ML20094H428
| ML20094H428 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/30/1995 |
| From: | Hetzel W, Klapproth J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20094H423 | List: |
| References | |
| 24A5187, 24A5187-R01, 24A5187-R1, NUDOCS 9511140172 | |
| Download: ML20094H428 (21) | |
Text
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i O
GENuclearEnergy CompanyProprietaryInformation 24A5187 Revision 1 ClassI November 1995 t
(
SUPPLEMENTAL RELOAD LICENSING REPORT for COOPER NUCLEAR STATION RELOAD 16 CYCLE 17
,~
i
.i D
A O
GE Nuclear Energy 24A5187 Revision 1 Class I N
November 1995 S
24A5187, Rev.1
(
Supplemental Reload Licensing Report for Cooper Nuclear Station Reload 16 Cycle 17 Approved d-7 Approved J. F. Klapproth, r
W. H. Hetze Fuel and Facility Licensing Fuel Project Manager I
COOPER STATION 24A5187 Reload 16 Rev.1 Important Notice Regarding Contents of This Report Please Read Carefully V
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 tme 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 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 (ex-pressed or implied) as to the completeness, accuracy or usefulness of the infonnation 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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Page 2
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COOPER STATION 24A5187 Reload 16 Rev.1 Aclinowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by A. F. Alzaben. The Supplemental Reload Licensing Report was pre-pared by A. E Alzaben. This document has been reviewed by W. E. Russell of Fuel Engineering and C. W.
9 Smith of Fuel L.icensing.
e 9
Page 3
COOPER STATION 24A5187 Reload 16 Rev.1 The basis for this report is General Electric Standanl Applicationfor Reactor Fuel, NEDE-240ll-P-A-10, Febmary 1991; and the U.S. Supplement, NEDE-24011-P-A-10-US, March 1991, 1.
Plant-unique Items Appendix A: AnalysisConditions Appendix B: Decrease in Core Coolant Temperature Events Appendix C: SRV Tolerance Analysis 2.
Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:
GE9B-P8DWB302-10GZ-80M-150-T (GE8x8NB) 14 48 GE9B-P8DWB320-10GZl-80M-150-T (GE8x8NB) 15 164 GE9B-P8DWB 348-i l GZ-80M-150-T (GE8x8NB) 16 136 GE9B-P8DWB348-12GZ-80M-150-T (GE8x8NB) 16 48 Erm GE9B-P8 DWB 348-11 GZ-80M-150-T (GE8x8NB) 17 152 4
Total 548 1
3.
Reference Core Loading Pattern Assumed previous cycle core average exposure at end of cycle:
24717 mwd /MT
( 22423 mwd /ST)
Assumed reload cycle core 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 I
- 1. The End of cycle core average exposure reflects the basis for the license work.
Page 4
COOPER STATION 24A5187 Reload 16 Rev.1 4.
Calculated Core Effective Multiplication and Control System Worth - No Voids,20 C Beginning of Cycle, kerreaive Uncontrolled 1.109 Fully controlled 0.965 Strongest control rod out 0.990 g
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 0.037 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 lb/hr)
GE8x8NB 1.20 1.78 1.40 1.000 7.522 99.8 1.14 Exposure: EHFP17-2205 mwd /MT to EHFPI'7 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt)
(1000 lb/hr)
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 withdrawal limiter:
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)
Page 5 1
COOPER STATION 24A5187 Reload 16 Rev.I 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 Turbine bypass system out of service:
No 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; GEXL-PLUS Exposure range: BOC17 to EHFP17-2205 mwd /MT Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.
(7eNBR)
(7eNBR)
FW Controller Failure 187 112 0.08 2
Turbine 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.
('/cNBR)
(c/eNBR)
FW Controller Failure 209 116 0.12 5
Turbine Trip w/o Bypass 288 115 0.12 6
Load lieject w/o Bypass 300 115 0.12 7
Page 6
COOPER STATION 24A5187 Reload 16 Rev.1
- 10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary Rod withdrawal enor (RWE) is analyzed in GE Licensing Report, Extendedload Line Limit and ARTSIm-provement Program Analysesfor CooperNuclearStation Cycle 14, NEDC-31892P, Januasy 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. The results obtained verified that the existing ARTS limits are still valid for this cycle.
- 11. Cycle MCPR Values 2 Safety limit:
1.06 Single loop operation safety limit:
1.07 Non-pressurization events:
Exposure range: BOC17 to EHFP17 Event GE8x8NB Loss of 100 *F feedwater heating 1.15 Inadvertent HPCI 1.22 Fuel loading error (Misoriented) 1.22 Fuel loading error (MislocateQ l.22 Rod withdrawal error (for RBM setpoint to 108%)
1.19 Pressurization events:
Exposure range: BOC17 to EHFP17-2205 mwd /MT Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.20 1.18 Turbine Trip w/o Bypass 1.22 1.15 Load Reject w/o Bypass 1.22 1.15 Exposure range: EHFP17-2205 MWD /MT to EHFP17 Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.22 1.19 Turbine Trip w/o Bypass 1.23 1.19 Load Reject w/o Bypass 1.23 1.19
- 2. For single-loop operation, the MCPR operating limit is 0 01 greater than the two-loop value.
Page 7
COOPER STATION 24A5187 Reload 16 Rev.I
- 12. Overpressurization Analysis Summary Psl Pv Plant Event (psig)
(psig)
Response
MSIV Closure (Flux Scram) 1217 1242 Figure 8
- 13. Loading Error Results Variable water gap misoriented bundle analysis:
Yes Event ACPR Fuelloading error (Misoriented) 0.16 Fuel loading error (Mislocated) 0.16
- 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 it.to 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. Pfefferien (GE), AcceptanceforReferencing ofLicensing TopicalReportNEDE-24011, Rev. 6, Amendment 8 Thermal Hydraulic Stability Amendment to GESTAR H, April 24,1985.
Cooper Nuclear Station recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1, Power Oscilla-tions in Boiling WaterReactors (BWRs), and has taken appropriate actions to address the identified concerns.
- 16. Loss-of-Coolant Accident Results LOCA method used: SAFE /REFLOOD/ CHASTE Reference the loss-of-Coolant A ccident Analysis Reportfor Cooper Nuclea r Power Station, NEDO-24045, August 1977, as amended.
Page 8
COOPER STATION 24A5187 Reload 16 Rev.I
- 16. Loss-of-Coolant Acci6ent Results (cont)
Bundle Type: GE9B-P8DWB348-1IGZ-80M-150-T (GE8x8NB)
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 ')1 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 49.73 54.81 5.83 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 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 (K ) that are applied to the full power nodal t
exposure-<lependent 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 9
COOPER STATION 24A5187 Reload 16 Rev.1 988[D@@de@]O'8BE9
@@b@@
@4@ @
52 so lL eMMMMMMMMMs m!MMMMMMMMMim sHMMMMMMMMMMs lLs MMMMMMMMMMHis
- M M M M M M M M M M M H H
- M M M M M M M M M M M M M
- MEMMMMMMMMMMM
':-*iH M M M M M M M M M Mi*
ll
- HMMMMMMMMMM*
- EMMMMMMMMHi*
- HMMMMMMMM*
- iMMHi"*
IIIIIIII 1 5 5 7 9 11 15 15 17 19 21 25 25 27 29 51 55 55 57 59 il 45 15 if 49 51 Fuel Type A=GE9B-P8DWB302-10GZ-80M-150-T (Cycle 14)
D=GE9B-P8DWB348-12GZ-80M-150-T (Cycle 16)
B=GE9B-P8DWB320-10GZl-80M-150-T (Cycle 15)
E=GE9B-P8DWB348-11GZ-80M-150-T (Cycle 17)
C=GE9B-P8DWB348-11GZ-80M-150-T (Cycle 16)
Figure 1 Reference Core Loading Pattern Page 10
COOPER STATION 24A5187 Reload 16 Rev.I
[\\
Vessel Press Rise (psi)
Neutron Flux
- - - - - Ave Surface Heat Flux
\\
- - - - - Safety Valve Flow 150.0 - --- Core inlet Flow 125.0 - --- Relief Valve Flow
- -- Core Inlet Subcooling
--- Bypass Valve Flow J
./.s.-
E.100.0 s\\
E 75.0 n
.\\
n m
'\\
x
~
\\
50.0 25.0 j
l
__ l...
I I
0.0
- 25.0 0.0 20.0 0.0 20.0 Time (SOC)
Time (SOC)
Level (inch-DEF-SEP-SKRT)
Void Reactivity
- - - - - Vessel Steam Flow
- - - - - Doppler Reactivity 150.0 - --- Turbine Steam Ftow 1.0 - --- Scram Reactivity
--- Feedwater Flow
--- Total Reactivity
(
5
(,
e 100.0 --
e.0
-- ew n-,- -y, r -
l m
e l-c I:
g i
ll l,
3 l
- E
\\'
50.0 l.',.,'
N -1.0 i
e h$
I' t
l' i
I.
K 1,
III I
0.0
- 2.0 0.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figure 2 Plant Response to FW Controller Failure (BOC17 to EHFP17-2205 mwd /MT)
Page 11
COOPER STATION 24A5187 Reload 16 Rev.I
-1 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
~
/ R 's 200.0 g 100.0 5
'.\\
W E
- N, E
~
'\\.,
~
l N'-
50.0 D:
100.0 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)
Void Resctivity
- - - - - Vessel Steam Flow
- - - - Doppler Reactivity 200.0 - --- Turbine Steam Flow 1.0
-- Scram Reactivity
--- Feedwater Flow
-- Total Reactivity G
E g
g 100.0 g, g Oh
.t.
\\
,~
~
Q' g
\\(
~
0.0 1_. l '__.'___..__ _________
R -1.0 t '\\
\\
\\,
gt l}
\\.
l
-100.0
- 2.0 O.0 3.0 6.0 0.0 3.0 60 Time (sec)
Time (sec)
Figure 3 Plant Response to 'Ibrbine 'IYip w/o Bypass (BOC17 to EHFP17-2205 mwd /MT)
I I
l l
Page 12 i
COOPER STATION 24A5187 Reload 16 Rev.1 Neutron Flux Vessel Press Rise (psi)
-- - Ave Surface Heat Flux
- - - - - Safety Valve Flow 150.0 - - - - Core Irdet Flow 300.0 - --- Relief Valve Flow
--- Dypass Valve Flow
/ 'N ).6 y 100.0
/
'. \\
200.0
'\\
E onE
.\\
E 1
\\
m Y
~
\\.
Y N'.
50.0
' ~:. %
100.0 f---------..-
/
/
I
/
I 0.0 O.0 O.0 30 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void Reactiv'ty
-a y
- - - - - Vessel Steam Flow
- - - Doppler Reactivity 200.0 - --- Turbine Steam Flow 1.0
-- Scram Reacti ity
--- Feedwater Flow
-- TotalReactivity Q
l1\\
r E
C v 100.0 m 9 0.0
- f. l 1 ' ' .~.~~. ' N' l r--- --%
_, e l '.
(
l '.1~~.*'
y o
l.
O
- I g-g 31 e
00 b,4-----------------
m -1.0
\\\\
e
\\
\\.
]
\\\\
\\.
-100.0 I
- 2.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Figure 4 Plant Response to Load Reject w/o Bypass (HOC 17 to EHFP17-2205 mwd /MT)
Page 13
COOPER STATION 24A5187 Reload 16 Rev.1 f\\
Neutron Flux Vessel Press Rise (psi)
- - - - - Ave Surface Heat Flux
- - - - Safety Valve Flow 150.0 - --- Core inlet Flow
/
125.0 - --- Relief Valve Flow
- -- Core inlet Subcooting
- - - Bypass Valve Flow J
./*,.
1
~~~~
g 75.0 g 100.0 a
w C
E Y
L Y
~~
l
\\'9 50.0 25.0 j
_L. '
l l
.I _I...
I I
0.0
- 25.0 0.0 20.0 0.0 20.0 Time (sec)
Time (sec)
I Level (inch-REF-SEP-SKRT)
Void Reactrvity k
- - - - - Vessel Steam Flow
- - - - - Doppler Reactivity 150.0 - --- Turbine Steam Flow 1.0 - --- Scram Reactivity h
--- Feedwater Flow
--- Total Reactivity g
I w
y
--}.\\
s 0.0
, m -:w-;.-- e
/
g 100.0 E
l' o
C I.
O g.
- ? '\\ '.(,.
i
.n 50.0 lll v li -1.0
(
e i
i'8 \\
i.
l I:
i li
\\
l I I' 0.0
- 2.0 O.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figum 5 Plant Response to FW Controller Failure (EHFP17-2205 mwd /MT to EHFP17)
Page 14
COOPER STATION 24A5187 Reload 16 Rev.1 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 g 100.0 %'.,'
,N g 200.0
\\(.
~
w v
w e
=
Y 50.0 s'% ',.'.* g N
100.0
/
~
I 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) oid R
- - - - - Vessel Steam Flow r Reactivity 200.0 - --- Turbine Steam Flow 1.0 Scram Reactivity
--- Feedwater Flow Total Reactivity b
,h
,, ~~~
s C
e*
g p 100.0 p, N - d..
g 0.0 le E
N '.
E
~P..*s,__
E g
g e
8 i e g
\\
l
.2-
\\
0.0 C
- 1.0
,\\
e
\\
\\.\\g
\\l e
I
'I I
-100.0
- 2.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Figu m 6 Plant Response to 'Ibrbine ' hip w/o Bypass (EHFP17-2205 mwd /MT to EHFP17)
Page 15
COOPER STATION 24A5187 Reload 16 Rev.1 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
/ N h%
v 100.0
,\\
v 200.0 2
'\\
2 m
e g's,
E c
Y s,.,
50.0 x ' '2 '. c.
100.0
/
I I
0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (soc)
Level (inch-REF-SEP-SKRT) oid Reacts
- - - - - Vessel Steam Flow ppler Reactivity 200.0 - --- Turbine Steam Flow 1.0 Scram Reactivity
--- Feedwater Flow Total Reactivity E
,q 2
I g
/
t 100.0 m, g 0.0 r,
'~---,..A,...,,_
' -**~
g:
g
\\.
~
m -1.0
\\*
M;', b - ~~ - - - - - - - - - - - - -
u 0.0
\\\\
.e t
\\,-
l
+
I
'I '
I
_100.0
- 2.0 O.0 3.0 6.0 00 3.0 6.0 Time (sec)
Time (sec)
Figure 7 Plant Response to Load Reject w/o Bypass (EHFP17-2205 mwd /MT to EHFPl7)
Page 16
a.
u2.
h-.m J~-l a-
.J.
2--e-44at... -+a
.-4_m.
A A--...+.s.4 m+-a,-au*e
+-- #.
.L_.
s4a
.A.,
4
..u
.a COOPER STATION 24A5187 Reload 16 Rev.I
. I i
1 Nsutron Flux Vessel Press Rise (psi)
A'e Surface Heat Flux
- - - - Safety Valve Flow
. O e inlet Flow 300.0 - --- Relief Valve Flow 150.0
--- Bypass Valve Flow Or\\ '.
200.0 100.0
\\
w w
E C
g
~
Y K
N 50.0 s'%-
100.0 I
j I
0.0 0.0 I
I O.0 4.0 8.0 0.0 4.0 8.0 i
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 -
I Reactivity
--- Feedwater Flow Q
a
~
m
\\
g 0.0 m\\ g g 100.0 m
' s g,, N %. ------ - -
8
\\"
E N.\\.
1ll
' N, E
g s
\\
.y g
0.0 as -1.0 e
o E
d
\\\\
I I
-100.0
- 2.0 0.0 4.0 8.0 0.0 4.0 8.0 l
l Time (sec)
Time (sec)
Figure 8 Plant Response to MSIV Closure (Flux Scram) l Page 17 l
COOPER STATION 24A5187 Reload 16 Rev.1 Appendix A Analysis Conditions To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle.
e Table A-1 l
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 analysis, Mlb/hr 9.56 Dome pressure, psig 1005.0 Turbine pressure, psig 955.1 No. of Safety / Relief Valves 8
q No. of Single Spring Safety Valves 3
i Relief mode lowest setpoint, psig 1113.0 Safety mode lowest setpoint, psig 1277.0 I
l Page 18
COOPER STATION 24A5187 Reload 16 Rev.1 Appendix B g
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 L FWH event.
For the HPCI event, the CPR is presented in Section 11. The transient analysis inputs used for the HPCI AOO are given in Table B-1.
Table 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.
g B-2 General Electric Standard Application for Reactor Fuel, NEDE-240l1-P-A, February 1991.
N= Nuclear input data: A=Used in transient analysis.
Generic exposure-independent values are used in Gener.1 Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-10, Febmary 1991.
Page 19
- COOPER STATION 24A5187 Reload 16 Rev.1 Appendix C SRV Tolerance Analysis The limiting overpressure event for Cooper is the main steam isolation valve closure with flux scram (MSIVF). The Cycle 17 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 17 reload is 1242 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 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 the MSIVF event with the upper limit SRV opening pressure set to 1210 psig. The peak vessel pressure for this case is 1302 psig at the vessel bottom, which is significantly below the vessel overpressure limit of 1375 psig. Thus, the Cycle 17 Upper limit case meets the ASME code requirement for the overpressure protection.
This evaluation demonstrate compliance to vessel overpressure limits for cycle 17 with the upper limit SRV pressure. Thus, the applicability of Reference C-1 can be extended to Cycle 17.
Reference C-1. SRV Setpoint Tolerance Analysisfor Cooper Nuclear Station, General Electric Company, NEDC-31628P, October 1988.
l 4
Page 20
COOPER STATION 24A5187 Reload 16 Rev.1 Nputron Flux Vessel Press Rise (psi)
A'e Surface Heat Flux
- - - - - Safety Valve Flow 150.0 - - - - -. O re Inlet Flow 300.0 - --- Relief Valve Flow
--- Bypass Valve Flow
.e
/,b 100.0
\\
200.0 8
8 e
T C
~
Y
~
\\p Y
~
g N
50.0 s*-~.,
100.0
'~~
e--________
I g
I 0.0 l
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 Flow 1.0 - --- Scram Reactivity
--- Feedwater Flow
'vity 2
/
0.0
] 100.0 m\\
g
', q<;',;
s..,...-
o
\\
,e g
0.0 M'l * *
- I
\\ 1 s
ca -1.0
\\
E
\\\\
\\
I
- 2.0 I
-100.0
^
0.0 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) r Page 21
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