Errata & Addenda Sheet 15,replacing Pages v/vi,3-2,4-3 & 4-16,to LOCA Analysis Rept for Dresden Units 2 & 3 & Quad Cities 1 & 2 Nuclear Power Station| ML20215A534 |
| Person / Time |
|---|
| Site: |
Dresden, Quad Cities, 05000000 |
|---|
| Issue date: |
05/21/1986 |
|---|
| From: |
GENERAL ELECTRIC CO. |
|---|
| To: |
|
|---|
| Shared Package |
|---|
| ML20215A501 |
List: |
|---|
| References |
|---|
| 79NED273, NEDO-24146A-ERR, NEDO-24146A-ERR-15, NUDOCS 8610060112 |
| Download: ML20215A534 (8) |
|
|
|---|
Category:FUEL CYCLE RELOAD REPORTS
MONTHYEARML20196G1241998-11-30030 November 1998
COLR for Quad Cities Unit 1 Cycle 16
ML20236T8331998-06-30030 June 1998
COLR for Dresden Station Unit 3,Cycle 15
ML20236T8391998-06-30030 June 1998
Rev 1 to EMF-96-141, Dresden Unit 3 Cycle 15 Reload Analysis Rept
ML20247N6281998-05-19019 May 1998
Rev 2 to COLR for Quad Cities Unit 2 Cycle 15
ML20198R1911997-11-0404 November 1997
COLR for Quad Cities Unit 2 Cycle 15
ML20198R1451997-11-0404 November 1997
Rev 1 to Quad Cities,Unit 2 Cycle 15 Colr
ML20117H7321996-05-31031 May 1996
Revised COLR for Quad Cities Unit 1,Cycle 15
ML20117H8791996-05-31031 May 1996
Revised COLR for Quad Cities Unit 2,Cycle 14
ML20077E3421994-12-0606 December 1994
Revised COLR for Quad-Cities Unit 2 Cycle 13
ML20069F7221994-05-31031 May 1994
Cycle 14 Core Operating Limits Rept
ML20045A5481993-05-31031 May 1993
Quad-Cities Nuclear Power Station Unit 1,Summary of Fuel Performance,End-of-Cycle 12, May 1993
ML20044F9931993-05-21021 May 1993
Revised Quad-Cities Unit 1,Cycle 13 Colr
ML20044F9731993-05-21021 May 1993
Revised, Quad-Cities Unit 2,Cycle 13 Colr
ML20125B5301992-12-0303 December 1992
Cycle 13 Colr
ML20101A0201992-01-31031 January 1992
Core Operating Limits Rept Quad Cities Nuclear Power Station,Unit 2,Reload 11 (Cycle 12)
ML20066F4001991-01-21021 January 1991
Core Operating Limits Rept for Quad-Cities Unit 1 Cycle 12
ML17202L2591990-05-31031 May 1990
Rev 1 to Core Operating Limits Rept,Dresden Station Unit 2, Cycle 12.
ML17202G7761989-12-31031 December 1989
Core Operating Limits Rept,Dresden Station Unit 2,Cycle 12,Rev 0.
ML20006D3521989-12-31031 December 1989
Core Operating Limits Rept,Dresden Station,Unit 3,Cycle 12,Rev 0
ML19325C9081989-10-11011 October 1989
Core Operating Limits Rept for Quad Cities Nuclear Power Station Unit 1,Reload 10 (Cycle 11).
ML20247D3171989-05-31031 May 1989
Core Operating Limits Rept for Quad Cities Nuclear Power Station Unit 2,Reload 9 (Cycle 10)
ML20247D3071989-05-31031 May 1989
Core Operating Limits Rept for Quad Cities Nuclear Power Station Unit 1,Reload 9 (Cycle 10)
ML20148N4111988-01-31031 January 1988
Rev 0 to, Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station Unit 2,Reload 9,Cycle 10
ML18149A4301986-09-30030 September 1986
Rev 1-A to Reload Nuclear Design Methodology.
ML20215A5451986-08-31031 August 1986
Errata & Addenda Sheet 14,replacing Pages v/vi,3-2,4-3,4-10, 4-15 & 4-16,to LOCA Analysis Rept for Dresden Units 2 & 3 & Quad Cities 1 & 2 Nuclear Power Station
ML20215A5271986-07-31031 July 1986
Rev 0 to Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station,Unit 2,Reload 8 (Cycle 9)
ML20215A5341986-05-21021 May 1986
Errata & Addenda Sheet 15,replacing Pages v/vi,3-2,4-3 & 4-16,to LOCA Analysis Rept for Dresden Units 2 & 3 & Quad Cities 1 & 2 Nuclear Power Station
ML20153G3051985-09-12012 September 1985
Cycle 10 Reload Analysis,Design & Safety Analyses for ENC XN-3 9x9 Reload Fuel
ML20153G3261985-09-12012 September 1985
Cycle 10 Plant Transient Analysis
ML20153G3521985-09-12012 September 1985
LOCA-ECCS Analysis Maplhgr,Results for 9x9 Fuel. Related Info Encl
ML20135H9491985-08-31031 August 1985
Rev 1 to Reload Nuclear Design Methodology
ML20212Q2301985-08-31031 August 1985
Rev 0 to Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station Unit 1,Reload 8 (Cycle 9)
ML20076G9741983-08-0101 August 1983
Cycle 9 Reload Analysis
ML20024C2481983-05-31031 May 1983
Rev 0 to Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station,Unit 2,Reload 6 (Cycle 7)
ML20071Q0871982-12-0606 December 1982
Nonproprietary Version of Revision 1 to Dresden Unit 2 Cycle 9 Reload Analysis
ML20030C1281981-06-30030 June 1981
Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station,Unit 2 Reload 5 (Cycle 6)
1998-06-30
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEAR05000254/LER-1999-004-03, :on 990910,CREVS Air Filtration Unit Was Declared Inoperable.Caused by Airflow Rate in Excess of TS Allowable.Airflow Was Adjusted During Performance of Surveillance Procedure Qcos 5750-021999-10-12012 October 1999
- on 990910,CREVS Air Filtration Unit Was Declared Inoperable.Caused by Airflow Rate in Excess of TS Allowable.Airflow Was Adjusted During Performance of Surveillance Procedure Qcos 5750-02
05000254/LER-1999-003-05, :on 990907,noted That HPCI Was Inoperable Due to Manual Closure of HPCI Steam Supply Isolation Valve.Cause Indeterminate.Suppl LER Will Be Issued Upon Determination of Root Cause of Control Switch Failure1999-10-0707 October 1999
- on 990907,noted That HPCI Was Inoperable Due to Manual Closure of HPCI Steam Supply Isolation Valve.Cause Indeterminate.Suppl LER Will Be Issued Upon Determination of Root Cause of Control Switch Failure
SVP-99-204, Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-09-30030 September 1999
Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20217A9931999-09-30030 September 1999
NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data
ML20249C8491999-09-30030 September 1999
1999 Third Quarter Rept of Completed Changes,Tests & Experiments Evaluated,Per 10CFR50.59(b)(2), for Dresden Nuclear Power Station. with
ML20216H8481999-09-23023 September 1999
Safety Evaluation Supporting Amends 190 & 187 to Licenses DPR-29 & DPR-30,respectively
ML20217A1691999-09-22022 September 1999
Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers
ML20212J0501999-09-21021 September 1999
Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1
SVP-99-179, Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-08-31031 August 1999
Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20210U7831999-08-0404 August 1999
Safety Evaluation Supporting Amends 189 & 186 to Licenses DPR-29 & DPR-30,respectively
ML20210L8661999-08-0202 August 1999
Safety Evaluation Accepting License 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs
SVP-99-155, Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-07-31031 July 1999
Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20210R6081999-07-31031 July 1999
Monthly Operating Repts for July 1999 for Dresden Nuclear Power,Units 1,2 & 3.With
ML20209E1291999-06-30030 June 1999
Monthly Operating Repts for June 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
ML20210D3071999-06-30030 June 1999
Corrected Page 8 to MOR for June 1999 for DNPS Unit 3
ML20209J3481999-06-30030 June 1999
1999 Second Quarter Rept of Completed Changes,Tests & Experiments, Per 10CFR50.59.With
SVP-99-148, Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-06-30030 June 1999
Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20196H8621999-06-30030 June 1999
NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept
ML20196F4261999-06-25025 June 1999
Safety Evaluation Supporting Amends 188 & 185 to Licenses DPR-29 & DPR-30,respectively
05000265/LER-1999-002-04, :on 990527,station Personnel Discovered Unsealed One Inch Diameter Pipe Penetrating Secondary Containment.Caused by Lack of Proper Work Control.Util Will Permanently Cap Spare Subject Pipe by 9907151999-06-25025 June 1999
- on 990527,station Personnel Discovered Unsealed One Inch Diameter Pipe Penetrating Secondary Containment.Caused by Lack of Proper Work Control.Util Will Permanently Cap Spare Subject Pipe by 990715
05000254/LER-1999-004-02, :on 990521,reactor Scram Occurred.Caused by Steam Intrusion Into Scram Discharge Volume.Procedure Qcop 12200-11, RWCU Sys Startup & Pump Operation, Revised1999-06-18018 June 1999
- on 990521,reactor Scram Occurred.Caused by Steam Intrusion Into Scram Discharge Volume.Procedure Qcop 12200-11, RWCU Sys Startup & Pump Operation, Revised
ML20195K1481999-06-16016 June 1999
Safety Evaluation Authorizing Relief Request RV-23A for Duration of Current 10 Yr IST Interval on Basis That Compliance with Code Requirements Would Result in Hardship Without Compensating Increase in Level of Quality & Safety
SVP-99-123, Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-05-31031 May 1999
Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20195G6381999-05-31031 May 1999
Monthly Operating Repts for May 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
ML20210H3501999-05-25025 May 1999
Rev 1 to Quad Cities Station IPEEE Submittal Rept
ML20207E1621999-05-25025 May 1999
Rev 1 to Quad Cities IPEEE, Consisting of Revised Chapter 4.0 Re Internal Fire Analysis
ML20195B2591999-05-19019 May 1999
Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations
05000254/LER-1999-001-04, :on 970715,three Spare Valves Were Determined to Require Monthly Verification in Accordance with TS SR 4.7.A.2.Caused by Inadequate Review of Procedure Rev. Surveillance Revised to Ensure to Continued Compliance1999-05-12012 May 1999
- on 970715,three Spare Valves Were Determined to Require Monthly Verification in Accordance with TS SR 4.7.A.2.Caused by Inadequate Review of Procedure Rev. Surveillance Revised to Ensure to Continued Compliance
SVP-99-102, Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with1999-04-30030 April 1999
Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with
SVP-99-104, Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-04-30030 April 1999
Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20206N2821999-04-30030 April 1999
Monthly Operating Repts for Apr 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
ML20205Q5291999-04-16016 April 1999
SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained
ML20205J6011999-04-0707 April 1999
Safety Evaluation Accepting Proposed Merger of Calenergy Co, Inc & Midamerican Holdings Co for Quad Cities Nuclear Power Station,Units 1 & 2
SVP-99-071, Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-03-31031 March 1999
Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20205N7491999-03-31031 March 1999
Monthly Operating Repts for Mar 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
ML20206B1901999-03-31031 March 1999
First Quarter Rept of Completed Changes,Tests & Experiments Per 10CFR50.59 for Dresden Nuclear Power Station. with
ML20205F4791999-03-31031 March 1999
Safety Evaluation Supporting Amends 187 & 184 to Licenses DPR-29 & DPR-30,respectively
ML20205D4171999-03-26026 March 1999
Safety Evaluation Supporting Amends 186 & 183 to Licenses DPR-29 & DPR-30,respectively
ML20205C5671999-03-19019 March 1999
Simulator Four-Yr Certification Rept
ML20204D9751999-03-17017 March 1999
Safety Evaluation Supporting Amends 185 & 182 to Licenses DPR-29 & DPR-30,respectively
ML20207D2341999-03-0101 March 1999
Post Outage (90 Day) Summary Rept, for ISI Exams & Repair/Replacement Activities Conducted 981207-1205
ML20204B1571999-02-28028 February 1999
Monthly Operating Repts for Feb 1999 for Quad Cities,Units 1 & 2.With
ML20207M6921999-02-28028 February 1999
Monthly Operating Repts for Feb 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
ML20207E9311999-02-26026 February 1999
Part 21 Rept Re Sprague Model TE1302 Aluminum Electrolytic Capacitors with Date Code of 9322H.Caused by Aluminum Electrolytic Capacitors.Affected Capacitors Replaced
SVP-99-021, Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With1999-01-31031 January 1999
Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With
ML20199M0891999-01-22022 January 1999
Part 21 Rept Re Failure of Square Root Converters.Caused by Failed Aluminum Electrolytic Capacitory Spargue Electric Co (Model Number TE1302 with Mfg Date Code 9322H).Sent Square Root Converters Back to Mfg,Barker Microfarads,Inc
ML20199C8951998-12-31031 December 1998
Monthly Operating Repts for Dec 1998 for Dnps,Units 1,2 & 3
ML20199D3261998-12-31031 December 1998
10CFR50.59 SER for 1998-04 Quarter, of Changes,Tests & Experiments.With
ML20205M7061998-12-31031 December 1998
Unicom Corp 1998 Summary Annual Rept. with
ML20205D1311998-12-31031 December 1998
1998 Decommissioning Funding Status Rept for Yr Ending 981231 for Quad Cities Nuclear Power Station,Units 1 & 2
1999-09-30
[Table view] |
Text
.-
%1WL*
CUCLEAR ENERGY BUSINESS OPER ATIONS o CENER AL ELECTRIC COMPANY SAN JOSE CALIFORNIA 95125 l'
l GEN ER AL $ ELECTRIC APPLICABLE TO:
" ** "" 'NEDO-24146A ERRATA And ADDENDA 79Nzn273 SHEET LOCA Analysis Report for TIT u I
Dresden Units 2, 3 and Quad May 1986 l
oan Cities 1.2 Nuclear Power Stations NOTE: Corroet s// copses of the applicable l
April 1979 ISSUE DATE gyg,gjof,,, gjfg,,,
l INSTRUCTIONS l
P RAG A L Nel GCORRECTIONS AND ADDITeoNS) t 1.
Page v/vi Replace with new page v/vi l
2.
Page 3-2 Replace with new page 3-2 3.
Page 4-3 Replace with new page 4-3 4.
Page 4-16 Replace with new page 4-16 i
l t
(Change brackets in right-hand margin indicate areas where report has been revised.)
l l
l 8610060112 860918 PDR ADOCK 05000265 P
PDR l
l Of l PAGE
r-NED0-24146A LIST OF WARTRS Table Title Pg i
1 Significant Input Parameters to the Loss-of-Coolant Accident 3-1 2
Summary of Break Spectrum Results 4-5 3
LOCA Analysis Figure Summary - Non-Lead Plant 4-6 4A MAPLEGR Versus Average Planar Exposure (7D212 - No Gad.)
4-7 4B MAPLHGR Versus Average Planar Exposure (7D212L) 4-7 l
4C MAPLHGR Versus Average Planar Exposure (7D230) 4-8 4D MAPLHGR Versus Average Planar Exposure (EEIC - Pu)
'4-8 4E MAPLHGR Versus Average Planar Exposure (8D250) 4-9 l
4F MAPLHGR Versus Average Planar Exposure (8D262) 4-9 4G MAPLHGR Versus Average Planar Exposure (8DRB265L) 4-10 4H MAPLHGR Versus Average Planar Exposure (Barrier LTA) 4--10
{
4I MAPLHGR Versus Average Planar Exposure (P8DRB282) 4-11 I
4J MAPLHGR Versus Average Planar Exposure (P8DRB265H/BP8DRB265H) 4-11 4K MAPLHGR Versus Average Planar Exposure (P8DRB239) 4-12 4L MAPLHGR Versus Average Planar Exposure (P8DGB284)*
4-12 4M MAPLHGR Versus Average Planar Exposure (P8DGB263L)*
4-13 4N MAPLHGR Versus Average Planar Exposure (P8DGB263H)*
4-13 40 MAPLHGR Versus Average Planar Exposure (P8DGB298)*
4-14 4P
- MAPLHGR Versus Average Planar Exposure (P8DRB265L)
.'and (FSDG2265L)*
4-14 4Q MAPLHGR Versus Average Planar Exposure (EP8DRB283H) 4-15 I.
4R MAPLHGR Versus Average Planar Exposure (BP8DRB282) 4-15 4S MAPLHGR Versus Average Planar Exposure (BP8DRB299) 4-16 4T MAPLHGR Versus Average Planar Exposure (BP8DRB299L) 4-16]
i
- Barrier fuel for the Barrier Fuel Demonstration Program l
l 1
v/vi
f~
NEDO-24146A i
3.
INPUT TO ANALYSIS A list of the significant plant input parameters to the LOCA analysis is presented in Table 1.
Table 1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Paraceters Core Thermal Power 2578 MWt, which corresponds to 102% of rated power Vessel Steam Output 9.96 x 106 lba/h, which corresponds to 102% of rated power Vessel Steam Dome Pressure 1020 psia Recirculation Line Break Area 4.18 ft (DBA) for Large Breaks - Suction Number of Drilled Bundles 156 Fuel Parameters:
Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Fuel Bundle Generation Kate Peaking Power Fuel Type Geometry (kW/ft)
Factor Ratio
7D2,12 - No Gad 7x7 17.5 1.57 1.2 B.
7D212L 7x7 17.5 1.57 1.2 C.
7D230 7x7 17.5 1.57 1.2 D.
EEIC - Pu 7x7 17.5 1.57 1.2 E.
8D250 8x8 13.4 1.57 1.2 F.
8D262 8x8 13.4 1.57 1.2 C.
8DRB265L 8x8 13.4 1.57 1.2 B.
Barrier LTA 8x8 13.4 1.57 1.2
~
~
1.
P8DRB282 " ~~~~
8x8 13.4 ~
1.57 1.2 J.
P8DRB265H/BP8DRB265H 8x8 13.4 1.57 1.2
~
K.
PSDRB239 8x8 13.4 1.57 1.2
- To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e.,1.2 divided by 1.02) for a bundle with an initial MCPR of 1.20, 3-1
NEDO-24146A Table 1 (Continued)
Fuel Parameters:
(Continued)
Peak Technical Initial Specification Design Minimus Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power Fuel Type Geometry kW/ft)
Factor Ratio L.
P8DGB284**
8x8 13.4 1.57 1.2 M.
P8DGB263L**
8x8 13.4 1.57 1.2 N.
P8DGB263Ha*
8x8 13.4 1.57 1.2" 0.
P8DGB298**
8x8 13.4 1.57 1.2 P.
P8DRB265L/
8x8 13.4 1.57 1.2 P8DGB265L**
Q.
BP8DRB283H 8x8 13.4 1.57 1.2 1.
BP8DRB282 8x8 13.4 1.57 1.2 S.
BP8DRB299 8x8 13.4 1.57 1.2 T.
BP8DRB299L 8x8 13.4 1.57 1.2
- Barrigt fuel for the Barrier Fuel Demonstration Program 3-2
NED0-24146A 4.5 RESULTS OF THE CHASTE ANALYSIS This code is used, with suitable inputs from the other codes, to calculate the fuel cladding heatup rate, peak cladding temperature, peak local cladding oxidation, and core-wide metal-water reaction for large breaks. The detailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction. The empirical core spray heat transfer and channel wetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel assembly. Iterative applications of CHASTE determine the maximum permissible planar power where required to satisfy the requirement's of 10CFR50.46 acceptance criteria.
The CHASTE resulta presented are:
Peak Cladding Temperature versus time e
o Peak Cladding Temperature versus Break Ares e
Peak Cladding Temperature and Peak Local oxidation versus Planar Average Exposure for the most limiting break size.
I l
e Maximum Averrage Planar Heat Generation Rate (MPLHGR) versus Planar Average Exposure for the most limiting break size A summary of the analytical results is given in Table 2.
Table 3 lists the figures provided for this analysis. The MPLBGR values for each fuel type for D2,3/QC1,2 are presented in Tables 4A through 4T.
4.6 METHODS In the following sections, it will be useful to refer to the methods used to analyze DBA, large breaks, and small breaks. For jet-pump reactors, these are defined as follows:
a.
DBA Methods. LAMB / SCAT / SAFE /DBA-REFLOOD/ CHASTE. Break size: DBA.
4-3
T NEDO-24146 b.
Large Break Methods (LBM). LAMB / SCAT / SAFE /non-DBA REFLOOD/ CHASTE.
Break sizes: 1.0 ft f_ A < DBA.
b.
Small Break Methods (SBM). SAFE /non-DBA REFLOOD. Beat transfer coefficients: nucleate boiling prior to core uncovery, 25 Btu /hr-ft
- F after recovery, core spray when appropriate. Peak cladding temperature and peak local oxidation are calculated in 2
non-DBA-REFLOOD. Break sizes: A < 1.0 ft,
O l
e l
i l
4-4 s
-.,,,-.---n,-nn-,ny,.=-
,.,-n,
-~--n
I l
NEDO-24146A e
Table 4Q MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Quad Cities 1,2 FUEL TYPE:
BP8DRB283H Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)
(kW/ft)
(*F)
Fraction 200 11.2 2128 0.028 1,000 11.2 2121 0.028 5,000 11.7 2157 0.030 10,000 12.0 2192 0.033 15,000 12.0 2199 0.033
'20,000 11.9 2195 0.033 25,000 11.4 2132 0.027 30,000 10.8 2051 0.038 35.000 10.3 1956 0.031 40,000 9.6 1841 0.009 45,000 9.0 1764 0.007 NOTE: Credit taken for the effects of pre-pressurization of the fuel rods.
Table 4R MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Quad Cities 1,2 FUEL TYPE:
BP8DRB282 Average Planar Exposure MAPLHGR PCT 0xidation CMWd/t)
(kW/ f t)
(*F)
Fraction 200 11.2 2131 0.029 1,000 11.2 2128 0.028 5,000 11.8 2178 0.032 10,000 12.0 2188 0.032 15,000 12.1 2199 0.033 20,000 11.9 2192 0.033 25,000 11.4 2129 0.027 i
30,000 10.8 2047 0.038 l
35,000 10.3 1957 0.031 40,000 9.6 1840 0.009 45,000 8.9 1761 0.007 I
NOTE: Credit taken for the effects of pre-pressurization of the. fuel rods.
I 4-15 a.__-,..-_--___-__...--_
f NEDO-24146A Table 4S MAPLHGR VERSUS AVERAGE PIANAR EXPOSURE PIANT: Quad Cities 1&2 PUEL. TYPE: BP8DRB299 Average Planar Exposure M&PLHGR PCT 0xidation (MWD /st)
(kW/ft)
(*F)
Fraction 200 10.90 2089 0.025 1,000 11.00 2090 0.025 5,000 11.50 2126 0.027 10,000 12.10 2199 0.033 15,000 12.10 2198 0.033 20,000 11.90 2197 0.033 25,000 11.50 2157 0.030 30,000 11.00 2072 0.049 35,000 10.30 1977 0.035 40,000 9.70 1879 0.021 45,000 9.00 1777 0.007 NOTE: Credit taken for the effects of pre-pressurization of the fuel rods.
Table 4T MAPLHGR VERSUS AVERAGE PIANAR EXPOSURE l
PIANT: Quad Cities 1&2 FUEL TYPE: BP8DRB299L Average Planar.
Exposure MAPLHGR PCT 0xidation (MWD /st)
(kW/ft)
(*F)
Fraction l
200 11.00 2094 0.025 l
1,000 11.00 2099 0.026 i
5,000 11.60 2145 0.029 10,000 12.20 2200 0.033 l
15,000 12.10 2198 0.033 20,000 12.00 2197 0.033 25,000 11.50 2139 0.027 30,000 10.90 2067 0.048 35,000 10.30 1973 0.034 40,000 9.70 1876 0.020 45,000 9.00 1776 0.007 i
NOTE: Credit taken for the effects of pre-pressurization of the fuel rods.
~
1 i
4-16
_ ~.
_ _, _ _ _ _ _ _ _ _ _ _ _ _ _ _