ML20030C128
| ML20030C128 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/30/1981 |
| From: | Engel R, Hilf C GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20030C123 | List: |
| References | |
| Y1003J01A23, Y1003J1A23, NUDOCS 8108250420 | |
| Download: ML20030C128 (22) | |
Text
...
Y1003J01 A23 REVISION O 1
i CLASSI JUNE 1981 i
i 9~ SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR QUAD CITIES NUCLEAR POWER STATION UNIT 2 RELOAD 5 (CYCLE 6) i l
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4 GENER AL h ELECTRIC isR*!!8!! oks88hkil:
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l Y1003J01A23 Rev. O Class I June 1981 i
SUPPLUtENTAL RELOAD LICENSING SUBMITTAL
~
FOR j
QUAD CITIES NUCLEAR POWER STATION UNIT 2 RELOAD 5 (CYCLE 6) i Prepared:
C. L. Hilf Approved:
R. E. Engel, anager Reload Fuel Licensing i
]
l NUCLEAR POWER SYSTEMS DIVIS ON e GENERAL ELECTRIC COMPANY SAN JOSE, CAllFORNI A 95125 l
GEN ER AL $ ELECTRIC i
Y1003J01A23 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Commonwea..h Edison Company (Edison) for Edison's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending Edison's operating license of the Quad Cities Nuclear Power Station Unit 2.
The information contained in this repot is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided t o General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the contract, as amended, between Common-wealth Edison Company and General Electric Company for Fuel Bundle Fabrication services for Quad Cities Nuclear Power Station Units 1 and 2, dated December 1,1978, as amended, 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 purpoce other than that for which it is intended, is not authorized; and with respect to such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the complete-ness, 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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11
Y 1003J01 A23 Rev. 0 1.
PLANT-UNIQUE ITEMS (1.0)*
a.
Plant parameter changes See Appendix A b.
ODYN Code for Transient Analyses See Appendix B c.
Loading Error See Appendix C d.
Loss-of-Coolant Accident Analysis See Reference 1 (pg 5) e.
R (item 4)
Value shown includes effect of B C settling (0.0005 Ak) 4 f.
Barrier Fuel Bundles See Appendix D g.
RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Type Number Number Drilled Irradiated Reload-l 8DB250 12 0
Reload-2 8DB250 96 0
8DB262 32 0
Reload-3 8DB262 180 0
Reload-4 PSDRB265L 180 180 New Reload-5 P8DRB265L 80 80 PEDGB263L**
32 32 P8DGB298**
32 32 P8DGB284**
68 68 P8DGB265L**
12 12 Total 724 404 3.
REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycle: 18,479 mwd /t Minimum previous cycle core average exposure at end of cycle 18,224 mwd /t from cold shutdown considerations:
Assumed reload cycle core average exposure at end of cycle:
17,374 mwd /t Core loading pattern:
Figure 1.
1 1
- ( ) refers to areas of discussion in " Generic Reload Fuel Application,"
NEDE-240ll-P-A-1, Augus t 1979.
- Barrier Fuel Bundles. See Appendix D.
1 i
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e
Y1003J01A23 Rev. 0 4.
CALCULATED CORE EFFECTIVE FULTIPLICATION AND CONTROL SYSTEM WORTH -
NO VOIDS, 20 C (3.3.2.1.1 AND 3.3.2.1.2)
BOC keff Uncontrolled 1.105 Fully Controlled 0.949 Strongest Control Rod Out 0.979 R, Maximum Increase in Cold Core Reactivity with 0.010 Exposure Into Cycle, Ak 5.
STANDB' LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) 1g31 (20*C, Xenon Free) 600 0.043 6.
RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2) (LFWH Event)
EOC6 Void Coefficient N/A*(c/% Rg)
-5.97/-7.47 Void ion (7l 32.63 oefficient N/A (c/*F)
-0.212/-0.202 ge Fuel Temperature (*F) 1196 Scram Worth N/A ($)
-46.31/-37.05 Scram Reactivity vs Tine Figure 2 i
l 7.
RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 8x8 P8x8R Peaking Factors (local, radial and axial) 1.22 1.20 1.62 1.75 1.40 1.40 R-Factor 1.098 1.051 l
Bundle Power (MWt) 5.505 5.921 Bundle Flow (10 lb/hr) 108.78 106.41 Initial MCPR 1.34 1.36
- N = Nuclear Input Data A = Used in Transient Analysis 2
l m
Y1003J01A23 Rev. 0 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
None 9.
CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Nominal ACPR Power Flow
?
Q/A sl v
py Transient Exposure
(%)
(%)
(%) j]Q_ (psig)
(psig) 8x8 P8x8R Response Load Rejection EOC6 100 100 497 119 1307 1322 0.27 0.29 Figure 3 Without Bypass Loss of 145*F 100 100 121 119 994 1044 0.17 0.17 Figure 4 Feedwater Heating Feedwater EOC6 100 100 192 115 1150 1189 0.14 0.15 Figure 5 Controller Failure 10.
LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(5.2.1)
MLHCR**
0 Rod Block Rod Position Limiting Reading (Feet Withdrawn) 8x8/P8x8R 8x8/P8x8R Rod Pattern 104 3.5 0.17/0.18 13.5/15.2 Figure 6 105 4.0 0.21/0.21 14.1/17.4 Figure 6 106 4.5 0.25/0.23 14.3/18.0 Figure 6 107*
5.0 0.28/0.25 14.5/18.2 Figure 6 108 5.5 0.31/0.27 14.6/18.4 Figure 6 109 6.0 0.33/0.29 14.6/18.4 Figure 6 110 8.5 0.39/0.35 14.6/18.4 Figure 6
- Indicates setpoint selected
- Densification penalty of 2.2% is included 3
_3 w
Y1003J01A23 Rev. 0 11.
CYCLE MCPR VALUES (5.2, APPENDIX C)
BOC6 to EOC6 Option A Option B Pressurization Events 8x8 P8x8R 8x8 P8x8R Generator Load Rejection 1.40 2.42 1.35 1.37 without Bypass Feedwater Controller Failure 1.26 1.27 1.19 1.20 Non-Pressurization Events 8x8 P8x8R Loss of 145'F Feedwater 1.24 1.24 Heating Fuel Loading Error 1.22 1.22 Rod Withdrawal Error 1.35 1.32 12.
OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Flcw sl v
(%)
(%)
(psig)
(psig)
Plant Response MSIV Closure 100 100 1326 1343 Figure 7 (Flux Scram) 13.
STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio: Figure 8 i
Reactore Core Stability:
l Decay Ratio, x /*0 0.53 2
(Natural Circulation / Extrapolated Rod Block)
Channel Hydrodynamic Performance Decay Ratio, x /*0 2
(Natural Circulation /
l Extrapolated Rod Block)
P8x8R channel 0.17 l
8x8 channel 0.27 4
L
Y1003J01A23 Rev. 0 14.
LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
See Reference 1.
15.
LOADING ERROR RESULTS (5.5.4)
See Appendix C.
~
16.
CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Maximum incremental control rod worth:
0.42% Ak
REFERENCES:
(1) " Loss-of-Coolant Accident Analysis Report for Dresden Units 2, 3 and Quad Cities Units 1, 2 Nuclear Power Stations," NED0-24146A, April 1979.
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- 12 OF THESE FUEL BUNDLES WILL BE 8ARRIER FUEL P80G82e6L Figure 1.
Reference Core Loading Pattern f
6
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Y1003J01A23 Rev. 0 100 50 C - 878 CRD IN PERCENT 1 - NOMINAL SCRAM CURVE IN (-$)
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NOTES:
i 1.
NLMBER INDICATES NUMBER OF NOTCHES WITHDRAWN l
OUT OF 48.
BLANK IS A WITHDRAWN ROD.
i 2.
ERROR ROD IS (18,35).
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i Figure 6.
Limiting RWE Rod Pattern l
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Decay Ratio l
l 13/14
Y1003J01A23 Rev. O APPENDIX A PLANT PARAMETER CHANGES Pressure Relief Systems (Table 5-4, pg 5-62, NEDO-24011)
Safety / Relief Valve setpoint (psig) 1115 + 1%
Safety / Relief Valve capacity (% rated steam flow) 29.2 Safety Valve capacity (% rated steam flow) 52.5 Transient Operating Parameters (Table 5-6, pg 5-o4, NEDO-24011)
Turbine Pressure (psig) 950 GETAB Initial Conditions Reactor Core Pressure (psia) 1035 Inlet Enthalpy (Btu /lb) 523.7 Non-Fuel Power Fraction 0.04 ATWS - RPT initiated at 1250 psig 4
4 l
l I
15/16 m
Y1003J01A23 Rev. O APPENDIX B ODYN TRANSIENT CODE All rapid pressurization and overpressure protection events have been analyzed using the ODYN transient code as speaified in Reference B-1.
Code overpres-sure protection analysis results are deterministic as discussed in Reference B-2.
The ACPR values given for the pressurization events in Section 9 are the plant-specific deterministic values calculated by ODYN based on the ini-tial MCPR given in Item 7 of this submittal. These ACPRs may be adjusted to reflect either Option A or Optien B ACPRs by employing the conversion method described in ' Reference B-2.
These adjustments are based on conservatism factors applied to the ratio ArPR/ICPR. The MCPR for the event is determined l
by adding the ACPR to the safety limit. Section 11 presents both the MCPRs for the nonpressurization events, as well as the adjusted MCPRs (Option A and i
Option B) for the pressurization events.
s The operating limit MCPR is the maximum MCPR of the following events:
(1) turbine trip or load rejection without bypass based on ODYN; 4
(2) feedwater controller failure event based on ODYN; (3) loss of feedwater heating event; (4) rod withdrawal error event; (5) bundle loading error accident; i
(6) minimum cequired by LOCA; and (7) minimum required by Reference B-3, Appendix C, Page C-65
)
where Items 3 through 7 are calculated as described in Reference B-3 but the MCPRs for the pressurization events analyzed with ODYN have been adjusted as follows:
17 i
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,y,r-yw m n,n-
~.,e-.e
-.,,.,,,,,,e,,,--_
i Y1003J01A23 Rev. 0 (1) MCPRs are adjusted for Option B for all plants choosing to operate l
under Option B which meet all scram specifications given in Reference B-4.
(2) MCPRs are determined by a linear interpolation between the Option A MCPR and the Option B MCPR for all plants choosing to operate under l
Option B which do not meet the scram time specificat on.
This interpolation is based on the tested measured scram time and is j
{
described in Reference B-4.
REFERENCES i
B-1.
Letter, R. P. Denise (NRC) to G. G. Sherwood (GE), January 23, 1980.
B-2.
Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment Method for Determination of Operating Limits", January 19, 1981.
B-3.
" Generic Reload Fuel Application", NEDE-24011-P-A-1, August 1979.
B-4.
Letter (with attachment), R. H. Buchholz (GE) to P. S. Check (NRC),
" Response to NRC Request for Information on ODYN Computer Model",
September 5, 1980.
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I 18 i
Y1003J01A23 Rev. O APPENDIX C BUNDLE LOADING ERROR EVENT ANALYSES The bendle loading error analyses, protzdures, and results for the rotated bundle are presented below. The mislocated bundle loading error event analy-sis is no longer being reported as discussed in Reference C-2.
C.1 ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error event analysis results presented in this supplement are based on the new analysis procedure described and approved in Reference C-1.
This new method of performing the analysis is based on a more accurate detailed analytical model.
The principal difference between the previous analysis procedure and the new analysis procedure is the modeling of the water gap along the axial length of the bundle. The pitvious analysis used a uniform water gap, whereas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interfacing between the top guide and the fuel spacer buttons, caused by misorientation, causes the bundle to lean. The effect of the variable water gap is to reduce the power peaking and the R-factor in the upper regions of the limiting fuel rod.
This results in the calculation of a reduced CPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry.
The results of the analysis indicate that the limiting event is a rotated P8DGB284 bundle resulting in a 17.5 kW/f t LHGR and a 0.15 ACPR (includes a 0.02 penalty due to variable water gap R-factor uncertainty) with a minimum CPR of >1.07.
The LHGR value includes a 2.'% power peaking penalty due to fuel densification.
19
Y1003J01A23 Rev. O REFERENCES C-1.
Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel (GE). MFN-200-78, dated May 8, 1978.
C-2.
Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Change in General Electric Methods for Analysis of Mislocated Bundle Accident",
November 14, 1980.
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1 20
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_m Y1003J01A23 Rev. 0 APPENDIX D The reload contains barrier fuel bundles for the barrier fuel demonstration i
j program. The bundles and the program are described in liEDO-24259-A, " Generic Information for Barrier Fuel Demonstration Bundle Licensing", February 1981.
4 j
{
Fuel types P8DGB263L, P8DGB298, and P8DGB284 refer, respectively, to the j
2.65% (lower gadolinia), 2.97%, and 2.83% average enrichment bundles described in CE report NEDO-24259-A.
i i
Fuel type P8DGB265L is a barrier fuel bundle of a standard nuclear design which is descrfJed in NEDE-24011-P-A-1, " Generic Reload Fuel Application".
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