ML20141B656

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Rev 1 to 24A5399, Suppl Reload Licensing Rept for Cooper Nuclear Station,Reload 17 Cycle 18
ML20141B656
Person / Time
Site: Cooper Entergy icon.png
Issue date: 05/09/1997
From: Brohaugh T, Reda R
GENERAL ELECTRIC CO.
To:
Shared Package
ML20141B653 List:
References
24A5399, 24A5399-R01, 24A5399-R1, NUDOCS 9705150374
Download: ML20141B656 (26)
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] GENuclearEnergy I.;

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+ 24A5399 Revision I ChusI May 1997

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Supplemental Reload Licensing Report for .

COOPER NECLEAR STATION Reload 17 Cycle 18  !

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l-9705150374 970509 I PDR t

P ADOCK 05000298['PDR. 2

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GE Nuclear Energy l

b 24A5399 Resision 1 r Class I May 1997 i

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24A5399,Rev. 1 Supplenental Reload Licensing Report l

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! Cooper Nuclear Station Reload 17 Cycle 18 l

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Approved - S' Approved TT R. . Reda, Manager T. R. Brohaugh Fuel and Facility Licensing Fuel Project Manager k

. COOPER STATION 24A5399 Reload 17 Rev.1 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. Nu: lear 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 t4ue 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 con-tract between NPPD and GE for nuclear fuel and related services for the nuclear system for Coo-per Nuclear Station and nothing contained in thb document shall be constmed 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 authorird; and w ith respect to any such unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty (ex-pressed 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.

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, COOPER STATION 24A5399 Reload 17 Rev.1 Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload .

Licensing Repon, were performed by A. E Alzaben. 'Ihe Supplemental Reload Licensing Report was pre--

pared by A. E Alzaben. This document has been verified by D. B. Waltermire of Nuclear Fuel Engineering.

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. COOPER STATION 24A5399

Reload 17 Rev.I

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The basis for this teport is GeneralElectric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-13, August 1996; and the U.S. Supplement, NEDE-24011-P-A-13-US. August 1996.

1. - Plant-unique Items Appendix A: Analysis Conditions

' Appendix B: Decrease in Core Coolant Temperature Events Appendix C: SRV Tolerance Analysis Appendix D: One 'Ibrbine Bypass Valve Out of Service -

2. Reload Fuel B9ndles Cycle Fuel Type Loaded Number

' Irrndiatad:

GE9B-P8DWB320-10GZ1-80M-150-T (GE8x8NB) 15 48 GE9B-P8DWB348-11GZ-80M-150-T (GE8x8NB) 16 136 GE9B-P8DWB348-12GL80M-150-T (GE8x8NB) 16 48 GE98-P8DWB348-11GZ-80M-150-T (GE8x8NB) 17 148 GE9B-P8DWB348-ilGZ-80M-150-T (GE8x8NB) 17 43 i U GE9B-P8DWB350-10GZ-80U-150-T (GE8x8NB) 18 160 GE9B-P8DWB348-ilGZ-80M-150-T (GE8x8NB) 18 4 Total 548

3. Reference Core Loading Pattern 2

[

Nominal previous cycle core average exposure at end of cycle: 26092 mwd /MT

( 23670 mwd /ST)

Minimum previous cycle core average exposure at end of cycle 25761 mwd /MT from cold shutdown considerations: ( 23370 mwd /ST)

Assumed reload cycle cora average exposure at beginning of 15342 mwd /MT cycle: ( 13918 mwd /ST)

P Assumed reload cycle core average exposure at end of cycle: 26585 mwd /MT

( 24118 mwd /S'I)

,, Reference core loading pattem: Figure 1

1. Re-4nnened from spent fuel pool (discharged mid-cycle 17 outage).

( 2. The end of cycle core average exposure seAcets the basis for the license work.

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. COOPER STATION 24A5399 Reload 17 Rev.1

4. Calculated Core Effective Multiplication and Control System Worth - No Voids,20 C Beginning of Cycle, kerreaive Uncontrolled 1.106 Fully controlled 0.965 Strongest control rod out 0.987 R Maximum increase in cold core reactivity with exposure into cycle, Ak 0.000
5. Standby Liquid Control System Shutdown Capability Boron Shutdown Margin (Ak)

(ppm) (20 C, Xenon Free) 660 0.039

6. Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters Exposure: BOC18 to EHFP18-2205 mwd /MT (2000 mwd /ST)

Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE8x8NB 1.20 1.74 1.40 1.000 7.376 100.7 1.17 Exposure: EHFP18-2205 mwd /MT (2000 mwd /ST) to EHFP18 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE8x8NB 1.20 1.69 1.40 1.000 7.157 102.0 1.21

7. Selected Margin Impruvement 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)

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L, COOPER STATION : 24A5399 Reload 17 Rev.1 I

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8. Operating Flexibility Options Single-loop operation: 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 hrbine 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
9. Core-wide AOO Analysis Results I

Methods used: GEMINI; GEXI -PLUS l

Exposure range: HOC 18 to EHFP18-2205 MWdIMT (2000 mwd /ST)

Uncorrected ACPR l Event Flux Q/A GE8x8NB Fig.

(%NBR) (%NBR)

FW Controller Failure 203 114 0.11 2 hrbine Trip w/o Bypass 270 112 0.09 3 Load Reject w/o Bypass 276 112 0.09 4 Exposure range: EHFP18-2205 mwd /MT (2000 mwd /ST) to EHFP18 Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.

(%NBR) (%NBR)

>. FW Controller Failure 275 119 0.15 5 Load Reject w/o Bypass 341 116 0.14 6 hrbine Trip w/o Bypass 327 116 0.14 7 Page 6

.-  ? COOPER STATION 24A5399 Reload 17 Rev.1

> '10.t Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary

~ Rod withdrawal error (RWE) is analyzed in GE Licensing Report, Extended lead Line Limit and ARTS Im-r ' pmvement Pwgram Analysesfor Cooper Nuclear Station Cycle 14, NEDC .R 892P, Revision 1, May 1991. l 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 Values3 In agreement with commitments to the NRC (letter from M. A. Smith to the Document Control Desk,10CFR Part 21, Reportable Condition, Safety Umit MCPR Evaluation, May 24,1996) a cyc1e-specific Safety Limit MCPR calculation was performed, and has been reported in both the Safety Limit MCPR and Operating Limit l MCPR shown below. His cycle specific SLMCPR was determined using the analysis basis documented in GESTAR with the following exceptions:
1. The actual core loading was analyzed. I
2. The actual bundle parameters (e.g., local peaking) were used.
3. The full cycle exposuir range was analyzed.

Safety limit: 1.06 Single loop operation safety limit:1.07 -

Non-pressurization events:

Exposure Range: BOC18 to EHFP18 GE8x8NB Loss of r) 'F feedwater heating 1.18 Fuel Loading Error (misoriented) 1.20 Fuel Loading Error (mislocated) 1.20 Rod withdrawal error (for RBM setpoint to 108%) 1.19 Pressurization events:

Exposure range: BOC18 to EHFP18-2205 mwd /MT (2000 mwd /ST) j Exposure point: EHFP18-2205 mwd /MT (2000 mwd /ST) '

Option A Option B

.. GE8x8NB GE8x8NB FW Controller Failure 1.22 1.20

., hrbine Trip w/o Bypass 1.24 1.17 Load Reject w/o Bypass 1.24 1.17

)

3. For single-loop operation, the MCPR operating Imut is 0.01 greaser than the two-loop value.

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1 COOPER STATION ' 24A5399 l', Reload 17 Rev.1 L

f Exposure range: EHFP18-2205 mwd /MT (2000 mwd /ST) to EHFP18 Exposure point: EHFP18 Option A Option B

} GE8x8NB GE8x8NB FW Controller Failure 1.25 1.22 Load Reject w/o Bypass 1.25 1.21 Tbrbine Trip w/o Bypass 1.25 1.21

12. Overpressurization Analysis Summary Psi Py Plant l

Event (psig) (psig) Response MSIV Closure (Flux Scram) 1219 1244 Figure 8

13. Loading Error Results i Variable water gap misoriented bundle analysis: Yes 4 Event ACPR Fuel loading error (Misoriented) 0.14 Fuel loading error (Mislocated) 0.14
14. Control Rod Dmp 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-240ll-P-A-13-US, August 1996. CNS implemented the BPWS into the Rod Wonh Minimizer (RWM) as documented in License Amendment No.117. 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 GE SIL-380 recommendations have been included in the Cooper Nuclear Station Technical Specifications; therefore, no stability analysis is required as documented in the letter, C. O. Thomas (NRC) to H. C. Pfefferlen

.. (GE), Acceptancefor Referencing ofLicensing Topical Report NEDE-24011, Rev. 6, Amendment 8, Thennal Hydraulic Stability Amendment to GESIAR 11, April 24,1985.

.. Cooper Nuclear Station recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1, Power Oscilla-tions in Boiling Water Reactors (B WRs), and has taken appropriate actions to address the identified concems.

I 4, includes a 0 00 penalty due to vanable water gap R-factor uncertainty.

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. . COOPER STATION 24A5399 Reload 17 Rev.1

16. Loss-of-Coolant Accident Results LOCA method used: SAFE /REFLOOD/ CHASTE Reicrence the loss-of-Coolant Accident Analysis Reportfor CooperNuclear Power Station, NEDO-24045, August 1977, as amended.

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, COOPER STATION 24A5399 Reload 17 Rev.I i

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

Bundle Type: GE9B-P8DWB350-10GZ-80U-150-T

?

i 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 l 2.00 2.20 -

11.85 11.88 3.00 3.31 12.00 12.03 ,

4.00 4.41 12.13 12.17 I 5.00 5.51 12.26 12.30 6.00 6.61 12.38 12.43 7.00 7.72 12.52 12.57 8.00 8.82 12.65 12.71 ,

9.00 9.92 12.80 12.87 10.00 11.02 12.84 12.91 12.50 13.78 12.81 12.87 15.00 16.53 12.52 12.54 20.00 22.05 11.78 11.78 l

25.00 27.56 11.05 11.05 35.00 38.58 9.75 9.75 45.00 49.60 7.96 7.96 l 49.68 54.76 5.67 5.68 49.69 54,78 -

5.67 l

NOTE:

Peak clad temperatures (PCT) are s 2181 'F at all exposures and local oxidation fractions are s 0.077 at all exposures.

,. When in single loop operation, a MAPLIIGR factor of 0.75 is substituted for the LOCA analysis factors of 1.0 and 0.86 contained in the flow dependent MAPLHGR curves (Kr) that are applied to the full power nodal exposuro-dependent limits.

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

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- , . COOPER STATION 24A5399 l Reload 17 Rev.1 I

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~

k 16. Ims-of-Coolant Accident Results (cont)

Bundle Type: GE9B-P8DWBW l1GZ-80M-150-T(GE8x8NB) e Average Planar Exposure MAPLHGR(kW/ft)

(GWd/ST) (GWd/MT) Most Limiting Least Limiting 0.00 0.00 10.85 11.82 0.20 0.22 10.90 11.87 1.00 1.10 11.01 11.96 2.00 2.20 11.17 12.08 3.00 3.31 11.36 12.19 4.00 4.41 11.56 12.32 5.00 5.51 11.76 12.44 6.00 6.61 11.91 12.55 7.00 7.72 12.07 12.65 8.00 8.82 12.23 12.68 9.00 9.92 12.38 12.67 10.00 11.02 12.48 12.80 12.50 13.78 12.61 12.93 15.00 16.53 12.47 12.60 20.00 22.05 11.79 11.91 25.00 27.56 11.05 11.17 35.00 38.58 9.69 9.74 45.00 49.60 7.86 8.09 49.56 - 54.63 5.62 5.91 49.59 54.66 -

5.90 49.68 54.76 -

5.85 44.73 54.81 -

5.83 NOTE:

Peak clad temperatures (PCT) are s 2127 *F at all exposures and local oxidation fractions are s 0.065 at all exposures.

When in single loop operation, a MAPLHGR factor of 0.75 is substituted for the LOCA analysis factors of 1.0 and 0.86 contained in the flow dependent MAPLHGR curves (Kr) that are applied to the full power nodal 3 exposure-dependent limits.

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

Page 11

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. COOPER STATION 24A5399

' Reload 17 Rev.1

/

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MMMMMMMMMMMMM
-*sM M M M M M M M M M Ms*

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ll *sMMMMMMMMMs*

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**sHMMs**

l l l l l l l l 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 83 35 37 59 41 43 45 97 49 51 Fuel Type A=GE9B-P8DWB320-10GZl-80M-150-T (Cycle 15) E=GE9B-P8DWB348-1IGZ-80M-150-T (Cycle 17)

B=GE9B-P8DWB348-1IG7-80M-150-T (Cycle 17) F=GE98-P8DWB348-1IGZ-80M-150-T (Cycle 18)

C=GE9B-P8DWB348-1IGZ-80M-150-T (Cycle 16) G=GE9B-P8DWB350-10GZ-80U-150-T (Cycle 18) '

,, D=GE9B-P8DWB348-12GZ-80M-150-T (Cycle 16)

Figure 1 Reference Core Loading Pattern _

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. COOPER STATION 24A5399 Reload 17 Rev.1 L

l Vessel Press Rise (psi)

P f4eutron Flux f

- - - - - Ave Surface Heat I'ux 150.0 - --- Core Irdet Flow ,

\ .

- - - - Safety Valve Flow 125.0 - --- Relief Valve Flow

- -- Core Irdet Subcoolrg ( --- Bypass Valve Flow

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g 100.0

@\

  • v 75.0 -

- ',\ S E '\ ' E e \\- *

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N's N

25.0 -

P f 'l- N l ( $

s' l I s..... u_____

' ' ' I 0.0 - 25.0 0.0 20.0 0.0 20.0 Time (sec) Time (sec) l 1

i Level (inch-REF-SEP-SKRT) Void Reactivity

- - - Vessel Steam Flow - - - - - Doppler Reactivity h 150.0 - : .-- Turbine Steam riow 1.0 - --- Scram Reactivity ,,

--- Feedwater Flow - -- Total Reactivity ,

Q

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e m .

N e g 100.0

--~ ~~l. 0 - - ~ ,'

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=

c:n l.; \

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...... i Is

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II lI O0 ' h - 2.0 I 0.0 20.0 0.0 20.0 Time (sec) Time (sec)

Figure 2 Plant Response to FW Controller Failure (BOC18 to EHFP18-2205 mwd /MT (2000 mwd /ST))

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. COOPER STATION 24A5399 l Reload 17 Rev.1 I

Neutron Flux . Vessel Press Rise (psi)

Ave Surface Heat Flux - - - - - Safety Valve Flow 150.0 - ---

- Core inlet Flow 300.0 - --- Relief Valve Flow

--- Bypass Valve Flow l

l 100.0 ,

% A

, ', 200.0 -

. .s .

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$  %., g $

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50.0 - EM , 100.0 -

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0.0 '

O.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

Level (inch-REF-SEP-SKRT) [ Void Reactivity Vessel Steam Flow - - ' Doppler Reactivity 200.0 - --- TurtWne Steam Flow 1.0 -

-- Scram Reactivity

--- Feedwater Flow -- Total Reactivity 6 'q *,.. i 2 ..

p100 m R l '

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-  %. s . ..

E I ,. 5 ,, d - % . ~ m s  :

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0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

'- Figure 3 Plant Response to 'Ibrbine 'lYip w/o Bypass (BOC18 to EHFP18-2205 mwd /MT (2000 mwd /ST))

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i , . COOPER STATION 24A5399 Reload 17 Rev.I l

l l

Neutron Flux Vessel Press Rise (psi)

- - 4

- Ave Surface Heat Flux - - - - - Safety Valve Flow 150.0 - - - - Coro inlet Flow 300.0 - --- Relief Valve Flow

--- Dypass Valve Flow

~

I

/ ',Qss 100.0 ,

, '. \ 200.0 -

m x e E 'N C

%.s- * ~

t 50.0 -

%' ' 4 100.0 -

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' ' / I 0.0 O.0 0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

Level (inch-REF-SEP-SKRT) Void Reactivity

- - - Vessel Steam Flow --

Doppler Reactivity 200,0 .- --- Turbine Steam Flow 1.0 -

-- Scram Reactivity

--- Foodwater Flow -- Total Reactivity E

f\ . ..-

ue *.. -

p 100.0 m, s , {0.0 ,,

s% ~:... -

e E ),', ,,

'W'A-.- 3 0.0 Q'

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g h -1.0 -

\-

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-100.0 I ' '

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O.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

I Figure 4 Plant Response to Load Reject w/o Bypass (BOC18 to EHFP18-2205 mwd /MT (2000 mwd /ST))

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, ., COOPER STATION 24A5399 Reload 17 Rev.1 i

i i

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~

Neutron Flux Vessel Press Rise (psi)

- - - - Ave Surface Heat f'ux '

, - - - - - Safety Valve Flow 150.0 - --- Core Irdet Flow , 125.0 - --- ReEsf Valve Flow

\

- -- Core triot SutWrg \ --- Bypass Valve Flow

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/ .

100.0 *

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8~~----- ... -Lf'),t g 75.0 E \ E

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' ' ' I 0.0 - 25.0 0.0 20.0 0.0 20.0 Time (sec) Time (sec)

? l Level (inch-REF-SEP-SKRT) Void Reactivity V l

- - - - Vessel Steam Flow - - - - - Doppler Reactivity ,

150.0 - - --- Turhina Steam f"1 w 9 1.0 - --- Scram Reactivity ,.-

i

--- Feedweter Flow --- Total Reactivity ,-

n

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-2.0 ' I 0.0 20.0 0.0 20.0 q Time (sec) Time (sec) {

Figum 5 Plant Response to FW Controller Failum (EHFP18-2205 mwd /MT (2000 mwd /ST) to EIIFP18)

Page 16

y .. COOPER STATION 24A5399 Reload 17 Rev.1 l

E

( Neutron Flux Vessel Press Rise (psi) l

- - Ave Surface Heat Flux - - - - Safety Valve Flow

' 150,0 - --- - Core Irdet Flow 300.0 - --- Relief Valve Flow

--- Bypass Valve Flow 100.0 . -

g 200.0

. s. a E \'. C

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N. - $

s.'

50.0 -  ; N* ..

c. 100.0 -

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0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

Level (6nch-REF-SEP-SKRT) oid Reacts

- - - - Vessel Steam Flow -----

pier Reactivity ._ s 200.0 - --- Turbine Steam Flow 1.0 - - - Scram Reactivity

--- Feedwater Flow -

Total Reactivity g ,\ ...

3 \ ,,.

t .

e t .-

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g 100.0 m {0.0 -

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m , ' i' ~we A f

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  • 00 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

Figure 6 Plant Response to Load Reject w/o Hypass (EHFP18-2205 mwd /MT (2000 mwd /ST) to EHFP18)

Page 17

. .. COOPER STATION 24A5399

' Reload 17 Rev.1 i

Neutron Flux Vessel Press Rise (psi)

- Ave Surface Heat Fha - - - - - Safety Valve Flow 150.0 - --- - Core inist Fk= 300.0 - --- Relief Valve Flow

--- Bypass Valve Flow l .

100.0 ,

200.0 -

e <.N., e

/t N. .

50.0 -

% s *% . 4 c. 100.0 -

l I I

' I '

0.0 0.0 l 0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec) l 1- Level (inch-REF-SEP-SKRT) 06d R l - - - - - Vessel Steam Flow -----

r Reactivity -

200.0 - --- Turbine Steam Flow 1.0 - - - Scram Reactivity

--- Feedwater Flow -

Total Reactivity g

2c ,

8 0.0 p 100.0 , ,-

c li; m , ..N d. , v . ,

t.

+ , . , _

g s... .

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1 \

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-100.0 I '

-2.0 \{

'I I 0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

Figure 7 Plant Response to 'Ibrbine Trip w/o Bypass (EHFP18-2205 mwd /MT (2000 mwd /ST) to EHFP8) .l Page 18

. ., COOPER STATION 24A5399 Reload 17 Rev.1 l

l h

hputron Flux Vessel Press Rise (psi)

--- A Surface Heat Flux - - - - Safety Valve Flow 150.0 - ----

Irdet Flow 300.0 - --- Relief Valve Flow

--- Bypass Valve Flow 100.0 - - * " " ' \ -

E

\,I -

C h.200.0 g ,

., 8

%,s .

50.0 - ' *- 100.0 -

I I

I 0.0 - 1 0.0

/ ' >

0.0 4.0 8.0 0.0 4.0 8.0 l

l Time (sec) nme (sec) l l

l f Level (inch-REF-SEP-SKRT) Void Reactivity l - - - - - Vessel Steam Flow - - - - - Doppler Reactivity 200.0 - --- Turbine 6 team Flow 1.0 - --- Scram Reactivity

--- Feedwater Flow --- Total Reactivity g-s

  • l \-

k*,,'.- -

1 EN " q , %,..' .',

l00

& m ,s 16

's -----------..... E .N ,

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.. l Time (sec) Time (sec)

'. Figure 8 Plant Response to MSIV Closure (Flux Scram)

Page 19

. . COOPER STATION . 24A5399 Reload 17 Rev.1 l

Appendix A Analysis Conditions L

To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle.

r 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 ,

'Ibrbine pressure, psig 955.1 No. of Safety / Relief Valves 8 No. of Single Spring Safety Valves 3 Relief mode lowest setpoint, psig 1113.0 Safety mode lowest setpoint, psig 1277.0 l'

t I

i I

Page 20

, ., COOPER STATION 24A5399 Reload 17 Rev.1

! Appendix B Decrease in Core Coolant Temperature Events l

l l

l The loss-of-feedwater heating (LFWH) and the HPCI inadvertent startup anticipated operational occur-rences (AOO) are the only cold water injection events checked on a cycle-by-<ycle basis.

The LFWH event was analyzed using the BWR Simulator code (Reference B-1). The use of this code is per-mitted in GESTAR II(Reference B-2). The transient plots, flux, and Q/A normally reported in Section 9 are not outputs of the BWR Simulator Code; therefore, these items are not included in this document for the LFWH event.

For Cycle 18, 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.

1 Refer.nces l 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 19%. l a

1 i

s.

Page 21

l

. - . , GOOPER STATION 24A5399 l Reload 17 Rev.I L

Appendix C SRV Tolerance' Analysis De limiting overpressure event for Cooper is the main steam isolation valve closure with fiax scram (MSIVF). De Cycle 18 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 18 reload is 1244 psig.

An SRV tolerance analysis was previously completed and reported in Reference C-1. To demonstrate the applicability of Referen~ C-1 results to Cycle 18, 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 1304 psig at the vessel bottom, which is significantly below the vessel overpressure limit of 1375 4 psig. Thus, the Cycle 18 Upper limit case meets the ASME code requirement for the overpressure protection. j This evaluation demonstrates compliance to vessel overpressure limits for Cycle 18 with the upper limit SRV  ;

pressure. Bus, the applicabihty of Reference C-1 can be extended to Cycle 18. {

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

(

)

Page 22

, ,, COOPER STATION 24A5399 Reload 17 Rev.1 i

l l

on Flux Vessel Press Rise (psi)

( -*--

A Surface Heat Flux - - - - Safety Valve Flow 150.0 - ---- Irdet Flow 300.0 - --- Relief Valve %w

--- Bypass Valve Flow

~

100.0 -- -

\ '., 200.0 -

e - e E g '., E Y

~

Y

~

\p ',

h '.,

60.0 - ~~.'% % .

. 100.0 -

l I

I 0.0 1 0.0 ' ' ~~ '

O.0 4.0 8.0 0.0 4.0 8.0 Time (sec) Time (sec)

L ewel(inch-REF--SEP-SKRT) Void Reactivity

- - - - - Vessel Steam Flow - - - - - Doppler Reactivity 200.0 - --- Turtdne Steam Flow 1.0 - --- Scram Reactivity

--- r dwat.r now ---T vity J - 5 '

Je \\. ,.-

8 .-

\;...-...

l E

g

~ 9

,'. g%, ;--

0 h** .

\

I g . .- - . g {.

0.0 -

Q

\ ---------

f

- 1.0 -

\.

l m i.

\

1

-100.0 ' ' l '

- 2.0 O.0 4.0 8.0 0.0 4.0 8.0

, Time (sec) Time (sec)

, Figure C-1 Plant Response to MSIV Closure (Flux Scram)

(SRV Tolerance Analysis) i 1

Page 23

, . , COOPER STATION 24A5399 Reload 17 Rev.1 l

l Appendix D l

One Turbine Bypass Valve Out of Service In order to support continued operation of Cooper Nuclear Station with the possibility that one bypass valve may be unavailable, the turbine bypass valve (BPV) out of service operation was evaluated, ne 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 18 is affected.

The efrect ofone BPV unavailable is to reduce the pressure relief capability in the early part of a pressurization l

event (i.e., before the relief and safety valves can open) and thus result in an increase in the ACPR. The i limiting pressurization events that are atslyzed 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 w;;hout 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.

De same conditions that were used for the Cycle 18 reload analysis for the FWCF were used, except that one f BPV was assumed to be unavailable. End c' Cycle 18 conditions were used as these are the most stringent. j A conservative representation for the BPV opening characteristic was assumed. Both Option A and Option l 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: BOCl8 to EHFPI8 l

Option A Option B l GE8x8NB 1.27 1.24 m.

Page 24

e ,. COOPER STATION 24A5399 Reload 17 Rev.I l

l. Neutron Flux I k Vowel Press Rise l - - - - Ave Surface Heat Ijux - - - - - Safety Valve Flow 150.0 - --- Core inlet Flow 125.0 - -- - Relief Valve Flow l

- -- Core inlet Suthg ,

--- Bypass Valve Flow a ~

~

/ ,

rk 100.0

)

g 75.0 g

= i  % I i E (. C l

  1. w

~

\'. # I

\ ', I I

\- I I 50.0 -

V 25.0 -

i N. s j

, j+-- - q . -

l

1. . . . . . b _

' ' l 0.0 - 25.0 '

0.0 20.0 0.0 20.0 Time (sec) Time (sec) l Level (inch-REF-SEP-SKRT) Void Reactivity 1 - - - - Vessel Steam Flow - - - - - Doppier ReactMty 150.0 1.0 - --- Scram ReactMty j *

-- : -- T'awna Stamm F19w ,-

--- Feedwater Flow --- Total Reactivity { ,

l sc {

i l

{

~ I g 100.0 "--- ]. 0.0 --

g E l, ,g e E l' , '

1

" l

.. 8 \

  1. I;: ':l,\ ,.'.:' '. # I i

.2 50 0 -

1l l ,, ,,

\

k-1.0 -

.,  ; y ,

.g . ll l '.i '.' 's l'Ys\ , I\

l%

  • s'... j' 0.0 '

Is l' \I - 2.0 '

l

' l 0.0 20.0 0.0 20.0

, Time (sec) Eme (sec)

Figure D-1 Plant Response to FW Controller Failure (One 'Ibrbine Bypass Valve Out of Service)

Page 25

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