the use of the new fuel design. The results obtained verified that the existing ARTS limits are still valid for this cycle.
11. Cycle MCPR Values 2 In agreement with commitments to the NRC (letter from M.A. Smith to the Document Control Desk, 10CFR Part 21, Reportable Condition, Safety Limit MCPR Evaluation, May 24,1996) a cycle-specific Safety Limit MCPR calculation was performed, and has been reported in both the Sar ety Limit MCPR and Operating Limit MCPR shown below. This cycle specific SLMCPR was determined using the analysis basis documented in GESTAR with the following exceptions:

2 For single-loop operation, the MCPR operating limit is 0.01 greater than the two-loop value.

Page 7

i COOPER STATION J11-03354-10 Reload 18 Rev. 0

1. The reference core loading was analyzed.

. 2. The actual bundle parameters (e.g., local peaking) were used.

3. The full cycle exposure range was analyzed.

Safety limit: 1.06 Single loop operation safety limit: 1.0)

Non-pressurization events:

Exposure range: BOC19 to EHFP19 GE8x8NB Loss of100 F feedwater heating 1.17 Fuel Loading Error (misoriented) 1.20 Fuel Loading Error (mislocated) 1.21 Rod withdrawal error (for RBM setpoint to 111%) 1.24 Pressurization events:

Exposure range: BOC19 to EHFP19-2205 mwd /MT (2000 mwd /ST)

Exposure point: EHFP19-2205 mwd /MT (2000 mwd /ST)

Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.29 1.22 Load Reject w/o Bypass 1.27 1.20 Turbine Trip w/o Bypass 1.25 1.18 Exposure range: EHFP19-2205 mwd /MT (2000 mwd /ST) to EHFP19 Exposure point: EHFP19 Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.32 1.24 Load Reject w/o Bypass 1.31 1.23 Turbine Trip w/o Bypass 1 30 1.22

12. Overpressurization Analysis Summary Psi Py Plant Event (psig) (psig) Response MSIV Closure (Flux Scram) 1222 1246 Figure 8 Page 8 i.

COOPER STATION J11-03354-10 Reload 18 Rev. 0

( 13. Loading Error Results Variable water gap misoriented bundle analysis: Yes 3 Event ACPR Fuel loading error (Misoriented) 0.14 Fuel loading error (Mislocated) 0.15

14. Control Rod Drop Analysis Results Cooper Nuclear Station operates in the banked position withdrawal sequence (BPWS), so the control rod drop accident analysis is not required. NRC approval to use the generic analysis is documented in NEDE-24011-P-A-13-US, August 1996. CNS implemented the BPWS into the Rod Worth Minimizer (RWM) as documented in License Amendment No. I17. Removal of the Rod Sequence Control System (RSCS) at CNS has been approved by the NRC in License Amendment No.156.
15. Stability Analysis Results Cooper Nuclear Station has implemented the Option 1-D stability solution, documented in the reference, Application ofthe " Regional Exclusion with Flow-Biased APRMNeutron Flux Scram " Stability Solution (Option 1-D) to the Cooper Nuclear Station, Licensing Topical Report, GENE-Al3 00395-01, Class I, November,1996. The continued validity of the reference results for Cycle 19 has been confirmed.
16. Loss-of-Coolant Accident Results LOCA method used: SAFER /GESTR-LOCA See the Cooper Nuclear Station SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis, NEDC-32687P, Revision 1, March 1997. The LOCA analysis results presented in NEDC-32687P is conservatively analyzed for the GE8x8NB fuel types. This analysis yields a Licensing Basis peak clad temperature (PCT) of 1570 F, a peak local oxidation fraction of <0.4%, and a core-wide metal-water reaction of <0.1%. The MAPLliGR multiplier for single loop operation (SLO) is 0.77. The SLO multiplier of 0.77 ensures that the PCT for ELO will always be bounded by that of two-loop operation.

NRC approval for single loop operation is documented in Amendment No. 94, dated September 24, 1985, to Cooper Nuclear Station Facility Operating License.

There is one new GE8x8NB fuel design loaded in Cycle 19. The most limiting and least limiting MAPLilGRs for the new fuel designs are as follows:

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

Page 9

\ .

COOPER STATION J11-03354-10 Reload 18 Rev. 0

16. Loss-of-Coolant Accident Results (cont.)

Bundle Type: GE9B-P8DWB350-10GZl-80U-150-T Average Planar Exposure MAPLHGR (kw/ft)

(GWd/ST) (GWd/MT) Most Limiting Least Limiting 0.00 0.00 11.59 11.61 0.20 0.22 11.63 11.65 1.00 1.10 11.71 11.74 2.00 2.20 11.84 11.88 3.00 3.31 11.99 12.03 4.00 4.41 12.14 12.17 5.00 5.51 12.26 12.30 6.00 6.61 12.39 12.43 7.00 7.72 12.53 12.57 8.00 8.82 12.66 12.71 9.00 9.92 12.81 12.87 10.00 11.02 12.85 12.91 12.50 13.78 12.79 12.87 15.00 16.53 12.51 12.54 20.00 22.05 11.78 11.78 25.00 27.56 11.05 11.05 35.00 38.58 9.74 9.75 45.00 49.60 7.92 7.96 49.61 54.68 5.66 5.70 49.68 54.76 --

5.67 Page 10 l

COOPER STATION J11-03354-10 Reload 18 Rev. 0 52 D D D D D D D D 50 E E F H H H H H H F E E 48 E D D F A H A H A A H A H A F D D E 46 D F G H A H A H B F F B H A H A H G F D 44 E F 11 A H H F B H F F H B F H 11 A H F E 42 D H H F H B F F H B F F B H F F B H F H H D 40 D H A H B F B H B F B B F B H B F B H A H D 38 D F A H B F B H B F B H H B F B H B F B H A F D 36 F A H H F B H F F B H F F 11 B F F H B F H H A F 34 E F H A F F H B F F H B F F B H F F B H F F A H F E 32 E H A H B H B F B H B F A A F B H B F B H B H A H E 30 D H H 3 H B F B 11 B F A H H A F B H B F B H B H H D 28 D C A F F F B H F F A H F F H A F F H B F F F A C D 26 D C A F F F B H F F A H F F H A F F H B F F F A C D 24 D H H B H B F B H B F A 11 11 A F B 11 B F B H B H H D 22 E H A H B H B F B H B F A A F B H B F B H B H A H E 20 E F H A F F H B F F H B F F B H F F B H F F A 11 F E

,, 18 F A H H F B H F F B H F F H B F F H B F H H A F 16 j D F A H B F B H B F B H H B F B H B F B H A F D 14 D 11 A H B F B H B F B B F B H B F B H A H D 12 D H H F H B F F H B F F B H F F B H F H 11 D 10 E F H A H H F B H F F H B F H H A H F E 8 f D F G H A H A H B F F B H A H A H G F D 6 f E D D F A H A H A A H A H A F D D E 4 i i E E F H H H H H H F E E l-4 , 1--

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 Fuel Type A=GE98-P8DWB350-10GZ-80U-150-T (Cycle 19) E=GE9B-P8DWB348-12GZ-80M-150-T (Cycle 16)

B=GE9B-P8DWB350-10GZl 80U-150-T (Cycle 19) F=GE9B-P8DWB348-1 IGZ-80M-150-T (Cycle 17)

C=GE9B-P8DWB348-1IGZ-80M-150-T (Cycle 17) G=GE9B-P8DWB348-1 IGZ-80M-150-T (Cycle 18)

D=GE9B-P8DWB348-l lGZ-80M-150-T (Cycle 16) H=GE98-P8DWB350-10GZ-80U-150-T (Cycle 18)

Figure 1 Reference Core Loading Pattern Page11

COOPER STATION J11-03354-10 Reload 18 Rev. O j .

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Page 12

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Page 13 -

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Page 14

COOPER STATION J110335410 Reload 18 Rev. 0 I \

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Page 15

COOPER STATION Jl103354-10 Reicad 18 Rev. 0

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Page 17

COOPER STATION J11-0335410 Reload 18 Rev. 0 1

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Page 18

. . j COOPER STATION Jl1-03354-10 Reload 18 Rev. 0 Appendix A Analysis Conditions To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle.

Table A-1 STANDARD Parameter Analysis Value Thermal power, MWt 2381.0 Core flow, Mlb/hr 73.5 Reactor pressure, psia 1035.0 Inlet enthalpy, BTU /lb 520.4 Non-fuel power fraction 0.038 Steam flow, Mlb/hr 9.56 Dome pressure, psig 1005.0 Turbine pressure, psig 955.1 No. of Safety /Relicf Valves 8 No. of Single Spring Safety Valves 3 Relief mode lowest setpoint, psig 1113.0 Safety mode lowest setpoint, psig 1277.0 Page 19

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COOPER STATION Jl1-03354-10 Reload 18 Rev. 0 Appendix B Decrease in Core Coolant Events The loss-of-feedwater heating (LFWH) and the HPCI inadvertent startup anticipated operational occurrences (AOO) are the only cold water injection events checked on a cycle-to-cycle basis.

The LFWH event was analyzed using the BWR simulator code (Reference B-1). The use of this code is -

permitted in GESTAR II (Reference B-2). The transient plots, flux, and Q/A normally reported in Section 9 are not cutputs of the BWR Simulator Code; therefore, these items are not included in this document for the LFWH event.

s For Cycle 19, the Inadvertent HPCI analysis was shown to be bounded by the LFWH event. This was done by showing the core inlet subcooling due to feedwater temperature reduction from HPCI plus the core inlet subcooling due to excess feedwater from HPCI is less than the core inlet subcooling for the LFWH event.

References B-1. Steady State Nuclear Methods, NEDE-30130-P-A and NEDO-30130-A, April 1985.

B-2. General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-13-US, August 1996.

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COOPER STATION Jl1-03354-10 Reload 18 Rev. O Appendix C SRV Tolerance Analysis The limiting overpressure event for Cooper is the main steam isolation valve closure with flux scram (MSIVF). The Cycle 19 reload evaluation was performed with the SRV and SV opening pressures at 3%

above their nominal values. The peak vessel pressure reported for the Cycle 19 reload is 1246 psig.

An SRV tolerance analysis was previously completed and reported in Reference C-1. To determine the applicability of Reference C-1 results to Cycle 19, an additional MSIVF event was analyzed with SRV opening pressure of 1210 psig (SRV upper limit). Except for the SRV opening pressure, this evaluation used the same analysis conditions as in the standard MSIVF analysis. Figure C-1 shows the reactor response for the MSIVF event with the upper limit SRV opening pressure set to 1210 psig. The peak vessel pressure for this case is 1305 psig at the vessel bottom, which is significantly below the vessel overpressure limit of 1375 psig. Thus, the Cycle 19 upper limit case meets the ASME code requirement for the overpressure protection.

This evaluation demonstrates compliance to vessel overpressure limits for Cycle 19 with the upper limit SRV pressure. Thus, the applicability of Reference C-1 can be extended to Cycle 19.

Reference C-1. SRVSetpoint Tolerance Analysisfor Cooper Nuclear Station, General Electric Company, NEDC-31628P, October 1988.

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COOPER STATION Jil-03354-10

'Reloa'd 18 Rev. 0 Appendix D One Turbine Bypass Out of Service In order to support continued operation of Cooper Nuclear Station with the possibility that one bypass valve is unavailable, the turbine bypass valve (BPV) out of service operation was evaluated. The objective of this evaluation was to calculate the MCPR for the limiting event with one BPV unavailable and determine whether the calculated MCPR specified for the most limiting event for Cycle 19 is affected.

The effect of one BPV unavailable is to reduce the pressure relief capability in the early part of a pressurization event (i.e., before the relief and safety valves can open) and thus result in an increase in the ACPR. The limiting pressurization events that are analyzed on a cycle-specific basis for Cooper are the turbine trip without bypass, the load reject without bypass, and the feedwater controller failure events. The turbine trip without bypass and the load reject without bypass events are not affected by one BPV being unavailable because the analyses do not take credit for any BPV's being available. Therefore, only the feedwater controller failure event (FWCF) was analyzed.

The same conditions that were used for the Cycle 19 reload analysis for the FWCF were used, except that one BPV was assumed to be unavailable. End of Cycle 19 conditions were used as these are most stringent. A conservative representation for the BPV opening characteristic was assumed. Both Option A and, Option B scram conditions were analyzed and the results are provided below. Figure D-1 shows the reactor response for the FWCF event with one BPV unavailable.

With one BPV unavailable, the MCPRs are as follows:

Exposure range: BOC19 to EHFP19 Option A Option B GE8x8NB 1.34 1.26 Page 23 l

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