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 Stations| ML20205F271 |
| Person / Time |
|---|
| Site: |
Dresden, Quad Cities, 05000000 |
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| Issue date: |
08/31/1985 |
|---|
| From: |
GENERAL ELECTRIC CO. |
|---|
| To: |
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|---|
| Shared Package |
|---|
| ML20205F242 |
List: |
|---|
| References |
|---|
| NEDO-24146A-ERR, NEDO-24146A-ERR-14, NUDOCS 8511050275 |
| Download: ML20205F271 (11) |
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|---|
Category:REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)
MONTHYEARML20198R2331997-04-30030 April 1997
Rev 2 to Quad Cities Unit 2 Cycle 145 Reload Analysis
ML20198T3251997-04-18018 April 1997
Non-proprietary Rev 1 to EMF-96-180, Quad Cities Unit 2 Cycle 15 Plant Transient Analysis
ML20078E5171995-01-25025 January 1995
Correction to Encl 3, Equivalent Margin Analysis Plant Applicability Verification Form of Topical Rept NEDO-32205-A,Rev 1
ML20078D8611995-01-0505 January 1995
Shroud Repair Seismic Analysis
ML20078A7451994-06-30030 June 1994
Evaluation of Indications Found at H5 Weld Location in Quad Cities Unit 1 Shroud
ML17187A8051993-01-31031 January 1993
Suppl 1 to GENE-770-26-1092, Long Term Containment Response to DBA-LOCA w/65 Fdegrees F Svc Water Temperature,80 Degrees F Initial Pool Temperature & Initiation of Containment Cooling at 1800 Seconds.
ML17252A7531992-11-30030 November 1992
Lpci/Containment Cooling Sys Evaluation.
ML20065Q8531990-11-30030 November 1990
Rev 0 to, Comm Ed Topical Benchmark of BWR Nuclear Design Methods
ML20006A8231990-01-10010 January 1990
Errata to Rev 3 to BAW-1543, Master Integrated Reactor Vessel Surveillance Program Consisting of Revised Tables 3-20 & E-1
ML19332E4071989-10-16016 October 1989
Errata & Addendum 1 to, Safety Evaluation to Justify Operation W/Loss of Jet Pump Flow Indication for Quad Cities 1 & 2.
ML19351A4191989-09-30030 September 1989
Mark-BW Reload LOCA Analysis for Catawba & McGuire Units.
ML19325C9131989-03-31031 March 1989
Safety Evaluation to Justify Operation W/Loss of Jet Pump Flow Indication for Quad Cities,Units 1 & 2.
ML20235M9111989-02-20020 February 1989
Nonproprietary Catawba Unit 1 Evaluation for Tube Vibration Induced Fatigue
ML20207L7781988-10-0606 October 1988
Nonproprietary WCAP 11936, McGuire Unit 2 Evaluation for Tube Vibration Induced Fatigue
ML20154M2421988-09-30030 September 1988
Simulate-3 Validation & Verification
ML20207L9261988-09-30030 September 1988
Nonproprietary Addendum 2 to COBRA-NC,Analysis for Main Steamline Break in Catawba Unit 1 Ice Condenser Containment (Response to NRC Questions)
ML20154A4131988-07-13013 July 1988
Extended Operating Domain/Equipment Out-of-Svc Analysis for Dresden Units 2 & 3
ML20154J9831988-04-30030 April 1988
RCS Flow Anomaly Investigation Rept
ML20147F4101988-02-0909 February 1988
Rev 49 to CE-1-A, QA Program for Nuclear Generating Stations
ML20237C4091987-11-30030 November 1987
Nonproprietary Rev 2 to Byron/Braidwood T-Hot Reduction Final Licensing Rept
ML20154A4021987-09-28028 September 1987
LOCA-ECCS Analysis for Dresden Units During Single Loop Operation W/Advanced Nuclear Fuels Corp Fuel
ML20235C4871987-07-31031 July 1987
Extended Operating Domain & Equipment Out-of-Svc for Quad Cities Nuclear Power Station Units 1 & 2. B Wolfe Affidavit Also Encl
ML20235C4331987-06-30030 June 1987
Rev 0 to Suppl Reload Licensing Submittal for Quad Cities Nuclear Power Station Unit 1,Reload 9
ML18149A5121986-11-30030 November 1986
Qualification of WRB-1 CHF Correlation in VEPCO Cobra Code.
ML20207H9761986-11-30030 November 1986
Nonproprietary Resistance Temp Detector Bypass Elimination Licensing Rept for Byron 1 & 2 & Braidwood 1 & 2
ML20210A6051986-09-0505 September 1986
Addendum a to Vol II of Nusco 140-2, Nusco Thermal Hydraulic Model Qualification
ML20155J4911986-05-31031 May 1986
Rev 0 to Core Protection Calculator/Control Element Assembly Calculator Software Mods for CPC Improvement Program Reload Data Block
ML20155J5231986-05-31031 May 1986
Rev 1 to Functional Design Requirement for Core Protection Calculator
ML20155J5051986-05-31031 May 1986
Rev 1 to Functional Design Requirements for Control Element Assembly Calculator
ML20203G5481986-03-31031 March 1986
Relaxed Power Distribution Control Methodology & Associated Fq Surveillance Tech Specs
ML20151Z0641986-01-31031 January 1986
Rev 0 to CEN-323-NP Reload Data Block Constant Installation Guidelines
ML20151Z0521986-01-31031 January 1986
Rev 3 to CEN-39(A)-NP, CPC Protection Algorithm Software Change Procedure
ML20140A6821985-12-31031 December 1985
Reactor Vessel Fluence & Ref Temp for Pressurized Thermal Shock Evaluations
ML20138M4211985-09-30030 September 1985
Justification for Raising Setpoint for Reactor Trip on High Pressure, Transient Assessment Program
ML20135H9491985-08-31031 August 1985
Rev 1 to Reload Nuclear Design Methodology
ML20135A5571985-08-31031 August 1985
B&W Owners Group Cavity Dosimetry Program
ML20205F2711985-08-31031 August 1985
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 Stations
ML20134N3021985-07-31031 July 1985
Rev 00 to Functional Design Requirement for Core Protection Calculator
ML20134N2941985-07-31031 July 1985
Rev 00 to Functional Design Requirement for Control Element Assembly Calculator
ML17346B1711985-07-31031 July 1985
Retran Code Transient Analysis Model Qualification.
ML20133H3591985-07-31031 July 1985
Statistical DNBR Evaluation Methodology
ML20138P1541985-07-0202 July 1985
Rev 1 to TMI-1 Nuclear Generating Station Natural Circulation Cooldown Analysis W/O Reactor Vessel Upper Head Void Formation, Topical Rept
ML20127K1381985-06-30030 June 1985
Nuclear Physics Methodology for Reload Design
ML20128Q2751985-05-31031 May 1985
Revised Nomad Code & Model
ML20129E9881985-05-30030 May 1985
VEPCO Reactor Sys Transient Analysis Using Retran Computer Code
ML20106A6481985-01-31031 January 1985
Errata & Addenda Sheet 13,replacing Pages v/vi,3-2,4-3, 4-15 & 4-16 to, LOCA Anaysis Rept for Dresden Units 2 & 3 & Quad Cities 1 & 2 Nuclear Power Stations
ML20111C1171985-01-31031 January 1985
Rev 1 to Comprehensive Vibration Assessment Program for Palo Verde Nuclear Generating Station Unit 1 (Sys 80 Prototype),Evaluation of Predictions & Pre-Core Hot Functional Measurement & Insp Programs
ML20137F0331985-01-15015 January 1985
Rev 1 to Criticality Safety Analysis,Dresden 2,3 Spent Fuels Storage Pool W/Exxon Nuclear Co,Inc 9x9 Reload Fuel
ML18142A1531984-12-31031 December 1984
Nonproprietary Version of Revised VEPCO Evaluation of Control Rod Ejection Transient.
ML17298B5401984-11-30030 November 1984
Summary Rept on Operability of Shutdown Cooling Sys Relief Valves for Palo Verde Units 1,2 & 3.
1997-04-30
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999
NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data
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
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
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
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
ML20210R6081999-07-31031 July 1999
Monthly Operating Repts for July 1999 for Dresden Nuclear Power,Units 1,2 & 3.With
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
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
ML20209E1291999-06-30030 June 1999
Monthly Operating Repts for June 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.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
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
ML20195G6381999-05-31031 May 1999
Monthly Operating Repts for May 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
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
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
ML20206N2821999-04-30030 April 1999
Monthly Operating Repts for Apr 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
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
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
ML20206B1901999-03-31031 March 1999
First Quarter Rept of Completed Changes,Tests & Experiments Per 10CFR50.59 for Dresden Nuclear Power Station. with
ML20205N7491999-03-31031 March 1999
Monthly Operating Repts for Mar 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
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
ML20205C5671999-03-19019 March 1999
Simulator Four-Yr Certification Rept
ML20207D2341999-03-0101 March 1999
Post Outage (90 Day) Summary Rept, for ISI Exams & Repair/Replacement Activities Conducted 981207-1205
ML20207M6921999-02-28028 February 1999
Monthly Operating Repts for Feb 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With
ML20204B1571999-02-28028 February 1999
Monthly Operating Repts for Feb 1999 for Quad Cities,Units 1 & 2.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
ML20199D3261998-12-31031 December 1998
10CFR50.59 SER for 1998-04 Quarter, of Changes,Tests & Experiments.With
ML20199C8951998-12-31031 December 1998
Monthly Operating Repts for Dec 1998 for Dnps,Units 1,2 & 3
ML20205M7061998-12-31031 December 1998
Unicom Corp 1998 Summary Annual Rept. with
SVP-99-007, Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With1998-12-31031 December 1998
Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With
ML20205D1311998-12-31031 December 1998
1998 Decommissioning Funding Status Rept for Yr Ending 981231 for Quad Cities Nuclear Power Station,Units 1 & 2
ML20196D9651998-11-30030 November 1998
Safety Evaluation Supporting Relief Requests CR-21 & CR-24, Respectively.Relief Request CR-23,proposed Alternative May Be Authorized,Per 10CFR50.55a & Relief Request CR-22 Was Withdrawn by Licensee
ML20197G8591998-11-30030 November 1998
Monthly Operating Repts for Nov 1998 for Dresden Nuclear Power Station.With
ML20196G1241998-11-30030 November 1998
COLR for Quad Cities Unit 1 Cycle 16
ML20196C8731998-11-30030 November 1998
Rev 0 to GE-NE-B13-01980-30-1, Fracture Mechanics Evaluation on Observed Indications at Two Welds in Recirculation Piping of Quad Cities,Unit 1 Station
ML20196C8391998-11-30030 November 1998
Rev 0 to GE-NE-B13-01980-030-2, Assessment of Crack Growth Rates Applicable to Induction Heating Stress Improvement (IHSI) Recirculation Piping in Quad Cities Unit 1
SVP-98-364, Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-11-30030 November 1998
Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With
ML20196A9761998-11-20020 November 1998
Safety Evaluation Re Licensee 180-day Response to GL 95-07, Thermal Binding of Safety-Related Power-operated Gate Valves
ML20196J0061998-11-19019 November 1998
Rev 66 to Topical Rept CE-1-A, QA Program
ML20196A4191998-11-19019 November 1998
Safety Evaluation Accepting QA TR CE-1-A,Rev 66 Re Changes in Independent & Onsite Review Organization by Creating NSRB
SVP-98-346, Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-10-31031 October 1998
Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With
SVP-98-358, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With1998-10-31031 October 1998
Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With
ML20195D2861998-10-31031 October 1998
Monthly Operating Repts for Oct 1998 for Dresden Nuclear Power Station.With
1999-09-30
[Table view] |
Text
__
t NUCLEA3 ENERJY BUSINESS OPER ATIONS o CENERAL ELECTRIC COMPANY g
SAN JOSE, CA LIFORhlA 95125 GEN ER AL $ ELECTRIC APPLICABLE TO:
"'""- NEDO-24146A ERRATA And ADDENDA T. I. E. No. 79NED273 SHEET TITLE LOCA ANALYSIS REPORT FDR 16 N O.
DRESDEN UNITS 2, 3 AND QUAD OATE AUGUST 1985 CITIES 1.2 NUCLEAR POWER STATI tis OE Cometau copise of use app # cable IS$*JE OATE APRIL 1979 publication as specified below.
REFERENCES
' p Yao[Ap'd tfNE , (co R RE: ONSA ADOITIONS)
- 1. Page v/vi Replace with new page v/vi
- 2. Page 3-2 Replace with new page 3-2
- 3. Page 4-3 Replace with new page 4-3
- 4. Page 4-10 Replace with new page 4-10
- 5. Page 4-15 Replace w1:h new page 4-15
- 6. Page 4-16 Replace with new page 4-16 (Change brackets in right-hand margin indicate areas where report has been revised.
R511050275 851029 PDR ADOCK 05000254 P PDR
, , 1 o f .1
NEDO-24146A LIST OF TABLES Title P_ age Table Significant Input Parameters to the Loss-of-Coolant Accident 3-1 1
Summary of Dreak Spectrum Results 4-5 2
LOCA Analysis Figure Summary - Non-Lead Plant 4-6 3
MAPLHGR Versus Average Planar Exposure (7D212 - No Gad.) 4-7 4A' MAPLHGR Versus Average Planar Exposure (7D212L) 4-7 4B '
MAPLHGR Versus Average Planar Exposure (7D230) 4-8 4C MAPLHGR Versus Average Planar Exposure (EEIC - Pu) 4-8 4D MAPLHGR Versus Average Planar Exposure (8D250) 4-9 4E MAPLHGR Versus Average P.lanar Exposure (8D262) 4-9 4F MAPLHGR Versus Average Planar Exposure (8DRB265L) 4-10 4G MAPLHGR Versus Average Planar Exposure (Barrier LTA) 4-10 4H 4I MAPLHGR Versus Average Planar Exposure (P8DRB282) 4-11 4J MAPLHGR Versus Average Planar Exposure (P8DRB265H/BP8DRB265H) 4-11 4K MAPLHGR Versus Average Planar Exposure (P8DRB239) 4-12 MAPLHGR Versus Average Planar Exposure (P8DGB284)* 4-12 4L MAPLHGR Versus Average Planar Exposure (P8DGB263L)* 4-13 4M 4N MAPLHGR Versus Average Planar Exposure (P8DGB263H)* 4-13 ,
MAPLRGR Versus Average Planar Exposure (PSDGB298)* 4-14 40 4P MAPLHGR Versus Average Planar Exposure (P8DRB265L) 4-14 and (P8DGB265L)*
MAPLHGR Versus Average Planar Exposure (BP8DRB283H) 4-15 4Q ~
4R MAPLHGR Versus Average Planar Exposure (BP8DRB282) 4-15 4S MAPLHGR Versus Average Planar Exposure (BP8DRB299) 4-16
- Barrier fuel for the Barrier Fuel Demonstration Urogram v/vi
NED0-24146A
- 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 Parameters Core Thermal Power 2578 MWt, which corresponds to 102% of rated power Vessel Steam output 9.96 x 106 lbm/h, which corresponds -
to 102% of rated power Vessel Steam Dome Pressure 1020 psia Rec.irculation 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
- A. 7D212 - 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. dD250 8x8 13.4 1.57 1.2 F. 8D262 8x8 13.4 1.57 1.2 G. 8DRB265L 8x8 13.4 1.57 1.2 H. Barrier LIA 8x8 13.4 1.57 1.2 I. P8DRB282 8x8 13.4 1.57 1.2 J. P8DRB265H/BP8DRB265H 8x8 13.4 1.57 1.2
~
K. P8DRB239 8x8 13.4
.L.57 1.2 To account for the 2% unc.ertainty 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 Minimum -
Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power Fuel Type Geomstry (kW/ft) Factor Ratio L. P8DGB284** 8x8 13.4 1.57 1.2 M. P8DGB263L** 8x8 13.4 1.57 1.2 N. P8DGB263Hi* 8x8 13.4 1.57 1.2
- 0. P 8DGB298** 8x8 13.4 1.57 1.2 P. P8DRB265L/ 820 13.4 1.57 1.2 P8DGB265L**
Q. BP8DRB283H 8x8 13.4 1.57 1.2 R. BP8DRB282 8x8 13.4 1.57 1.2 S. BP8DRB299 8x8 13.4 1.57 1.2 ,
- Barrier fuel ter the Barrier Fuel Demonstration Program O
t
=
3-2
- NEDO-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 troasient gap conductance, clad swelling and rupture, and metal-water reaction. Tne empirical core spray heat transfer and channel wetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entire LOCA trancient 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 requirements of 10CFR50.46 acceptance criteria.
The CHASTE results presented are:
o Peak Cledding Temperature versus time 4 e Peak Cladding Temperature versus Break Area e Peak Cladding Temperature and Peak Local Oxidacion versus Planar .
Average Exposure for the most limiting break size e Maximum Average Planar Heat Generation Rate (RSPLHGR) versus Planar Average Exposure for the most limiting break size A summary of thp analytical results is given in Table 2. Table 3 lists the figures provided for this analysis. The MAPLHGR values for each fuel type .
for D2,3/QC1,2 are presented in Tables 4A through 45. .
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
NEDO-24146
- b. Large Break Methods (LBM). LA'!B/ SCAT / SAFE /non-DBA REFLOOD/ CHASTE.
Break sizes: 1.0 ft < A < DBA.
- b. Small Break Methods (SBM). SAFE /non-DBA REFLOOD. Heat 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 non-DBA-REFLOOD. Break sizes: A < l.0 ft .
O 4-4
NEDO-24146A Table 4E MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Dresden 2.3/ Quad Cities 1,2 FUEL TYPE: 8D250 Average Planar-Exposure MAPLEGR PCT 0xidation (mwd /t) (kW/ft) ('F) Fraction 200 11.2 2106 0.024 1,000 11.3 2109 0.024 5,000 11.9 2169 0.029 10,000 12.1 2179 0.029 15,000 12.2 2198 0.031 20,000 12.0 2199 0.031 25,000 11.5 2148 0.027 30,000 10.6 2020 0.017 35,000 9.6 1888 0.011 40,000 9.0 1808 0.008 Table 4F MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Dresden 2,3/ Quad Cities 1,2 FUEL TYPE: 8D262 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 11.1 2104 0.024 1,000 11.3 2107 0.024 5,000 11.9 2166 0.029 10,000 12.1 2175 0.029 15,000 12.2 2199 0.031 20,000 12.0 2199 0.032 !
25,000 11.6 2157 0.028 30,000 10.7 2042 0.019 35,000 9.8 1915 0.012 40,000 9.2 1831 0.009 4-9 1
NED0-24146A Table 4G MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Dresden 2,3/ Quad Cities 1,2 FUEL TYPE: 8DRB265L Average. Planar Exposure MAPLHGR PCT 0xidation
-(mwd /t) (kW/ft) ( F) Fraction 200 11.6 2189 0.035 1,000 11.6 2188 0.034
-5,000 11.3 2198' O.034 10,000 11.9 2196 0.033 15,000 11.9 2198 0.034 20,000 11.7 2195 0.034
.25,000 11.3 2154 0.030 30,000 10.7~ 2075 0.023 35,000 10.2 2002 0.018 40,000 9.6 1913 0.013 Table 4H MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: . Quad Cities 1 FUEL TYPE: Barrier LTA Average Planar Exposure MAPLHCR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 11.6 2171 0.032 1,000 11.6 2178 0.033
. 5,000 12.0 2198 0.034 4
10,000 12.1 2195 0.033 15,000 12.0 2200 0.033 20,000 11.7 2187 0.032 25,000 11.0 2115 0.025 30,000 10.5 2029 0.019 40,000 9.3 1913 0.013 45,000 8.6 1840 0.010 ~
, 50,000 8.0 1785 0.009 55,000 7.3- 1730 0.007 .
4-10
NEDO-24146A 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 t.he fuel rods.
Table 4R MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Quad Cities 1,2 FUEL TYPE: BP8DRB282 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /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 30,000 10.8 2047 0.038 35,000 10.3 1957 0.031 40,000 9.6 1840 0.009 45,000 8.9 1761 0.007 NOTE: Credit taken for the effects of pre-pressurization of the fuel rode. .
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NEDO-24146A Table 4S MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Quad Cities 1&2 FUEL TYPE: BP8DRE299_
Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 2d0 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 ef fects of pre-pressurization of the fuel rods. .
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