ML20117K332
| ML20117K332 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/30/1996 |
| From: | Brohaugh T, Reda R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20117K327 | List: |
| References | |
| 24A5187, 24A5187-R03, 24A5187-R3, NUDOCS 9606110120 | |
| Download: ML20117K332 (23) | |
Text
.- -
Attschusnt 2 e
, [O[ g to NLS960113 T
GE Nuclear Energy 24A5187 Revision 3 Class I
-l June 1996 i
4 l
24A5187, Rev. 3
)
Supplemental Reload Licensing Report 2
for Cooper Nuclear Station Reload 16 Cycle 17 1
l l
l i
Approved 1
Approved R. J. Reda, M ger T. R. Brohaugh Fuel and Facility Licensing Fuel Project Manager 9606110120 960606 PDR ADOCK 05000290 P
COOPER STATION 24A518}
Reload 16 Rev.
Important Notice Regarding Contents of This Report Please Read Carefully
' Ibis 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. 'Ihe information contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GB at the time this report was prepared.
'Ibe 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 this document shall be construed as changmg said 1
contract. 'Ihe use of this information except as defined by said contract, or for any purpose other l
than that for which it is id-tad. is not authorized; and with respect to any such unauthorized use, neither GE nor a ey of the contributors to this document rmkes any representation or warranty (ex-pressed or implied) as to the comphe accuracy or usefulness of the information contained in this Mment 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.
Page 2
COOPER STATION 24A5187 Reload 16 Rev. 3 Acknowledgement ne engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by A. E Alzaben. The Supplemental Reload Licensing Report was pre-pared by A. E Alzaben. His tevision of this document (Rev. 3) has been verified by G. N. Marrotte of Fuel.
Engineering.
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Page 3
COOPER STATION 24A5187 Reload 16 Rev. 3 Tbe basis for this report is General Electric Standard Applicationfor Reac:or Fuel, NEDE-24011-P-A-11, November 1995; and the U.S. Supplement, NEDB-24011-P-A-11-US, November 1995, 1.
Plant-unique Items Appendix A: Analysis Conditions Appendix B: Decrease in Core Coolant Temperature Events Appendix C: SRV Tolerance Analysis Appendix D: One habine Bypass Valve Out of Service 2.
Reload FuelBundles Q&
FuelType Loaded Number Irradiated:
GE9B-P8DWB302-10GZ-80M-150-T (GE8x8NB) 14 48 GE9B-P8DWB320-10GZ1-80M-150-T (GE8x8NB) 15 164 GE9B-P8DWB348-1107,-80M-150-T (GE8x8NB) 16 136 GE9B-P8DWB348-120Z-80M-150-T (GE8x8NB) 16 48 Scm GE9B-P8DWB348-11GZ-80M-150-T (GE8x8NB) 17 152 Total 548 3.
Reference Cort Imding Pattern!
Assumed previous cycle core average exposure at end of cycle:
24717 mwd /MT (22423 mwd /ST)
Assumed reload cycle cose average exposure at beginning of 15417 mwd /MT cycle:
( 13986 mwd /ST)
Assumed reload cycle core average exposure at end of cycle:
25888 mwd /MT
( 23486 mwd /ST)
Reference core loading pattern:
Figure 1
- 1. The end of cycle core averase exposwe reGects the basis for the liocese work.
Page 4
COOPER STATION 24A5187 Rev. 3 Reload 16 4.
Calculated Cort Effective Multiplication and Control System Worth - No Voids,20'C Beginning of Cycle, ken cev.
Uncontrolled 1.109 Fully controlled 0.%5 Strongest control rod out 0.990 1 Maximum increase in cold core reactivity with exposure into cycle, Ak 0.000 5.
Standby Liquid Control System Shutdown Capability Boron Shutdown Margm(Ak)
(ppm)
(20*C, Xenon Free) 660 0.037 6.
Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters Exposure: BOC17 to EHFP17-2205 mwd /MT Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt)
(1000 livkr)
GE8x8NB 1.20 1.78 1.40 1.000 7.522 99.8 1.14 Exposun: EHFP17-2205 mwd /MT to EHFP17 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt)
(1000 Ildhr)
GE8x8NB 1.20 1.72 1.40 1.000 7.290 101.1 1.18 7.
Selected Margin Improvement Options Recirculation pump trip:
No Rod withdrawallimiter:
No Thermal power monitor:
No improved scram time:
Yes (ODYN Option B)
Exposure dependent limits:
Yes Exposure points analyzed:
2 (EHFP-2205 mwd /MT, EHFP) l Page 5
COOPER STATION 24A518]
Rev.
Reload 16 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 Ttubine bypass system out of service:
No One turbine bypass valve out of service:
Yes Safety / relief valves out of service:
No Feetwater heaters out of service:
No ADS out of service:
No 9.
Core-wide AOO Analysis Results Methods used: GEMINI; GEXI<-PLUS Exposure range: BOC17 to EHFP17-2205 mwd /MT Uncorncted ACPR Event Mux Q/A GE8x8NB Fig.
(%NBR)
(%NBR)
FW Controller Failure 187 112 0.08 2
Bubine Trip w/o Bypass 276 111 0.07 3
Load Reject w/o Bypass 285 111 0.07 4
Exposure range: EHFP17-2205 mwd /MT to EHFP17 Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.
(%NBR)
(%NBR)
FW Controller Failure 209 116 0.12 5
Turbine tip w/o Bypass 288 115 0.12 6
Load Reject w/o Bypass 300 115 0.12 7
Page 6
COOPER STATION 24A5187 Reload 16 Rev.3 i
- 10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary Rod withdrawal error (RWE) is analyzed in GE Licensing Report, Extended load Une Umit and ARTS Im.
provement Program Analysesfor CooperNuclearStation Cycle 14, NEDC-31892P, January 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 Cycle 17 fuel design. He results obtained verified that the existing ARTS limits are still valid f* this cycle.
2 l
11, Cycle MCPR Values In agreement with commitments to the NRC (letter from M. A. Smith to the Document Control Desk,10CFR Part 21, Reportable Conditlon, Safety Dmit MCPR Evaluation, May 24,19%) a cycle-spectiic Safety Limit MCPR calculation was performed, and has been reported in both the Safety Limit MCPR and Operating Limit MCPR shown below. His cycle specific SLMCPR was determined using the analysis basis documented in j
GESTAR with the following exceptions:
- 1. De actual core loading was analyzed.
4 l
- 2. He actual bundle parameters (e.g., local peaking) were used.
~
- 3. He full cycle exposure range was analyzed, i
Safety limit:
1.07 Singleloop operation safety limit:
1.08 1
Non-pressurization events:
l Exposure range: BOC17 to EHFP17 Event GE8x8NB Loss of 100 'F feedwater heating 1.16 Inadvertent HPCI 1.23 1.23 Fuelloading error (Misoriented) 1.23 Fuelloading error (Mislocated)
Rod withdrawal error (for RBM setpoint to 108%)
1.20 Pressurization events:
Exposure range: BOC17 to EHFP17-2205 mwd /MT Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.21 1.19 Turbine Trip w/o Bypass 1.23 1.16 1.23 1.16 Load Reject w/o Bypass
- 2. For single-loop operanon. the MCPR operating limit is 0.01 gttater than the two-loop value.
Page 7
l
~. _ _
_ -~
COOPER STATION 24A5187 Reload 16 Rev.3 Exposure range: EHFP17-2205 MWD /MT to EHFP17 Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.23 1.20 hrbine Trip w/o Bypass 1.24 1.20 l
Load Reject w/o Bypass 1.24 1.20 t
- 12. Overpressurization Analysis Summwy i
Psi Py Plant
(
Event (psig)
(psig)
Response
MSIV Closure (Flux Scram) 1217 1242 Figure 8
- 13. Loading Ermr Results Variable water gap misoriented bundle analysis:
Yes Event ACPR l
Fuelloadmg error (Misoriented) 0.16 l
Fuelloadmg error (Mislocated) 0.16
- 14. Contml Rod Dmp Analysis Results CooperNuclear 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-US, March 1991. CNS implemented the BPWS into the Rod Worth Minimizer (RWM) l as h=nted in License AmeMment No.117. Removal of the Rod Sequence Control System (RSCS) at l
CNS has been approved by the NRC in License Amendment No.156.
- 15. Stability Analysis Results GE SIL-380 recommeMations have been included in the Cooper Nuclear Station Technical Specifications; therefore, no stability analysis is required as documented in the letter,C. O. nomas (NRC) to H. C. Pfefferten (GB), Acceptancefor Referencing ofIlcensing TopicalReport NEDE-24011, Rev. 6, Amendment 8, Thenna!
l Hydraulic Stability Amendment to GESTAR 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 (BWRs), and has taken appropriate actions to address the identified concems.
- 16. Loss-of-Coolant Accident Results LOCA method used: SAFE /REFLOOD/ CHASTE i
Reference the Exss-of-Coolant Accident Analysis Reportfor CooperNuclear Power Station, NEDO-24045, August 1977, as amended.
Page 8
i
^
j COOPER STATION 24A5187 Rev. 3 Relo,ad 16
- 16. Loss-of-Coolant Accident Results (cont)
Bundle Type: GE9B-P8DWB348-11GZ-80M-150-T (GE8x8NB)
Average Planar Em e MAPLHGR(kW/ft)
(GWd/ST)
(GWd/MT)
Most Limiting Least Limiting I
O.00 0.00 10.85 11.82 j
0.20 0.22 10.90 11.87 1.00 1.10 11.01 11.%
2.00 2.20 11.17 12.08 3.00 331 1136 12.19 4.00 4.41 11.56 1232 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 1238 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 f
20.00 22.05 11.79 11.91 l
25.00 I
27.56 11.05 11.17 35.00 38.58 9.69 9.74 45.00 49.60 7.86 8.09 l
49.56 54.63 5.62 5.91 5.90 49.59 54.66 5.85 l
49.68 54.76 49.73 54.81 5.83 I
NOTE:
Peak clad temperature (PCT) are s 2127 'F at all exposure and local oxidation fractions are s 0.065 at all exposures.
When in single loop operation, a MAPLHOR 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 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, i
Page 9
ha IgSTATION 24g5]
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- -sMMBEMBEMMMMMMm
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- MMMMMMMMMMME8E ll:MBEE8MMBEMMMMMME8 ll:MiGMBEMBEBEBEMMMMM
- -
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FuelType B-P8D 3h OGZ -Sek WT Cyc E9B-P8 WBWil
- S' Cy le I c-GE98-P8Dwanticz-80u-twt (cycle 16)
Figure 1 Reference Core Loading Pattern Page 10
-~. -.- -
COOPER STATION 24A5187 l,
Reload 16 Rev. 3 i
Neutron Fkm
['\\
Vessel Press Aloe (pel)
\\
- - - - - Safety h FW
..... Aw Surfeos Heat Fkm 150.0 - --- Core inlet Flow 125.0 - --- Rollet Velve Flow
- -- CoreinletSuboooung
--- Bypees Velve Flow
/./.J
--_ f--.....----1 P{\\
l 100.0
.g 75.0
'\\
/
'.\\
~
~
\\
50.0 25.0 l
1 1
I
- l...
I I
0.0
- 25.0 0.0 20.0 0.0 20.0 l
Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void ReactMty
..... Vessel Steam Flow
- - - - - Doppler ReactMty 150.0 - --- Turtaine Steam Flow 1.0 - --- Scram Reecevity
--- Feedwater Flow
--- Total ReectMty g
l
\\
w f,
iQ T.\\
p 0.0 w n w.-;- =' e 100.0
~
l'. \\
j
~
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i-a:
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o i
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- 1.0 50.0 e>
t, IN i
~
l' I,
l 1
K t.
NI hl 0.0
- 2.0 0.0 20.0 0.0 20.0 Time (sec) mme (sec)
Figure 2 Plant Response to FW Controller Failure (BOC17 to EHFP17-2205 mwd /MT) 1 4
i Page 11 1
i COOPER STATION 24A5187 l
Relo-d 16 Rev. 3 t
Neutron Fkm Vessel Prus Fuse (pel)
- Aw Surface Heat Fkm
- - - - - Safety VsNe Flow Core inlet Flow 300.0 - --- Renef V No Flow 150.0
--- sypees v N. now I R' '
m 100.0
,\\
200.0 3
\\
w
.s e
s cc Y
\\~.
Y 60.0 ED:,
100.0
(-----.---.,_
,1 I
/
I 0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 Eme(sec)
Vme (sec)
L.wi0nch-REF-SEP-SKm)
- k[VoidRe=My
- - - Doppiw Reecevity
- - - - - Vessel Steam Flow 200.0 - --- Tuit* ste=n now 1.0
-- scram R.=ey
--- FeooweterFio.
-- ToisiReecevity a
- (\\
0.0
- f. j g 100.0 p, %
,_. t.v ~ -,- -- -- m l.
\\l,V**
ux
' 5 e
\\
\\
e ',
[
\\I 0.0 CM-'-------------
as -1.0 I
\\
W
\\.
\\
\\
\\.
I I
-100.0
- 2.0 O.0 3.0 6.0 00 3.0 6.0 Time (sec)
Time (sec) i Figure 3 Plant Response to'Ihrbine'IYip w/o Bypass l
(BOC17 to EHFP17-2205 mwd /MT) l i
I I
Page 12
COOPER STATION 24A5187 Beload 16 Rev. 3 l
l i
l Neutmn Flux Vessel Preu fuse (pel)
- - Ave Surface Heat Flux
- - - - - Safety Valve Flow j
- core inlet Flow 300.0 - --- Relief Wye Flow 150.0 - - -
--- Bypees Wye Flow
/ #N p' %
/
\\
'.\\
h
.0 200
] 100.0 e
.\\
e E
\\
E Y
~
Y l
50.0 S'%
a 4
100.0 y---~~~---.
I
/
I
/
I 0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 Eme (sec)
Eme(sec) i l
- \\ # old ReactMty
\\
vV t.evel(inciv-REF-SEP-SKRT)
..... Venel Steam Flow
- - - - Doppler ReactMty 200.0 - --- Turt*w Steam Flow 1.0
-- Scram ReactMty
--- Feedwater Flow
-- TotalReactMty
,.. **'~..
0.0 t
a n
,, y w._
_.m
~ il',\\
g 100.0 m.
m' y
N\\
0.0 Mwb-
-10
\\
/
~~-------
e E
\\.
l(
l)
\\.
l
-100.0
- 2.0 O.0 3.0 6.0 0.0 3.0 6.0 Eme (sec)
Eme (sec)
Figure 4 Plant Response to Load Reject w/o Bypass (BOC17 to EHFP17-2205 mwd /MT) l i
Page 13
COOPER STATION 24A5187 Reload 16 Rev. 3 4
Neutron Flux
/ k Vessel Press Rise (pel)
- - - - Ave Surface Heat Flux
- - - Safety Velve Flow 150.0 - --- Core inlet Flow l
125.0 - --- Relief Velve Flow
- -- Coreini.tsubcooiing
--- sypees verve Fio.
J
./",.
- ~~~~
'I 30 75
] 100.0 i
}
Y
~
\\
a m
L Y
^
Y, I.
50.0 25.0 lI...
I I
0.0
- 25.0 0.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Level (Irch-REF-SEP-SKITT)
Void ReactMty
\\
..... Vessel Stoem Flow
- - - - - Doppler Reecuvity
--- Turbine Steam Flow 1.0 - --- Scram ReactMty k
150.0
--- F.edwei.e Fio.
--- Toisi R.acimiy 7
g i
i I
e
- m-r:e;=
',/
100.0 0.0 r
- 1. \\
L mg 13, 1
l{
50.0
/ ;k',v,.
f,.
- 1.0 l.p
' I' E
I I'7 k
~
I.
1 l
1; li
\\
l II I I' 0.0
- 2.0 O.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figurt 5 Plant Response to FW Controller Failure (EHFP17-2205 mwd /MT to EHFP17)
Page 14
COOPER STATION l
Reload 16 24A5187 Rev.3
--- NeutronFlux Vessel Press Rise (pel)
- - Ave Surface Heat Flux
- - - - Safety Valve Flow 150.0 - --
Core inlet Flow 300.0
- - ReRef Valve Flow
--- Bypass Valve Flow Y
A 100
.N 200.0 I
.0 s.
a 8
K*N, '~.., '.,.. ' t 60.0 g
100.0
/
I l
0.0 O.0 0.0 3.0 6.0 0.0 3.0 6.0 Vme (sec)
Time (sec)
Level (inch.4EF-SEP-SKRT) old
.... - Vessel Steam Flow Reecovity 200.0
--- TurtWne Steam Flow 1.0 Scram Reecevtty
--- Feedwater Flow Total ReactMty
.\\'.
m:,
1 100.0 l
i h 0.0 y'..
' ~'
I
'~
h
- '~~
N', '
P ~.:-a s m, _
a l
\\
O.0 Cb1-
- 1.0 s\\
e 1 ')
\\
\\
-100.0
-2.0
'I I
O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Eme (sec)
Figure 6 Plant Response to Thrbine 'IYip w/o Bypass (EHFP17-2205 mwd /MT to EHFP17)
I i
Page 15
COOPER STATION 24A5187 Reload 16 Rev. 3 j
Neutron Flux Vasel Prem Rise (pel)
Ave t nfaoe Heat nux
- - - - Safety h Flow 150.0 - -
- Core inlet Flow 300.0 - --- ReRef h Flow
--- sypees vow now
'T N 100.0 200.0 a
sq.,
m s.
t
\\s.
50.0 c% s *., * :,.
100.0 e---------
/
I I
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) old Vessel Steam Flow Reecevity
--- Turtine Steam Flow 1.0 Scram ReeclMty 200.0
--- Feedwater Flow TotalRoadMty E
n I.
100.0 0.0 r,
m'%
s s
~ y \\ **~
~~ T ~ A s -
g M'
f
\\\\
Nl, $ ~ ~~ - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ -
- 1.0
~
0.0 e
s, I'
l.
I
'I '
I
-100.0
- 2.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Figure 7 Plant Response to Load Reject w/o Bypass (EHFP17-2205 mwd /MT to EHFP17)
=
Page 16
COOPER STATION 24A5187 Reload 16 Rev. 3 N e Flux Vessel Press Rise (peQ A e W Heat Flux
- - - - - W W Flow
. o e iniet Fio.
300.0 - --- Reser veno Fio.
150.0 - ---
--- Bypees Veno Flow
,k
/
200.0
\\
10 100 7.g 3
e
~
N s s0.0 N'-..
100.0 l
g Il 0.0 O.0 O.0 4.0 8.0 0.0 4.0 8.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-FART)
Void ReacthMy
..... Vessel Shm Flow
- - - - - Doppsw Reecevtty 200.0 - --- Turke Steam Flow 1.0
--- Scram Reecevity ReactMiy
--- Feedwater Flow
/
s
\\.
,o*
0.0 g 100.0 s,,,
', N 3't. sN----.---.
\\'
g g
A 0.0 N
- 1.0 g
e
)\\
\\
' I
- 2.0
-100.0 O.0 4.0 8.0 0.0 4.0 8.0 Time (sec)
Time (sec)
Figum 8 Plant Response to MSIV Closure (Flux Scram)
Page 17
COOPER STATION 24A5187 Reload 16 Rev. 3 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.,
l Analysis Value Thermal power, MWt 2381.0
'dore flow, Mlb/hr 73.5 Reactor pressure, psia 1035.0 Inlet enthalpy, B'IU/lb 520.4 l
Non-fuelpower fraction 0.038 Steam flow analysis,Mlb/hr 9.56 Dome pressure,psig 1005.0
'Ihrbine pressure, psig 955.1 No.of Safety / Relief Valves 8
No. of Single Spring Safety Valves 3
Relief modelowest setpoint, psig 1113.0 Safety modelowest setpoint, psig 1277.0 Page 18
COOPER STATION 24A5187 Reload 16 Rev. 3
' Appendix B Decrease in Core Coolant ihmperaturc Events 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-cycle basis.
The LFWH event was analyzed using the BWR Simulator code (Reference B-1). 'Ihe use of this code is per-mitted in GESTAR H (Reference B-2). h 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 the HPCI event, the CPR is presented in Section 11.h transient analysis inputs used for the HPCI AOO - -- - - - -
are given in Table B-1.
1 hble B-1 Void fraction (%)
43.66 Average fuel temperature (*F) 1099 Void coefficient N/A* (c/%RG)
-8.14/-10.18 Doppler coefficient N/A* (c/*F)
-0.191/-0.181 Scram worth N/A* ($)
References B-1. Steady state Nuclear Methods, NEDE-30130-P-A and NEDO-03130-A, April 1985.
B-2 General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A, February 1991.
N= Nuclear input data: A=Used in transient analysis.
Generic exposure-mdependent values are used in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-10, February 1991.
Page 19
COOPER STATION 24A5187 Reload 16 Rev. 3 Appendix C SRV Tolerance Analysis l
De limiting overpressure event for Cooper is the main steam isolation valve closure with flux scram j
(MSIVF). De Cycle 17 reload evaluation was performed with the SRV and SV opening pressures at 3%
above their nominal values. De peak vessel pressure reported for the Cycle 17 reload is 1242 psig.
l An SRV tolerance analysis was previously completed and reported in Reference C-1. To demonstrate the i
applicability of Reference C-1 results to Cycle 17, 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 j
the MSIVF event with the upper limit SRV opening pressure set to 121'O psig. He peak vessel pressure for i
this case is 1302 psig at the vessel bottom, which is significantly below the vessel overpressure limit of 1375 psig. Rus, the Cycle 17 Upper limit case meets the ASME code requirement for the overpressure protection.
a His evaluation demonstrate compliance to vessel overpressure limits for cycle 17 with the upper limit SRV pressure. Bus, the applicability of Reference C-1 can be extended to Cycle 17.
2 i
l Reference C-1. SRVSetpoint Tolerance Analysisfor Cooper Nuclear Station, General Electric Company, NEDC-3162SP, October 1988.
i Page 20
COOPER STATION 24A5187 Reload 16 Rev. 3 N sutron Flux Vessel Press Rise (psi)
Ae Surface Heat Flux
- - - - - Safety Velve Flow l
150.0 - __
. O e inlet Flow 300.0 - --- ReHof Valve Flow
--- Bypass Valve Flow a*.
a
"/,%
's g 100.0
\\
g 200.0
\\
3
~e m
E CE Y
\\p%s Y
~
~.
50.0 100.0 l
I I
0.0 I
'.0 0.0 O.0 4.0 8.0 0.0 4
8.0 Time (seci; Time (sec)
Level (irxh-REF-SEP-SKITT)
Void ReactMty
..... Vessel Steam Flow
- - - - - Doppler ReactMty 200.0 - ---- Turbine Steam Flow 1.0 - --- Scram ReactMty
_ _. Foodw ter F'ow y
E
\\
{1 On e~
n' s.
s.x.-...-
g*d.
dl-g 0.0 M.l..-
I. -i.0
\\,,..
g
\\ 1 s _ __ _ _ _._ _._..
g E
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I I
-100.0
- 2.0 O.0 4.0 8.0 0.0 4.0 8.0 Time (sec)
Time (sec)
Figure C-1 Plant Plesponse to MSIV Closure (Flux Scram)
(SRV Tolerance Analysis) l Page 21 i
_,.. =.. - -
COOPER STATION 24A5187 R, eload 16 Rev. 3 Appendix D One Thrbine Bypass Valve Out of Service l
in order to support continued operation of Cooper Nuclear Station with the possibility that two bypass valves may be available, 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 17 is affected.
'Ihe effect ofone BPV unavailable is to reduce the pressure reliefcapability 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. 'Ibe limiting pressurization events that are analyzed on a cycle-specific basis for Cooper are the turbine trip w/o l
bypass, the load reject w/o bypass, and the feedwater controller failure events. 'Ibe turbine trip w/o bypass, the load reject w/o bypass 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.
1 The same conditions that were used for the Cycle 17 reload analysis for the FWCF were used, except that one BPV was assumed to be unavailable. End of Cycle 17 conditions were used as these are the most stringent.
A conservative representation for the BPV opening charactenstic was assumed. Both Option A and Option B scram conditions were analyzed and the results are provided below.
With one BPV unavailable, the MCPRs are as follows:
j l
Exposure range: BOC17 to EHFPl7 Option A DRtionl GE8x8NB 1.25 1.22 1
l Page 22 i
'l LIST OF NRC COMMITMENTS ATTACHMENT 3 l
Correspondence No:
NLS960113 The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE COMMITMENT OR OUTAGE j
The District will implement the new operating limits corresponding with the revised SLMCPR values.
This Prior to Startup involves revision of the CNS Core Operating Limits f
he Current Report, and revision of the inputs to the plant process Outage.
computer in accordance with the operating limits specified in Revision 3 to the CNS Supplemental Reload Licensing Report.
The District will transmit the revised Core Operating Prior to Startup Limits Report to the NRC.
from the Current Outage.
1 1
l PROCEDURE NUMBER O.42 l
REVISION NUMBER 1 l
PAGE 9 OF 11 l
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