ML16256A128
| ML16256A128 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 08/25/2016 |
| From: | Entergy Operations |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16256A115 | List:
|
| References | |
| W3F1-2016-0053 | |
| Download: ML16256A128 (6) | |
Text
WSES-FSAR-UNIT-3 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCE The following topical reports are incorporated by reference.
Report FSAR Number Author and Title Date to NRC Section CENPD-26 Combustion Engineering, Inc.
August 1971 3.6, 6.2 (with Suppl.
"Description of Combustion 1 through 5)
Engineering Loss of Coolant Calculational Procedures" CENPD-42 Combustion Engineering, Inc.
August 1972 3.6 "Dynamic Analysis of Reactor Vessel Internals Under Loss of Coolant Accident Conditions with Application to C-E 800 Mwe Class Reactors" CENPD-67 Combustion Engineering, Inc.
September 1973 10.3 (with Suppl.
"Iodine Decontamination 1 and 2 and Factors During PWR Steam Addenda 1 Generation and Steam Venting" and 2)
CENPD-87 "Safety Related Research Development 1.5 for CE PWRs, Program Summaries" CENPD-98 Combustion Engineering, Inc.
July 1973 4.4, 15.0 "Coast Code Description" CENPD-105 Combustion Engineering, Inc.
June 1973 4.3 "Fast Neutron Attenuation by the ANISN-SHADRAC Analytical Method" CENPD-107 Combustion Engineering, Inc.
August 1974 15.0 (with Suppl.
"CESEC" 1 through 4)
CENPD-118 Combustion Engineering, Inc.
September 1974 4.3 "Densification of Combustion Engineering Fuel" CENPD-132 Combustion Engineering, Inc.
September 1974 6.2, 15.6 (with Suppl.
"Calculative Methods for the 6.3 1 and 2)
C-E Large Break LOCA Evaluation Model"
WSES-FSAR-UNIT-3 1.6-2 Revision 11-A (02/02)
Report FSAR Number Author and Title Date to NRC Section CENPD-133 Combustion Engineering, Inc.
September 1974 6.2, 15.6 (with "CEFLASH-4A Fortran IV 6.3 Suppl. 2)
Digital Computer Program for Reactor Blowdown Analysis" CENPD-134 Combustion Engineering, Inc.
September 1974 6.2, 15.6 (with "COMPERC-II A Program for 6.3 Suppl. 1)
Emergency Refill-Reflood of the Core" CENPD-135 Combustion Engineering, Inc.
September 1974 4.2, 6.3, (with "STRIKIN-II A Cylindrical 15.6 Suppl.
Geometry Fuel Rod Heat 2 and 4)
Transfer Program" CENPD-136 Combustion Engineering, Inc.
September 1974 4.2, 15.6 "High Temperature Properties of Zircaloy and UO2, for use in LOCA Evaluation Model" CENPD-137 Combustion Engineering, Inc.
September 1974 6.3, 15.6 (with "Calculative Methods for the Suppl. 1)
C-E Small Break LOCA Evaluation Model"
¨(DRN 01-758)
CENPD-138 "PARCH, A FORTAN IV Digital Program February 1975 15.6 to Evaluate Pool Boiling, Axial Rod and Coolant Heatup" (with Supplement 1)
CENPD-139 Combustion Engineering, Inc.
September 1974 4.1, 4.2, (with "C-E Fuel Evaluation Model" 4.3, 6.3 Suppl. 1)
CENPD-145 Combustion Engineering, Inc.
April 1975 4.3, 7.7 "A Method of Analyzing In-Core Detector Data in Power Reactors" CENPD-148 Combustion Engineering, Inc.
November 1974 4.6 "Review of Reactor Shutdown System (PPS Design) for Common Mode Failure Susceptibility" CENPD-153 Combustion Engineering, Inc.
August 1974 4.3 "Evaluation Uncertainty in the Nuclear Form Factor
WSES-FSAR-UNIT-3 Report FSAR Number Author and Title Date to NRC Section 1.6-3 CENPD-153 Measured by Self Powered (cont)
Fixed In-Core Detector Systems" CENPD-155 Combustion Engineering, Inc.
October 1974 5.3 "C-E Procedure for Design, Fabrication, Installation and Inspection of Surveillance Specimen Brackets Attached to Reactor Vessel Beltline Region" CENPD-161 Combustion Engineering, Inc, June 1975 4.1, 4.2, "TORC - A Computer Code for 4.4, 15.0 Determining the Thermal Margin of a Reactor Core" CENPD-162 Combustion Engineering, Inc.
May 1975 4.4 (with "CHF Correlation for C-E Fuel Suppl. 1)
Assemblies with Standard Spacer Grids - Part 1; Uniform Axial Power Distribution" CENPD-168 Combustion Engineering, Inc.
September 1976 3.6, 6.2 Rev. 1 "Design Basis Pipe Breaks for the Combustion Engineering Two Loop Reactor Coolant System" CENPD-169 Combustion Engineering, Inc.
August 1975 4.3, 7.2 "Assessment of the Accuracy 7.7 of PWR Operating Limits as Determined by Core Operating Limit Supervisory System" CENPD-170 Combustion Engineering, Inc.
August 1975 7.2 "Assessment of the Accuracy of the PWR Safety System Actuation as Performed by the Core Protection Calculators" CENPD-178P Combustion Engineering, Inc.
October 1976 3.9, 4.2 and 178 "Structural Analysis of the 16 x 16 Fuel Assembly for Combined Seismic and Loss-of-Coolant-Accident Loadings"
WSES-FSAR-UNIT-3 Report FSAR Number Author and Title Date to NRC Section 1.6-4 CENPD-179 Combustion Engineering, Inc.
April 1976 4.2 "C-E Thermo-Structural Fuel Evaluation Method" CENPD-183 Combustion Engineering, Inc.
August 1975 15.0 "C-E Methods for Loss of 15.3 Flow Analysis" CENPD-187 Combustion Engineering, Inc.
October 1975 4.2 (with "Method of Analyzing Creep Suppl. 1)
Collapse of Oval Cladding" CENPD-190 Combustion Engineering, Inc.
January 1976 15.4 "C-E Method for Control Element Assembly Ejection Analysis" CENPD-198P Combustion Engineering, Inc.
December 1975 4.2 and 198 "Zircaloy Growth-In-Reactor Dimensional Changes in Zircaloy-4 Fuel Assemblies" CENPD-206 Combustion Engineering, Inc.
December 1976 4.4 "Comparison of TORC Code Predictions with Experimental Data" CENPD-207 Combustion Engineering, Inc.
June 1976 4.4 "Critical Heat Flux Corre-lation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, Non-Uniform Axial Power Distributions" CENPD-213 Combustion Engineering, Inc.
February 1976 6.3, 15.6 "Application of FLECHT Reflood Heat Transfer Coefficients to Combustion Engineering 16 x 16 Fuel Bundles" CENPD-225P Combustion Engineering, Inc.
October 1976 4.2, 4.4 and 225 "Fuel and Poison Rod Bowing" CENPD-252 "Method for the Analysis of Blowdown July 1979 3.9E
WSES-FSAR-UNIT-3 1.6-5 Revision 304 (06/10)
Report Number Author and Title Date to NRC FSAR Section WCAP 7709-L Electric Hydrogen Recombiners for PWR Containments April 1972 6.2.5
(DRN 03-2054, R14)
CEN-367-A Combustion Engineering, Inc.
Leak-Before-Break of Primary Coolant Loop Piping in Combustion Engineering Designed Nuclear Steam Supply System February 1991 3.6
(DRN 03-2054, R14)
(EC-13881, R304)
WCAP-11596-P-A Westinghouse Electric Company, Qualification of the PHOENIX -
P/ANC Nuclear Design System for Pressurized Water Reactor Cores June 1988 4.2, 4.3A, 15.1 WCAP-10965-P-A Westinghouse Electric Company, ANC: A Westinghouse Advanced Nodal Computer Code September 1986 4.2, 4.3A, 15.1 WCAP-10965-P-A Addendum 1 Westinghouse Electric Company, ANC: A Westinghouse Advanced Nodal Computer Code:
Enhancements to ANC Rod Power Recovery April 1989 4.2, 4.3A, 15.1 WCAP-16045-P-A Westinghouse Electric Company, Qualification of the Two-Dimensional Transport Code
- PARAGON, August 2004 4.3A WCAP-16072-P-A Westinghouse Electric Company, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs August 2004 4.2, 4.3A CENPD-404-P-A Combustion Engineering, Inc.,
Implementation of ZIRLO Material Cladding in CE Nuclear Power Fuel Assembly Designs, November 2001 4.2, 4.3A, 15.0 WCAP-16500-P-A Westinghouse Electric Company, CE 16 x 16 Next Generation Fuel Core Reference Report, August 2007 4.2, 4.3A WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A Westinghouse Electric Company, Optimized ZIRLOTM, July 2006 4.2, 4.3A, 15.0
EC-13881, R304)
WSES-FSAR-UNIT-3 1.6-6 Revision 305 (11/11)
(EC-13881, R304)
Report Number Author and Title Date to NRC FSAR Section WCAP-16523-P-A Westinghouse Electric Company, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing
- Vanes, August 2007 4.4, 15.0, 15.1, 15.2, 15.3, 15.4 CENPD-387-P-A Combustion Engineering, Inc.,
ABB Critical Heat Flux Correlations for PWR Fuel, May 2000 4.3A, 15.0 WCAP-15996-P-A, Rev. 1 Westinghouse Electric Company, Technical Manual for the CENTS
- Code, March 2005 15.0, 15.1, 15.2, 15.3, 15.4 CEN-356(V)-P-A, Revision 01-P-A Combustion Engineering, Inc.,
Modified Statistical Combination of Uncertainties, May 1988 4.2, 4.3A (EC-13881, R304)
(EC-19087, R305)
WCAP-17817-P Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for Waterford Steam Electric Station, Unit 3 Using Leak-Before-Break Methdology February 2010 3.6.3 (EC-19087, R305)