ML19350A529
| ML19350A529 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/31/1981 |
| From: | Engel R, Zarbis W GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19350A525 | List: |
| References | |
| Y1003J01A17, Y1003J1A17, NUDOCS 8103160428 | |
| Download: ML19350A529 (31) | |
Text
/
l Y1003JCIA17 Class I January 1981 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR l
COOPER NUCLEAR STATION, UNIT 1 1
RELOAD 6 I
s 4
s'.
Prepared:
/
[or
- b. A. Zar s, Licensing Engineer l
Reload Fuel Licensing IJ i
I
\\
t Approved: h[
f R. E. Engel, Manager Reload Fuel Licensing 4
4 1
i NUCLEAR POWER SYSTEMS DIVISICid e GENERAL ELECTRIC COMPANY f
SAN JOSE. CAUFORNI A 95125 4
GENER AL $ ELECTRIC E03/6,o#2r
Y1003J01A17 k
I:2CETAIJ NOT:CE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report vaa prepared by General Electric actsty for Nebraska Public Pouer District (NPPD) for NPPD's use uith the U.S. Nuclear Regulatory Comiacion (USNRC) for amending NPPD's opemting license of the Cooper Nuclear Station. The infomation contained in this report is believed by General Electric to be an accurate and true representation of the facta knoun, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting infomation in this document are contained in the contract betueen Nebraska Public Pouer District and Generet Electric Company for nuclear fuel and related services for the nuclear ayatem for Cooper Nuclear Station and nothing contained in this document shall be construed as changing said contract.
The use of this infoma-tion 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 General Electric Company nor any of the contributors to this document makes any representation or carranty (express or implied) as to the complete-neas, accumcy, or usefulness of the infomation contained in this document or that auch use of such infomation 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
.auch info mation.
FJ
Y1003J01A17 1.
PLANT-UNIQUE ITEMS (1.0)*
Bundle Loading Error Analysis - Appendix A GETAB Analysis Initial Condit! ions - Appendix B Densification Power Spiking - Appendix C Transient Analysis Results - Appendix D Overpressurization Analysis - Appendix E 2.
RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Type Number N_ umber Drilled Irradiated 7DB250 (Initial Core) 48 48 Irradiated 8DB274L (R2) 64 64 Irradiated 8DB274L (R3) 24 24 Irradiated 8DRB283 (R3) 76 76 Irradiated 8DRB283 (R4) 112 112 Irradiated P8DRB283 (RS) 80 80 Irradiated P8DRB265L (RS) 32 32 New P8DRB283 (R6) 76 76 New P8DRB265L (R6) 36 36 Total 548 548 3.
REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure:
16,975 Wd/t Assumed reload cycle exposure:
17,441 Wd/t l
Core loading pattern:
Figure 1 i
l l
- (
) refers to areas of discussion in " Generic Reload Fuel Application,"
NEDE-24011-P-A-1. August 1979.
1
Y1003J01A17 4.,
CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTDI UORTH -
N0 VOIDS, 20'C (3.3.2.1.1 AND 3.3.2.1.2)
BOC k,ff Uncontrolled 1.111 Fully Controlled 0.955 Strongest Control Rod Out 0.986 R, itaximum Increase in Cold Core Reactivity 0.000 wit! Exposure Into Cycle, Ak 5.
STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAI ABILITY (3.3.2.1.3)
Shutdown Margin (Ak) m (20'C, Xenon Free) 600 0.036 6.
RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)
EOC l
Void Coefficient N/A* (c/% Rg)
-7.296/-9.12**
Void Fraction (%)
40.03 Doppler Coefficient N/A (C/%'F)
-0.226/-0.215**
Average Fuel Temperature (*F) 1357 Scram Worth N/A ($)
-37.61/-30.09**
Scram Reactivity vs Time Figure 2**
7.
RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)
Exposure 7x7 8x8/8x8R/P8x8R Peaking factors (. local, (1.24, 1.209, 1.4)
(1.20, 1.544, 1.4) radial and axial)
R-Factor 1.10 1.051 Bundle Power (MWt) 5.489 6.553 Bundle Flow 122.6 110.4 (103 lb/hr)
Initial MCPR' 1.21 1.29
\\
- N = Nuclear Input Data A = Used in Transient Analysis
- Used in non-pressurization transients only.
2
Y1003J01A17 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
None 9.
CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Core h
Q/A p
p Power Flow (%
(%
si v
8x8R/ Plant Transient Exposure (%)
(%) NBR) NBR) (psig) (psig) 7x7 8x8 P8x3R Response Load (See Appendix D)
Rej ection Without Bypass i
Loss of BOC-EOC 104.3 100 123 121.7 1022 1069 0.12 0.14 0.14 Figure 3 100'F Feedwater Heating Feedwater (See Appendix D)
Controller Failure 10.
LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT SLWiARY (5.2.1)*
Rod Block Rod Position Limiting Reading (%)
(Feet Withdrawn) 8x8 8x8R/P8x8R 8x8 8x8R/P8x8R Rod Pattern 104 3.0 D.10 0.07 14.9 16.3 Figure 4 105 3.5 0.16 0.08 15.4 17.0 Figure 4 106**
3.5 0.16 0.08 15.4 17.0 Figure 4 107 4.0 0.20 0.14 15.6 17.5 Figure 4 108 5.5 0.28 0.15 15.6 17.5 Figure 4 109 7.0 0.30 0.19 15.6 17.5 Figure 4 110 8.0 0.30 0.19 15.6 17.5 Figure 4
- Since the 7x7 fuel is on or near the periphery of the core far removed from the error rod, the Rod Withdrawal Error has no impact on this fuel.
-** Indicates setpoint selected.
- Includes densification penalty of 2.2%.
3
Y1003J01A17 11.
CYCLE MCPR VALUES (5.2)
Pressurization Events:
Exposure Transient Option A Option B (7x7/8x8/8x8R/P8x8R)
(7x7/8x8/8x8R/P8x8R)
BOC Load Rejection 1.26/1.32/1.32/1.34 1.16/1.20/1.20/1.22 Without Bypass to.
Feedwater Controller 1.21/1.25/1.26/1.28 1.15/1.19/1.20/1.22 EOC Failure Nonpressurization Eve' ts:
a (7x7/8x8/8x8R/P8x8R)
BOC Loss of Feedwater 1.19/1.21/1.21/1.21 Heating g,
Fuel Loading Error
---/1. 24 /1. 24 /1. 24 E0C Rod Withdrawal Error
---/1. 23 /1.15/1.15 Minimum Required 1.20/1.20/1.20/1.20 by LOCA Minimum Required by
- 1. 2 3 /---/---/---
NRC-SER for Ref. D-3, Appendix D 12.
OVERPRESSURIZATION ANALYSIS SLMfARY (5.3)
(See Appendix E) 13.
STABILITY ANALYSIS RESLT.TS (5.4)
Decay Ratio:
Figure 5 Reactor Core Stability:
Decay Ratio, x /*0 2
(105% Rod Line - Natural Circulation Power) 0.80 Channel Hydrodynamic Performance Decay Ratio, x /*0 2
(105% Rod Line - Natural Circulation Power) 8x8R/P8x8R channel 0.28 8x8 channel 0.37 7x7 channel 0.0 4
Y1003J01A17 1
14.
LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
" Supplemental Reload Licensing Submittal For Cooper Nuclear Station Unit 1 Reload 5", NEDO-24230, December 1979.
15.
CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient:
Figure 6 Accident Reactivity Shape Functions:
Figures 7 and 8 Scram Reactivity Functions:
Figures 9 and 10 Plant specific analysis results Parameter not bounded: Scram Reactivity Function, cold startup Resultant Peak Enthalpy = 154.6 cal /gm 16.
LOADING ERROR RESULTS (5.5.4)
Limiting Event:
Rotated Bundle (P8DRB283)
MCPR:
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FUEL TYPE A = INITIAL CORE TYPE II E = R3 8DRB283 I = R6 P8DRB265L B = INITIAL CORE TYPE III F = R4 8DRB283 J = R6 P8DRB283 i
l C = R2 8DB274L G = R5 P8DRB265L l D = R3 8DB274L H = R5 P8DRB283 Figure 1.
Reference Core Loading Pattern P00RORSINAI.
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Scram Reactivity and Control Rod Drive Specifications 7
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v-
Y1003J0LA17 i
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Doppler Reactivity Coefficient Comparison for RDA 11
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Y1003J01A17 APPENDIX A BUNDLE LOADING ACCIDENT The loading error accident (nislocated bundle and rotated bundle) for Cooper Cycle 7 has been analyzed using the " revised methods" (Reference A-1).
These analysis methods have been previously applied to Cooper (References A-2 and A-2) and these applications approved by the NRC (References A-4 and A-5).
The loading error will ese influence the operating CPR limit. The most severe MLHGR predicted for a misloaded bundle is 17.7 kWf t.
This value l
includes an allowance of 2.2% for power spiking due to fuel densification i
(see Appendix C).
REFERENCES A-1 Letter, D. G. Eisenhut (NRC) to R. E. Engel (GE), transmitting Safety Evaluation Report on "new calculational procedures...for the fuel bundle loading error analyses", May 8, 1978.
A-2.
Letter, J. Pilant (NPPD) to George Lear (NRC), April 14, 1978.
l-A-3 Letter, J. Pilant (NPPD) to T. A. Ippolito (NRC), August 16, 1978.
A-4 Letter, George Lear (NRC) to J. M. Pilant (NPPD), May 2,1978.
A-5 Letter, T. A. Ippolito (NRC) to J. Pilant (NPPD), August 25, 1978.
P 17/18
Y1003J0Lt17 APPENDIX B I
GETAB INITIAL CONDITIONS Table 5-8 of Reference B-1 states the "Nonvarying Plant GETAB Analysis Initial Conditions". The Cooper core pressure is given as 1045 psia. A value of 1035 psia, which more nearly reflects actual plant data, was assumed for this submittal.
Reference B-1 has been amended to eliminate this discrepancy.
REFERENCES B-1 Licensing Topical Report, " General Electric Boiling Water Reactor, Generic Reload Fuel Application", NEDE-240ll-P-A-1, August 1979.
i 0
19/20
Y1003J01A17 APPENDIX C DENSIFICATION POWER SPIKING Reference C-1 documents the NRC staff position that "... it (is) acceptable to remove the 8xs and 8x8R spiking penalty factor from the plant Technical Specification for those operating BWR's for which it can be shown that the predicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGR's".
i l
The Cooper Reload-6 submittal contains-the required information to remove the power spiking penalty from the Cooper Technical Specifications. Section 10, Rod Withdrawal Error, and Appendix A (Bundle Loading k.cident) include the
-densification effect in the calculated LHGR of the 8x8 fuels.
-REFERENCES C-1
" Safety Evaluation of the General Electric Methods for the Consideration of Power Spiking Due to Densification Effects in BWR 8x8 Fuel Design and Performance", Reactor Safety Branch, DOR, May 1978.
21/22
~
Y1003J01A17 I
APPENDIX D CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
All rapid pressurization and overpressure protection events have been analyzed using the ODYN transient code as specified in Reference D-1.
Code overpressure protection analysis results are deterministic, as discussed in Reference D-2.
The ACPR values given for the pressurization event-in Section 9 are the plant-specific deterministic values calculated by ODYN based on the initial CPR given in Item 7 of this submittal. These ACPRs may be adjusted to reflect either Option A or Option B ACPRs by employing the conversion method described in Reference D-1 or D-2, respectively. These adjustments are based on conservatism factors applied to the ratio ACPR/ICPR.
The MCPR for the event is determined by Eddi.sg the ACPR to the safety limit.
Section II presents both the MCPRs for the nonpressurization events, as well as the adjusted MCPRs (for both Option A and Option B) for the pressurization events.
The operating limit MCPR is the maximum MCPR of the following events:
l (1) turbine trip or load rejection without bypass based on ODYN; (2) feedwater controller failure event based on ODYN;
-(3) loss of feedwater heating event; (4) rod withdrawal error event; (5) bundle loading error accident; (6) minimum required by LOCA; and (7) minimum required by Reference D-3, Appendix D where the loss of feedwater heating, rod withdrawal error, and loading error MCPRs cre calculated as described in Reference D-3.
However, the MCPRs for the pressur-ization events analyzed with ODYN have been adjusted as follows:
(1) MCPRs adjusted for Option A (adding'O.044 to ACPR/ICPR) for all plants choosing to operate under Option A'.
23
Y1003J01A17 (2) MCPRs adjusted for Option B for all plants choosing to operate under Option B which meet all scram specifications given in Reference D-2.
(3) MCPRs are determined by a linear interpolation between the Option A MCPR and the Option B MCPR for all plants choosing to operate under Option B which do not meet the scram time specification. This inter-
.polation is based on the tested measured scram time and is described in Reference D-2.
General Electric's one-dimensional core transient model ODYN computer code has been used for pressurization transient analysis (Reference D-4).
0 h
QiA si v
Plant Power ow Transient Exposure (%)
(%) (% NBR) ( NBR) (psig) ( g ) 7x7 8x8 8x8R P8xBR Response Load BOC-EOC 104.3 100 501.5 122.3 1179 1213 0.14 0.19 0.19 0.21 Figare D-1 Rejection Without Bypass Feedwater BOC-EOC 104.3 100 314.4 119.1 1135 1172 0.09 0~.13 0.14.0.16 Figure D-2 p
Controller Failure BEFERENCES D-1 Letter, R. P. Denise (NRC) to G. G. Sherwood (GE), January 23, 1980.
D-2 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.
D-3 " Generic Reload Fuel Application", NEDE-240ll-P-A-1, August 1979.
D-4 "One-Dimensional Core Transient Model", NEDE-24154P, Volumes 1, 2 and 3, October 1978.
24
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Y1003J01A17 APPENDIX E OVERPRESSURIZATION ANALYSIS SDNARY (5.3)
General Electric's one-dimensional core transient model ODYN computer code has been used for pressurization transient analysis (Reference E-1).
Power Core Flow s1 y
Plant Transient
(%)
(%)
(psig)
(psig)
Response
MSIV Closure 104.3 100 1243 1270 Figure E-1 (Flux Scram)
REFERENCE E-1 NEDE-24154P, Volumes 1, 2, and 3 "One-Dimensional Core Transient Model",
October 1978.
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