ML20073L667
| ML20073L667 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/31/1994 |
| From: | Klapproth J, Paone C GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20073L638 | List: |
| References | |
| 23A7244, 23A7244-R, 23A7244-R00, NUDOCS 9410130235 | |
| Download: ML20073L667 (19) | |
Text
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O GE Nuclear Energy 23A7244 Revision 0 Class I May 1994 23A7244, Rev. O Supplemental Reload Licensing Report for River Bend Station Reload 5 Cycle 6 Approved 3 g,(f g
Approved J. F. Klapproth ager C,Paone Fuel Licensing Fuel Project Manager 9410130235 941007 8
PDR ADOCK CGUJ
RIVER BEND 23A7244 Reload 5 -
Rev.0-important Notice Regarding Contents of This Report Please Read Carefully This repon was prepared by General Electric Company (GE) solely for Gulf States Utilities Company (GSU) for GSU's use with the U. S. Nuclear Regulatory Commission (USNRC) for amending GSU's operating li-cense of the River Bend Station. The infom1ation contained in this repon is believed by GE to be an accurate e
and true 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 contract between GSU and GE for fuel bundle fabrication and related services for River Bend Station and nothing contained in this document shall be construed as changing said contract. The use of this information, except as defined by said contracts, by anyone other than GSU for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty (expressed or implied), as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any itsponsibility forliability or damage of any kind which may result from such use of such information.
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. RIVER BEND
,3A7244 Reload 5 Rev.O Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Repon, were performed by P. K. Wu of Advanced Engineering. The Supplemental Reload Licens-ing Repon was prepared by M. E. Harding of Nuclear Fuel-Europe. This document has been verified by C. W. Smith of Fuel Licensing.
4 Page 3
RIVER BEND 23A7244 Reload 5 Rev.O l
The basis for this repon is General Electric Standard Applicationfor Reactor Fuel. NEDE-240 l 1-P-A-t o, February 1991; and the U.S. Supplement. NEDE-240ll-P-A-14US, March 1991.
1.
Plant-unique Items Appendix A: Analysis Conditions Appendix B: Basis for Analysis of Loss-of-feedwater Heater Event 2.
Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:
G E8 B-P8SQB322-80Z-120M4WR-150-T (B S322B) (GE8x8EB) 3 36 GE88-P8SQB322-9GZ-120M-4WR-150-T (BS322C) (GE8x8EB) 3 20 G E8 B-P8 S Q B 333-10GZ-120M-4 WR-150-T (G E8x8 E B) 4 160 G E8 B-P8SQB331-i l GL 120M4WR-15%T (GE8 x8EB) 4 16 GE8 B-P8SQB334-10GZ-120M-4 WR-150-T (G E8x8EB) 5 200 Mem GE8B-P8SQB334-11GZ-120M-4WR-150-T (GE8x8EB) 6 56 GE88-P8SQB334-10GZ2-120M-4WR-150-T (GE8x8EB) 6 136 Total 624 3.
Reference Core Loading Pattern Nominal previous cycle core average exposure at end of cycle:
23494 mwd /MT i
( 21314 mwd /ST)
Minimum prevous cycle core average exposure at end of cycle 22494 mwd /MT from cold shutdown considerations:
( 20406 mwd /ST)
Assumed reload cycle core average exposure at beginning of 13423 mwd /MT cycle:
( 12177 mwd /ST)
Assumed reload cycle core average exposure at end of cycle:
24343 mwd /MT
( 22083 mwd /ST)
Reference core loading pattem:
Figure 1 Page 4
RIVER BEND' 23A7244 Reload 5 Rev.0 4.
Calculated Core Effective Multiplication and Control System Worth - No Voids,20 C Beginning of Cycle, kerrecce Uncontrolled 1.125 i
Fully controlled 0.9568 Strongest control rod out 0.988 R. Maximum increase in cold core reactivity with l
exposure into cycle, Ak 0.000 5.
Standby Liquid Control System Shutdown Capability l
Boron Shutdown Margin (Ak) 1 (pprn)
(20*C, Xenon Free) l 660 0.028 l
6.
Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis
]
Initial Condition Parameters 1
Exposure: BOC6 to EOC6 i
l Peaking Factors l
Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt)
(1000 lb/hr) l GE8x8EB 1.20 1.55 1.40 1.051 7.004 108.5 1.14 j
l Exposure: BOC6 to EOC6 FEEDWATER HEATER OUT OF SERVICE Peaking Factors Fuel Bundle Bundle Initial -
Design Local Radial Axial R-Factor Power Flow MCPR
( M W t)
(1000lb/hr)
GE8x8EB 1.20 1.57 1.40 1.051 7.074 107.4 1.16 I
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RIVER BEND 23A7244 Reload 5 Rev.0 7.
Selected Margin Improvement Options Recirculation pump trip:
Yes Rod withdrawallimiter:
Yes Thermal power monitor:
Yes impmved scram time:
No Measured scram time:
No Exposure dependent limits:
No Exposure points analyzed:
1 8.
Operating Flexibility Options Single-loop operation:
Yes Load line limit:
No Extended load line limit:
No Maximum extended load line limit:
No increased core flow throughout cycle:
No Flow point analyzed:
N/A Increased core flow at EOC:
No Feedwater Heater OOS:
Yes Final feedwater temperature reduction:
No ARTS Program:
No Maximum extended operating domain:
No Moisture separator reheater OOS:
No Turbine bypass system OOS:
No Safety /relicf valves OOS:
No ADS OOS:
No Main steam isolation valves OOS:
No Page 6
RIVER BEND 23A7244 Reload 5 Rev.0 9.
Core-wide AOO Analysis Results!
Methods used: GEMINI; GEXL-PLUS Exposure range: HOC 6 to EOC6 Uncorrected ACPR Event Flux Q/A GE8x8EB Fig.
(%NBR)
(%NBR)
FW Controller Failure 228 109 0.06 2
Load Reject w/o Bypass 312 110 0.07 3
Press. Regulator Failure 145 103 0.03 4
Loss of 100* Feedwater Heating 0)
U) 0.12 0)
Exposure range: BOC6 to EOC6 FEEDWATER HEATER OUT OF SERVICE Uncorrected ACPR Event Flux Q/A GE8x8EB Fig.
(%NBR)
(%NBR)
FW Controller Failure 260 113 0.09 5
- 10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary The generic bounding BWR/6 rod withdrawal error analysis described in NEDE-24011-P-A-US is applied.
l j
- 1. see Appendix B.
l Page 7 l
RIVER BEND 23A7244
. Reload 5 Rev.0 23
- 11. Cycle MCPR Values Safety limit:
1.07 Single loop operation safety I. nit: 1.08 Non-oressurization evt 01s1 Exposure Range: BOC6 to EOC6 GE8x6EB Rod Withdrawal Error 1.18 Loss of 100 F Feedwater Heating 1.19 Fuel Loading Error 1.22 Pressurization events:45 Exposure range: BOC6 to EOC6 Exposure point: EOC6 GE8x8EB FW Controller Failure 1.14 Load Reject w/o Bypass 1.14 Press. Regulator Failure 1.11 Exposure range: BOC6 to EOC6 Feedwater Heaters Out of Service Exposure point: EOC6 GE8x8EB FW Controller Failure 1.17
- 12. Overpressurization Analysis Summary 1
Psi Pv Plant Event (psig)
(psig)
Response
MSIV Closun:(Flux Scram) 1215 1257 Figure 6
- 2. GEMINI ODYN adjustment factors are provided in the letter from J.S. Chamley (GE) to M. w. Fiodges (NRC), GEMINI ODYN Adjust-ment Factors for BWR/6. dated July 6,1987. The hmiung transients for River Bend Station, Cycle 6, are rod withdrawal error and loss of
)
100* F feedwater heaung.
- 3. see letter, J. F. Klapproth (GE) tor R. C. Jones, Jr. (NRC), Rotated Bundle Evaluauon, July 20,1992.
- 5. The FWCF with Feedwater Heaters Out of Service is reported here for informauen only.
Page 8
-RIVER BEND 23A7244 Reload 5 Rev.O 6
- 13. Loading Error Results 7
Variable water gap misoriented bundle analysis: Yes Misoriented Fuel Bundle ACPR G E8 B-P8 SQ B 334-10GZ2-120M-4W R-150-T (GE8 x8 E B) 0.15
- 14. Control Rod Drop Analysis Results River Bend Station is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis is not required. NRC approval is documented in NEDE-24011-P-A-10-US, March 1991.
- 15. Stability Analysis Results GE SIL-380 recommendations have been included in the River Bend Station operating procedures and Tech-nical Specifications: therefore, the stability analysis is not required. NRC approval for deletion of a cycle-speciGc stability analysis is documented in Amendment 8 to NEDE-240ll-P-A-US. River Bend Station recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1, Power Oscillations in Boiling Water Reactors (BWRs), and will comply with the recommendations contained therein.
1
- 6. see letter, J. F. Klapproth (GE) to R. C. Jones, Jr. (NRC), Rotated Bundle Evaluauon, July 20,1992.
- 7. Includes a 0.02 penalty due to vanable water gap R-(actor uncertamty.
Page 9
6
' RIVER BEND 23A7244 Feload 5 Rev.O lt'. Loss-of-coolant Accident Results8 LOCA method used: SAFE /RER.OOD (See River Bend Station Final Safety Analysis Report)
Bundle Type: GE8B-P8SQB334-10G22-120M-4WR-150-T Average Planar Exposure AfAPLHGR(kw/ft)
(GWd/ST)
(GWd/MT)
Most Limiting Least Limiting 0.00 0.00 11.?6 11.88 0.20 0.22 11.42 11.91 1.00 1.10 11.54 12.00 2.00 2.20 11.71 12.13 3.00 3.31 11.89 12.28 4.00 4.41 12.08 12.43 5.00 5.51 12.28 12.60 6.00 6.61 12.48 12.76 7.00 7.72 12.69 12.94 8.00 8.82 12.91 13.11 9.00 9.9'2 13.13 13.29 10.00 11.02 13.34 13.48 12.50 13.78 13.57 13.63 15.00 16.53 13.30 13.31 20.00 22.05 12.63 12.64 25.00 27.56 11.95 11.96 35.00 38.58 10.46 10.52 45.00 49.60 9.08 9.21 50.00 55.12 6.95 7.03 The peak clad temperature (PCT)is s 2150 F at all exposures; the local oxidatiori(fraction)is s 0.073 at all exposures. The MAPLHGR multiplier for single-loop operation (SLO) is 0.84.
- 8. For format explansuon, see letter LS. Chamley (GE) to M. W. Hodps (NRC). Recanrnended MAPtJIGR Technical Specificauons for Muluple Latuce Fuel Designs, March 9.1987. Most Lmitmg and Le.ut brnaung refer to the lowest and highest imuts, respecuvely, of any ennched latuce m the bundle.
Page 10
RIVER BEND 23A7244 Reload 5 Rev.0
- 16. Loss-of-coolant Accident Results (continued)'
LOCA method used: SAFE /REFLOOD(See River Bend Station Final Safety Analysis Report)
Bundle Type: GE8B-P8SQB334-l lGZ-120M-4WR-150-T Average Planar Exposure MAPLHGR(kw/ft)
(GWd/ST)
(GWd/MT)
Most Limiting Least Limiting 0.00 0.00 10.86 11.24 0.20 U.22 10.93 11.30 1.00 1.10 11.09 11.44 1
2.00 2.20 11.30 11.63 3.00 3.31 11.53 11.83 l
4.00 4.41 11.76 12.05 5.00 5.51 12.01 12.27
)
6.00 6.61 12.27 12.50 l
7.00 7.72 12.53 12.74 l
8.00 8.82 12.81 12.98 9.00 9.92 13.09 13.22 10.00 11.02 13.36 13.46 12.50 13.78 13.58 13.61 15.00 16.53 13.29 13.31 20.00 22.05 12.63 12.63 25.00 27.56 11.95 11.96 35.00 38.58 10.46 10.52 45.00 49.60 9.08 9.20 50.00 55.12 6.95 7.02 i
The peak clad temperature (PCT) is s 2150 F at all exposures; the local oxidation (fraction)is s 0.073 at all exposures. The MAPLHGR multiplier for single-loop operation (SLO) is 0.84.
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- 9. For format explanauon, we letter LS. Chamley (GE) to M. W. Halges (NRC), Recommended MAPLHGR Technical Specificauons for l
Muluple tauice Fuel Designs, March 9.1987. Most Lanutmg and Least Lirniung refer to the lowest and highest 1: nuts. respecuvely, of any I
ennched lattxe in the bundle.
l Page 11 l
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4 s.
mememmem mMMMMMMM.
mMMMMMMMMMm 288MMMMMMMMMM.
MMMMMMMMMMMMM ll-mH M M M M M M M M M M M Me i;:8M M M M M M M M M M M M M8
- 8H M M M M M M M M M M M M8 ll: 8 M M M M M M M M M M M M M E l'-* M M M M M M M M M M M M M *
- MMMMMMMMMMMMM "MMMMMMMMMMM"
- MMMMMMMMM*
"MMMMMMM*
1 2
IIIIIilI 4
I 3 $ 7 9 11 13 18 17 le 21 23 N 17 N 31 33 M 37 N 41 43 48 47 44 51 53 SS Fuel Type P8 1
-4
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F E B-P8 4 10GZ2 M-4WR T
fdst:2'iltiRind"ilt?
Figust i Reference Core Loading Pattern Page 12
e '
RIVER BEND 23A7244 a
Reload 5 Rev.O
\\
Vessel Press R2se (pst)
}
Neutron Rux Ave Surface lier: Flux
- - - Safety Valve now 150.0 - ---
Core Inlet Flow
{
125.0 - --- Rehef Valve Flow Core Inle Subcooling
)
Bypass Valve Row N
\\
,p 1 1 I l 100 0
==a' I
75 0
,I I
3 f 1 3
I- [
e
\\,
S l
Qe lI l
i
'. s I
.\\N 25 0 l
l 50.0 I
)
3;
.I.-.-
I I
0.0
-25.0 0.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void Reactivity q
Vessel Steam Row
- - - - - Doppler Reacovity Turbme Sicam Row 1.0 - --- Scrain Reactivity 150.0 Feedwater Row Total Reacuvity m,
3 100.0
- - ' / '
00 t \\
- l 3
\\'
i i
a In ;\\-
e y
t s:
1 50.0 j,,' *\\,.
d -1.0 p
m l
- l..
i,. : \\
x
(
f.-
I e,..::;: \\.
'\\
00
-2.0 0.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figure 2 Plant Response to FW Controller Failure (BOC6 to EOC6) 1 l
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. Page 13
)
RIVER BEND 23A7244 Reload 5 Rev.0
+
Neutron Rux Vessel Press Rise (pa)
Ave Surface Heat Rua
- + * - Safety Valve Row 150 0
- -- Core Irdet Row 300.0 - ---
Relief Valve Row
--- Bypass Valve Row a
b \\
100 30 9'\\
] 200.0 2
e 50.0 100.0 f
~~-I I
\\
l i
l I
\\
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-SKRD Void R crivity
- f Dopple Reactivity Vessel Steam Flow 200.0 - --- Turbine Steam Row 1.0 Scram tivity
--- Feedwater Bow Total Re uvuy g
.,4 m
a M
g 0.0 100.0 e,
y.
. i.
, 1...
g
(
l, u
- s
[.
i.i i
l.
0.0 m, a -,- - - A- - - - - -
d -i.0
\\\\
a g-u
\\\\
\\.
\\\\;
\\
I l
-100.0
-2.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)
Time (sec)
Figure 3 Plant Response to Load Reject w/o Bypass (BOC6 to EOC6)
Page 14
. RIVER BEND 23 A7244 -
4 Reload 5 Rev.O
{
l 3
i f
F O
Neutron Rux Wssel Press Rise (pu)
Ave Surface IIcat Rus
- - - - - Safety Wlve Row q
150 0 - ---
Core Inle Row MO Relief Wlve Row
- 1 Bypass Wlve Flow
,,,,-N
\\
200.0 100.0 e.
b
\\
b a
'.,N e
's-,
e s*s j
50.0 100.0
[
~3 g
\\
\\
g
\\
g
\\
l l
0.0 O.0 I
0.0 5.0 10.0 0.0 5.0 10.0 Time (SCC)
Time (SCC)
I 1.evel(inch-REF-SEP-SKRT)
Void eactivity
..... Vessel Steam Row r Reactivig Turbme Steam Row I.0 - - - Se Reactivity 200.0 Feedwater Flow Total R tmty N
se
.e v
3 (n y
~. c _.
L On y
3....s q,.. -
3 v.
s.
I
\\.
.=
\\
, o a
5
- 1.0 00
~
x
\\
\\
\\
\\
l le j
)
I
-100.0
~2.0 O.0 5.0 10.0 0.0 5.0 10 0 Time (Sec)
Time (Sec)
Figure 4 Plant Response to Pressure Regulator Failure (BOC6 to EOC6) l 1
Page 15 l
p.
RIVER BEND 23A7244 Reload 5 Rev.O l
7' Neutron Flux Vessel Press Ri fe (psi)
- - - Ave Surface licat Rux
- - - Safety Valve Rw 150 0 125 0 - --- Retief Valve R>w
- - - - - - Core Inlet Rt,
--- Core Inlet Su:coohng Bypass Valve 11ou s
} \\
o
- ~ ^
O
, 75 0 I
\\
, 100.0
~- -
3 4
3 l
l S
\\
S I I N'.s l
T l
'. \\
l l
\\
25.0 50 0
'\\
l l
,s
..I L..
'. s ~~
a -.J t
1 l
0.0
-25.0 0.0 20.0 00 20.0 Time (sec)
Time (sec)
Level (inch-REF-4EP-SKRT)
Vod Rescuvtty Vessel Steam Row Doppict Rendutir 150.0 - --- Turbine Steam Row I.0 - --- Scram Reactinty
--- Feedwater Row Toul Reactivi r
.^*
'q y
l G
g o
im 0
- l.
g, 0.0 W i-.-.g ' /
.g g
-c,::
g c
u t
y H\\',.,
e x
s I
c.. -
2 d -1.0 F
Sa0 a
g.. -
x t;',\\;,
L 1:::
R:,'
N:'
0.0
-2.0 O.0 20.0 0.0 20.0 Time (sec)
Time (sec)
Figure 5 Plant Response to FW Controller Failure (BOC6 to EOC6 Feedwater Heater Out of Service)
Page 16 i
)
' RIVER BEND 23A7244 Reload 5 Rev.0 l
l Neutron Rux Vessel Pres
- Ri::'p-)
/,ve Surface Heat Rux Safety Wlve Row I50 0 - ----
Inte: Ro, 300.0 - ---
Rehef Valve Row
- -- Bypass Wlve Flow (y
..V
"',(,,,T '.
/
100.0 200 0 l
y 3
3
'% ' N '.
J j
i N
l 1
'7 %
1 50.0 100.0 l
g- - - - - - - -
l-l/
j
.,'5'-
I I
I 0.0 0.0 O.0 4.0 8.0 0.0 4.0 S.O Time (Sec)
Time (sec) l l
1 Level (inch-REF-SEP-SKRT)
Voi es ty j
Wssel Steam Flow i
Do ler eactivit 200.0 - --- Turbine Steam Row 1.0 - --- Sc Rescuvity Feedwater Row oul Reachvity n
e, j
n
.~
i 1
1
.A
- -.'- a s
- y
- s m\\
0.0 1
100.0 m e'. \\. W.
)
5
.1
'. N :.
j
\\
x g
V i
- \\
\\
\\\\
N 4
0.0
%'l/.*
I, N---------
-1.0 M
\\
i
\\
\\,
\\
\\
I I
-100.0
-2.0 O.0 4.0 8.0 0.0 4.0 8.0 Time (Sec)
Time (sec)
Figure 6 Plant Response to MSIV Closure (Flux Scram) l l
Page 17 l
RIVER BEND 23A7244 0
' Reload 5 Rev.O i
j l
Appendix A Analysis Conditions 1
To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle.
Table A-1 Analysis Value Parameter STANDARD FWHOOS Thermal power, MWt 2894.0 2894.0 Core flow, Mlb/hr 84.5 84.5 Reactor pmssure, psia 1055.0 1055.0 Inlet enthaJpy, BTU /lb 527.9 514.2 Non-fuel power fraction 0.041 0.041 Steam flow analysis, Mlb/hr 12.45 12.45 Dome pressure, psig 1025.0 1025.0 Turbine pressure, psig 986.0 986.0 No. of Dual Mode S/R Valves 16 16 Relief mode lowest setpoint, psig 1133.0 1133.0 Safety mode lowest setpoint, psig 1177.0 1177.0 Safety / Relief Valve Capacity,Ib/hr 831,000 831,000 Reference Pressure, psig 1080 1080 1
Page 18 l
i
fi 3 (,.
3-RIVER BEND 23A7244 4-Reload 5 Rev.0 I
i i
Appendix B Basis for Analysis of Loss-of-feedwater Heater Event The loss-of-feedw ater heating event was analyzed at 102% rated power using the BWR Simulator Code (Reference B-1). Tht, use of this code is permitted in GESTAR II (Reference B-2). The transient plots, neutron flux and heat flux values normally reported in Section 9 are not an output of the BWR Simulator Code; therefore, these items are net included in this document.
Rderences B-1. Steady-State Nuclear Methods NEDE-30134P-A, and NEDO-3013%A, April 1985 B-2. General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A (latest approved version)
I l
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