ML20138G460
| ML20138G460 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/30/1997 |
| From: | Brohaugh T, Reda R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19355F138 | List: |
| References | |
| 24A5399, 24A5399-R, 24A5399-R00, NUDOCS 9705060303 | |
| Download: ML20138G460 (25) | |
Text
__
h, GENuclearEnergy 24A5399 Revision 0 Class I April 1997 Supplemental Reload Licensing Report for COOPER NUCLEAR STATION Reload 17 Cycle 18 l
7 F A888R 8I888L e ~
P PDR
I'D GE Nuclear Energy 24A5399 Revision 0 ClassI April 1997 24A5399, Rev. O Supplemental Reload Licensing Report for Cooper Nuclear Station Reload 17 Cycle 18 Approved b
Approved f
'. J. Reda, Mana er T. R. Brohaugh Fuel and Facility Licensing Fuel Project Manager
- COOPER STATION 24A5399 Reload 17 Rev.0 Important Notice Regarding Contents of This Report Please Read Carefully This repon 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 tme representation of the facts known, obtained or provided to GE at the time this report was prepared.
The only undenakings 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 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 authorized; and with 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.
Page 2
- COOPER STATION 24A5399 Reload 17 Rev.0 Acknowledgement The engineering and reload licensing analyses, which form tir technical basis of this Supplemental Reload Licensing Report, were performed by A. F. Alzaben. The Supplemental Reload Licensing Report was pre-pared by A. F. Alzaben. This document has been verified by D. B. Waltermire of Nuclear Fuel Engineering.
Page 3
- COOPER STATION 24A5399 Reload 17 Rev.0 The basis for this report is General 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.
1.
Plant-unique items Appendix A: Analysis Conditions Appendix B: Decrease in Core Coolant Temperature Events Appendix C: SRV Tolerance Analysis Appendix D: One Turbine Bypass Valve Out of Service 2.
Reload Fuel Bundles Cycle Fuel'I)pe Loaded Number Irradiated:
GE9B-P8DWB320-10GZl-80M-150-T (GE8x8NB) 15 48 GE9B-P8DWB348-llGZ-80M-150-T (GE8x8NB) 16 136 GE9B-P8DWB348-12GZ-80M-150-T (GE8x8NB) 16 48 GE98-P8DWB348-llGZ-80M-150-T (GE8x8NB) 17 148 GE9B-P8DWB348-11GZ-80M-150-T (GE8x8NB) 17 41 Scz GE9B-P8DWB350-10GZ-80U-150-T (GE8x8NB) 18 160 GE9B-P8DWB348-llGZ-80M-150-T (GE8x8NB) 18 4
Total 548 2
3.
Reference Core Loading Pattern 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 core average exposure at beginning of 15342 mwd /MT cycle:
( 13918 mwd /ST)
Assumed reload cycle core average exposure at end of cycle:
26585 mwd /MT
( 24118 mwd /ST)
Reference core loading pattern:
Figure 1
- 1. Re-ir.serted from spent fuel pool (discharged mid-cycle 17 outage).
- 2. The end of cycle core average exposure reflects the basis for the hcense work.
Page 4
- COOPER STATION 24A5399 Reload 17 Rev.0 4.
Calculated Core Effective Multiplication and Control System Worth - No Voids,20 C Beginning of Cycle, kenen,y, 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 Improvement Options Recirculation pump trip:
No Rod withdrawal limiter:
No Thermal power monitor:
No Improved scram t!me:
Yes (ODYN Option B)
Measured scram time:
No Exposure dependent limits:
Yes Exposure points analyzed:
2 (EHFP-2205 mwd /MT, EHFP)
Page 5
l COOPER STATION 24A5399 Reload 17 Rev.0 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 teniperature 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 9.
Core-wide AOO Analysis Results Methods used: GEMIN1; GEXL-PLUS Exposure range: BOC18 to EHFP18-2205 mwd /MT (2000 mwd /ST)
Uncorrected ACPR Event Flux Q/A-GE8x8NB Fig.
(%NBR)
(c/cNBR)
FW Controller Failure 203 114 0.11 2
Turbine Trip w/o Bypass 270 112 0.09 3
Load Reject w/o Bypass 276 112 0.09 4
Exposure range: ElIFP18-2205 mwd /MT (2000 mwd /ST) to EIIFP18 Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.
(%NBR)
(c/cNBR)
FW Controller Failure 275 119 0.15 5
/
Load Reject w/o Bypass 341 116 0.14 6
Tbrbine Trip w/o Bypass 327 11t.
0.14 7
Page 6
- COOPER STATION 24A5399 i
Reload 17 Rev.0 1
- 10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary Rod withdrawal error (RWE) is analyzed in GE Licensing Report, Extended LoadLine Limit and ARTS 1m-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 new fuel design. The results obtained verified that the existing ARTS limits are still valid for this cycle.
- 11. Cycle MCPR Values 3 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, l 996) a cycle-specific Safcty Limit MCPR calculation was performed, and has been reported in both the Safety 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:
1.
The actual 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:1.07 Non-pressurization events:
Exposure Range: HOC 18 to EHFP18 GE8x8NH Loss of 100 *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: HOC 18 to EHFP18-2205 mwd /MT (2000 mwd /ST)
Exposure point: EHFP18-2205 mwd /MT (2000 mwd /ST)
Option A Option H GE8x8NH GE8x8NH FW Controller Failure 1.22 1.20 Turbine 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 hmit is 0.01 greater than the two-loop value.
Page 7
- COOPER STATION 24A5399 Reload 17 Rev.0 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 Turbine Trip w/o Bypass 1.25 1.21
- 12. Overpressurization Analysis Summary Psl Pv Plant Event (psig)
(psig)
Response
MSIV Closure (Flux Scram) 1219 1244 Figure 8
- 13. Loading Error Results Variable water gap misoriented bundle analysis: Yes4 Event ACPR Fuel loading error (Misoriented) 0.14 Fuel loading error (Mislocated) 0.14
- 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-US, March 1991. 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 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 TopicalReport NEDE-24011, Rev. 6. Amendment 8. Thermal Hydraulic Stability Amendment to GESTAR 11, April 24,1985.
Cooper Nuclear Station recognizes the issuance of NRC Bulletin No. 88-4)7, Supplement 1, Power Oscilla-tions in Boiling WaterReactors (BWRs), and has taken appropriate actions to address the identified concerns.
4 includes a 0.00 penalty due to vanable water gap R-factor uncertainty.
l Page 8
' COOPER STATION 24A5399 Relot.d 17 Rev.0
- 16. Loss-of-Coolant Accident Results LOCA method used: SAFF1REFLOOD/ CHASTE Reference the Loss-of-Coolant Accident Analysis Reportfor CooperNuclear Power Station, NEDO-24045, August 1977, as amended.
Page 9
' COOPER STATION 24A5399 Reload 17 Re v. 0
- 16. Loss-of-Coolant Accident Results (cont)
Bundle Type: GE9B-P8DWB350-10GZ-80U-150-T Average Planar Exposure MAPLilGR(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 11.85 11.88 3.00 3.31 12.00 12.03 4.00 4.41 12.13 12.17 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 25.00 27.56 11.05 11.05 35.00 38.58 9.75 9.75 45.00 49.60 7.96 7.96 49.68 54.76 5.67 5.68 49.69 54.78 5.67 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 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 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 10
]
z
- COOPER STATION 24A5399 l
Reload 17 Rev.0
- 16. Loss-of-Coolant Accident Results (cont)
Bundle Type: GE9B-P8DWB348-1 I GZ-80M-150-T (GE8x8NB)
Average Planar Exposure MAPLIIGR(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.I7 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 49.73 54.81 5.83 NOTE:
Peak clad temperatures (ICT) 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 MAPLIIGR 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.
Page 11
~
. COOPER STATION 24A5399 Reload 17 Rev.0
@ @ @ @ @ @b@ @
52 8 8 0 9 fD O+8(D 8 0 8 8 50 mMMMMMMMMHm esHMMMMMMMMis mM M M M M M PsH M M Mm
- - e s M M M M M M M M M M H a s
- M M M M M M M M M M H H H
- &s H M &s M M M s+s s+s M M M M
- MEMs+sMs+sHs+sHMMMM
- -*s M M M M M M M M M M M s*
- *MMMMMMMMMMM*
- sHMMMMMMMMs*
"MMMMM&sMMM*
- sM M &ss**
IIIIIIIi 1 3 5 7 9 11 13 15 17 10 21 23 25 27 29 31 33 35 37 39 91 43 45 97 99 51 Fuel Type A=GE9B P8DWB320-10GZ1-80M-150-T (Cycle 15) E=GE9B-P8DWB348-11GZ-80M-150-T (Cycle 17)
B=GE9B-P8DWB348-11GL80M-150-T (Cycle 17) F=GE9B-P8DWB348-11GZ-80M-150-T (Cycle 18)
C=GE9B-P8DWB348-11GZ-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 Page 12
' COOPER STATION 24A5399 Reload 17 Rev.O Neutron Flux
/
Vessel Press Rise (psi)
\\
- - - - Ave Surface Heatfux
- - - - - Safety Valve Flow 150 0 - --- Core inlet Flow 125.0 - --- Relief Valve Flow rg
(
--- Bypass Valve Flow
- - - Core inlet Subcooli
\\
~
' / "..... ---- -
Q.*I g 100.0
'\\
g 75.0 E
')
E Y
g~
,\\'
f I
i 50 0 N's 25.0
' l' ' j-N
- s
(
'i I
\\
a.....u-0.0
- 25.0 O.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void Reactivity
- - - - - Vessel Stsam Flow
- - - - - Doppler Reactivity 150 0 ---:.-- Turhina Sinam Flow 1.0 - --- Scram Reactivity
- - - Feedwater Flow
--- Total Reactivity e
s e
\\
E g 100.0
--];
0.0
/
T6 l.
E lV E
l' 8
\\
8
~
I,.
2-
\\
'+
b'.,
y
).
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50.0 l,W i:
e 1:. n i.
\\
i b,
g II
\\I l'
h
'I I
O.0
- 2.0 0.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figure 2 Plant Response to FW Controller Failure (HOC 18 to EHFP18-2205 mwd /MT (2000 mwd /ST))
Page 13
' COOPER STATION 24A5399 Reload 17 Rev.O 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
% A I
v 100.0
\\
v 200.0 2
'\\
2 m
N m
E
'N E
\\'g,
8
$0.0 N
'w 100 0
/
\\
{
/
0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Level (inch-BEF-SEP-SKRT)
Void Reactivity
- - - - - Vessel Steam Flow Doppler Reactivity 200.0 - --- Turbine Steam Flow 1.0
-- Scrarn Reactivity
--- Feedwater Flow
-- Total Reactivity 6
.\\
r E
I 8
l t 100 0 p,%
g 0.0 s '~. -
y I '.
' N'. d%..
m og 3
l.
o x
{
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0.0 C l I - " '. - - - - - - - - - - - - -
E-10 i\\
\\.
\\
1
\\'
-100 0 I
l
- 2.0 00 30 60 0.0 30 60 Time (sec)
Time (sec)
Figure 3 Plant Response to Tbrbine Trip w/o Bypass (BOC18 to EIIFP18-2205 mwd /MT (2000 mwd /ST))
Page 14
' COOPER STATION 24A5399 Reload 17 Rev.0 Neutron Flux Vessel Press Rise (psi)
- - Ave Surface Heat Flux
- - - - - Safety Valve Flow 150.0 - -.-
- Core inlet Flow 300.0 - --- Rehef Valve Flow
--- Bypass Vafve Flow
/ %Q-N 100 0
',N 200.0
<a x
co E
{
'N E
t 50.0
<'u 100.0 c------.---__
/
/
/
/
0.0 O.0 O0 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
--- Feedwater Flow
-- Total Reactivity
!E
.\\
t g
~
c t 100.0 m, g 0.0 B
I,
%. 5,.
5 y '. '.
I.
.'w%.-
o s
o g
.1'p.'-.'.------------.
5 -1.0
\\\\
O0
\\
E
\\'
\\
I
\\ ';
1
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-100.0
- 2.0 0.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Figure 4 Plant Response to Load Reject w/o Hypass (HOC 18 to EHFPI8-2205 mwd /MT (2000 mwd /ST))
Page 15
' COOPER STATION 24A5399
_ Reload 17 Rev.0
\\
Vessel Press Rise (psi)
Neutron Flux
/
- - - - - Ave Surface Heat I'ux
- - - - - Safety Valve Flow 150.0 - --- Core inlet Flow 125.0 - --- Relief Valve Flow
(
- -- Core inlet Subcooli-g
\\
--- Bypass Valve Flow
'N,
.s p
M
,\\
75.0 100.0 e
\\
e Q
t e
r~~
V, l
I
$0.0
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{l' ' I 7~s l
L
,N l
I s......t _ _.
0.0
- 25.0 I
OO 20.0 0.0 20.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void Reactivity
- - - - - Vessel Steam Flow
- - - - - Doppler Reactivity Turbine.SteamEJ9w 1.0 150.0 -- m - -
Scram Reactivity
\\
--- Feedwater Flow
--- Total Reactivity n
=
O m
c s
V 100.0
-- li
- 8. 0.0
-l
\\,,, '
e lis I..
E E
l' O
j D
I
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,i s,
2 h
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50.0 ll *,,,.'
e l'd * \\
C ll 1'
\\
II h,
['
II I
h I
0.0
- 2.0 O.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figure 5 Plant Response to FW Controller Failure (EHFPI8-2205 mwd /MT (2000 mwd /ST) to EHFPI8)
Page 16
l
- COOPER STATION 24A5399 i
Reload 17 Rev.O f
Neutron Flux Vessel Press Rise (psi) 1
- - Ave Surface Heat Flux
- - - - - Safety Valve Flow 150.0 - ---
- Core inlet Flow 300.0 - --- Relief Valve Flow I
I
--- Dypaas Valve Flow 1
a 100.0 200.0 E
\\-
E c
v-c
\\',
N.,'. *..
50.0
- --M.
100.0
/
I
/
0.0 0.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT) oid' React j
- - - - Vessel Steam Flow
oppler Reactivity 200.0 - --- Turbine Steam Flow 1.0
- Scram Reactivity
--- Feedwater Flow Total Reactivity j
l
~
t?,
~, -
j m
c o*
C u 1000 m g 0.0 g
"W ': :. -_-., _
e m:
i.
l-4, 'i - 'i - - - - - - - - - - - - -
f -1.0'-
h 0.0
\\ 'l x
\\s'
\\
i
'I I
-100.0
- 2.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec) i Figure 6 Plant Response to Load Reject w/o Bypass (EHFPI8-2205 mwd /MT (2000 mwd /ST) to EHFPI8)
Page 17
' COOPER STATION 24A5399 Reload 17 Rev.0 1
Neutron Flux Vesset Press Rise (psi)
-- Ave Surface Heat Flux
- - - - - Safety Valve Flow 150.0 - ---
- Core inlet Flow 300.0 - --- Relief Valve Flow
--- Bypass Valve Flow 100.0
,g 200.0 w
s.
a m
\\*
x
\\' -
f
~
s
= =..
50.0 24, 100.0 g
/
0.0 O.0 O.0 3.0 6.0 0.0 30 6.0 Time (sec)
Time (sec)
Level (inch.-REF-SEP-SKRT) oid Rea Vessel Steam Flow oppler Reactivity Turbine Steam Flow 1.0 - - - Scram Reactivity 200.0 - ---
--- Feedwater Flow Total Reactivity G
a c
g i
7 100.0 g 0.0
, g p, N a...
w i.
e
- s... -
v.,m o
x s
\\
O (l
'. l
~
\\,
U kl. - ' '- - - - - - - - - - - - - -
M. -1.0 h\\.
0.0 1
\\
\\ I.
s
\\}
-100.0 I
A I
- 2.0 O0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Figure 7 Plant Response to 'Ibrhine 'IYip w/o Bypass (EHFP18-2205 mwd /MT (2000 mwd /ST) to EHFP8)
Page 18
- COOPER STATION 24A5399 Reload 17 Rev.O utron Flux Vessel Press Rise (psi)
A e Surface Heat Flux
- - - - Safety Valve Flow 150.0 - ---.
Inlet Flow 300.0 - --- Relief Valve Flow
--- Bypass Valve Flow 100.0 -- "'
\\
200.0
\\,
m 3
a
,x t
N s
50.0 N
,]
100.0 I
I I
0.0 I
0.0 O.0 4.0 8.0 0.0 4.0 8.0 Time (sec)
Time (sec)
Level (inch REF-SEP-SKRT)
Void Reactivity
- - - - - Vessel Steam Flow
- - - - - Doppler Reactivity 200.0 - --- Turbine Steam Flew 1.0 - --- Scram Reactivity
--- Feedwater Flow
--- Total Reactivity e
y
\\
s
\\.
,s k k00 31NO m3, m
'.sD.,..................-
w
.N e
4' m
g
,~.
~
0.0 W' '
\\
.\\
I3o -1.0
\\
\\
ex
\\
'\\
g
\\
~ 100.0 I
' I
- 2.0 0.0 4.0 8.0 0.0 4.0 8.0 Time (sec)
Time (sec)
Figure 8 Plant Response to MSIV Closure (Flux Scram)
Page 19
1 COOPER STATION 24A5399 Reload 17 Rev.O 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 / Relief Valves 8
No. of Single Spring Safety Valves 3
Relief mode lowest setpoint, psig 1113.0 Safety mode lowest setpoint, psig 1277.0 Page 20
1
' COOPER STATION 24A5399
. +
Reload 17 Rev.0 Appendix B Decrease in Core Coolant Temperature 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). 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.
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-240l1-P-A, February 1991, Page 21
V
' COOPER STATION 24A5399 Reload 17 Rev.0 Appendix C SRV Tolerance Analysis The limiting overpressure event for Cooper is the main steam isolation valve closure with flux scram (MSIVF). The Cycle 18 reload evaluation was performed with the SRV and SV opening pressures at 3%
above their nominal values. The peak vessel pressure reponed 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 Reference 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 psig. Thus, the Cycle 18 Upper limit case meets the ASME code requirement for the overpressure protection.
This evaluation demonstrates compliance to vessel overpressure limits for Cycle 18 with the upper limit SRV pressure. Thus, the applicability of Reference C-1 can be extended to Cycle 18.
Reference C-1. SRV Setpoint Tolerance Analysisfor Cooper Nuclear Station, General Electric Company, NEDC-31628P, October 1988.
I Page 22
' COOPER STATION 24A5399 Reload 17 Rev.O utron Flux Vessel Press Rise (psi)
A Surface Heat Flux
- - - - - Safety Valve Flow 150.0 - ----
Inlet Flow 300.0 - --- Relief Valve Flow
--- Bypass Valve Flow
--"k/
g 100.0
\\
g 200.0
'\\
e m
E
~
g
[
Y
\\p s
g N
50.0 s'
100.0 l
I I
't-0.o O.0 0.0 4.0 8.0 0.0 4.0 8.0 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
--- Feedwater Flow
- - - Tota e etivity G
sg
? ~-~ %.
,'Il,....
k-
',s \\
.N s
N D,, ;,
E
- ss
.h,,.-
g O
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o
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3
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\\
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x
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1 I
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-100 0
-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)
Page 23 l
.s COOPER STATION 24A5399 Reload 17 Rev.O Appendix D One %rbine 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. 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 18 is affected.
The effect of one B PV 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 18 reload analysis for the FWCF were used, except that one BPV was assumed to be unavailable. End of Cycle 18 conditions were used as these are the 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: BOC18 to EHFP18 Option A OptiotLB GE8x8NH 1.27 1.24 f
Page 24
. ' COOPER STATION 24A5399 Reload 17 Rev.0 f
Neutron Flux k
Vessel Press Rise (p
- - - - - Ave Surface Heat Ijux Safety Valve Flow 150.0 - --- Core inlet Flow 125.0 - --- Relief Valve Flow
- -- Core inlet Subcooli-g
--- Bypass Valve Flow
,/
~
U g 100.0 g
75 0 i
I I
E
(,
E l
~
\\'.
1 yl
\\ '.
l N-50.0 Y
25.0 I
I
. +,-. q.__
i I
~
1 J....
m._
0.0
-25.0 I
OO 20.0 0.0 20.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void Reactivity
- - - - - Vessel Steam Flow
Doppler Reactivity 150.0 ---.r -- Irhina Steam Fl9w 1.0 - --- Scram Reactivity
--- Feedwater Flow
--- Total Reactivity 69 e
e u>
E t 100.0
-- 7i S
-.1 0.0 c--
a>%
- l..
E
\\',
- I:,\\ [l',,
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C
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I y
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t D
0.0
- 2.0 I
0.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figure D-1 Plant Response to FW Controller Failum (One 'Ibrbine Hypass Valve Out of Service)
Page 25 1