ML23291A373
| ML23291A373 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/12/2023 |
| From: | Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23291A105 | List:
|
| References | |
| PLA-8081 | |
| Download: ML23291A373 (1) | |
Text
SSES-FSAR 1.6 MATERIAL INCORPORATED BY REFERENCE Within each subsection of the text, applicable supporting technical material is referenced.
References are cited in a list of references at the end of a section. Most of the references are cited as a particular technical basis for BWR plant design and analysis. However, a few of the references cited in the FSAR refer to technical work performed by GE or Bechtel which is specifically applicable to the Susquehanna SES. The references in this category generally provide a full development and analysis of some aspect of plant technology. These special references are incorporated by reference into this FSAR, thereby becoming part of the license application.
1.6.1 GENERAL ELECTRIC TOPICAL REPORTS Table 1.6-1 is a list of all GE topical reports and other GE reports which are incorporated in whole or in part by reference in this FSAR, and have been filed with the NRC.
1.6.2 BECHTEL TOPICAL REPORTS Table 1.6-2 is a list of Bechtel topical reports which have been incorporated in whole or in part by reference in this FSAR.
1.6.3 OTHER TOPICAL REPORTS Table 1.6-4 is a list of other reports which have been incorporated in whole or in part by reference in this FSAR and which have been filed with the NRC.
Rev. 53, 04/99 1.6-1
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION APED-4824 Maximum Two-Phase Vessel Blowdown from Pipes 6.2 (April 1965)
APED-5286 Design Basis for Critical Heat Flux Condition in BWRs (September 1.5 1966)
APED-5458 Effectiveness of Core Standby Cooling Systems for General Electric 5.4 Boiling Water Reactors (March 1968)
APED-5460 Design and Performance of General Electric BWR Jet Pumps 3.9 (July 1968)
APED-5555 Impact Testing on Collet Assembly for Control Rod Drive 4.6 Mechanism 7RDB144A (November 1967)
APED-5640 Xenon Considerations in Design of Large Boiling Water Reactors 4.1,4.3 (June 1968)
APED-5652 Stability and Dynamic Performance of the General Electric Boiling 4.1 Water Reactor APED-5706 In-Core Neutron Monitoring System for General Electric Boiling 7.6, 7.7 Water Reactors 7.6.2a.5 (November 1968, Revised April 1969)
APED-5736 Guidelines for Determining Safe Test Intervals and Repair Times for Appendix Engineered Safeguards (April 1969)
APED-5750 Design and Performance of General Electric Boiling Water Reactor 5.4 Main Steam Line Isolation Valves (March 1969)
APED-5756 Analytical Methods for Evaluating the Radiological Aspects of the 15.4 General Electric Boiling Water Reactor (March 1969)
GEAP-10546 Theory Report for Creep-Plast Computer Program 4.1 (January 1972)
NEDE-10313 PDA-Pipe Dynamic Analysis Program for Pipe Rupture Movement 3.6 (Proprietary Filing)
NEDE-11146 Design Basis for New Gas System 11.3 (July 1971) (Company Proprietary)
FSAR Rev. 63 Page 1 of 8
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION NEDO-20944-(P) (A) BWR/4 and BWR/5 Fuel Design 4.2 (October 1976)
NEDE-21156 Supplemental Information for Plant Modification to Eliminate 4.4 Significant In-Core Vibration (January 1976)
NEDE-21175-3-P BWR/6 Fuel Assembly Evaluation of Combined Safe 3.9 Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings (Amendment No. 3)
(July 1982)
NEDE-21354-P BWR Fuel Channel Mechanical Design and Deflection 3.9 (September 1976)
NEDE-22290-A, Safety Evaluation of the General Electric Hybrid 1 4.2 Supp. 1 Control Rod Assembly for the BWR 4/5 C Lattice (July 1985)
NEDE-22290-A, Safety Evaluation of the General Electric Duralife 230 4.2 Supp. 3 Control Rod Assembly (May 1988)
NEDE-23014 HEX 01 User's Manual 15.2 (July 1976)
NEDE-31096-A Anticipated Transients Without Scram Response to the 7.1, 7.2 NRC ATWS Rule 10CFR50.62 (December 1985)
NEDO-10173 Current State of Knowledge, High Performance BWR 11.1 Zircaloy-Clad UO2 Fuel (May 1970)
NEDO-10299 Core Flow Distribution in a Modern Boiling Water Reactor as 4.4 Measured in Monticello (January 1971)
NEDO-10320 The General Electric Pressure Suppression Containment 6.2 Analytical Model (April 1971)
Supplement 1 (May 1971)
NEDO-10349 Analysis of Anticipated Transients Without Scram 15.8 (March 1971)
FSAR Rev. 63 Page 2 of 8
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION NEDO-10466 Power Generation Control Complex Design Criteria and 3.12.3.4.2.1 (f)
Safety Evaluation (February 1972)
NEDO-10505 Experience with BWR Fuel Through September 1971 11.1 (May 1972)
NEDO-10527 Rod Drop Accident Analysis for Large Boiling Water Reactors 15.4 (March 1972) Supplement 1 (July 1972) Supplement 2 (January 1973)
NEDO-10585 Behavior of Iodine in Reactor Water During Plant Shutdown 15.6 and Startup (August 1972)
NEDO-10602 Testing of Improved Jet Pumps for the BWR/6 Nuclear 3.9 System (June 1972)
NEDO-10734 A General Justification for Classification of Effluent 11.3 Treatment System Equipment as Group D (February 1973)
NEDO-10739 Methods for Calculating Safe Test Intervals and 6.3 Allowable Repair Times for Engineered Safeguard Systems (January 1973)
NEDO-10751 Experimental and Operational Confirmation of Offgas 11.3 System Design Parameters (January 1973) (Company Proprietary)
NEDO-10801 Modeling the BWR/6 Loss-of-Coolant Accident: Core 1.5 Spray and Bottom Flooding Heat Transfer Effectiveness (March 1973)
NEDO-10802 Analytical Methods of Plant Transient Evaluations for 4.4, 5.2, 15.1, General Electric Boiling Water Reactor 15.2, 15.3, (February 1973) 15.4, 15.5 NEDO-10846 BWR Core Spray Distribution 1.5 (April 1973)
NEDO-10899 Chloride Control in BWR Coolants 5.2 (June 1973)
FSAR Rev. 63 Page 3 of 8
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION NEDO-10958 General Electric BWR Thermal Analysis Basis (GETAB): 4.3, 4.4, 15.0 Data, Correlation, and Design Application (November 1973)
NEDO-10958-A General Electric BWR Thermal Analysis Basis (GETAB): 1.5, 15.4, 16.1 Data, Correlation, and Design Application (January 1977)
NEDO-10959 General Electric BWR Thermal Analysis Basis (GETAB): 15.0 Data, Correlation, and Design Application (November 1973)
NEDO-20231 Emergency Core Cooling Tests of an Internally 1.5 Pressurized, Zircaloy-Clad, 8 x 8 Simulated BWR Fuel Bundle (December 1973)
NEDO-20360 General Electric Boiling Water Reactor Generic Reload 15.4 Application for 8x8 Fuel (May 1975)
NEDO-20533 The General Electric Mark III Pressure Suppression Containment System Analytical Model (June 1974)
NEDO-20566 General Electric Company Model for Loss-of-Coolant 3.9 Accident Analysis in Accordance with 10 CFR 50, Appendix K (January 1976)
NEDO-20626 Studies of BWR Designs for Mitigation of Anticipated 15.8 Transients without Scrams (October 1974)
NEDO-20626-1 Studies of BWR Designs for Mitigation of Anticipated 15.8 Transients without Scrams (June 1975)
NEDO-20626-2 Studies of BWR Designs for Mitigation of Anticipated 15.8 Transients without Scrams (July 1975)
NEDO-20631 Mechanical Property Surveillance of Reactor Pressure 5.3 Vessels for General Electric BWR/6 Plants (March 1975)
FSAR Rev. 63 Page 4 of 8
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION NEDO-20922 Experience with BWR Fuel Through September 1974 11.1 (June 1975)
NEDE-20944 BWR/4 and BWR/5 Fuel Design 4.3 (October 1976)
NEDO-20953 Three-Dimensional Boiling Water Reactor Core Core 15.4 Simulator (May 1976)
NEDO-21142 Realistic Accident Analysis for General Electric Boiling 15.4, 15.6, 15.7, Water Reactor - The RELAC Code and User's Guide, to 15.2 be issued (December 1977)
NEDO-21143 Conservative Radiological Accident Evaluation - The 15.4, 15.6, CNAC01 Code 15.7 (March 1976)
NEDO-21159 Airborne Release from BWRs for Environment Impact 11.1 Evaluations (March 1976)
NEDO-24548 Technical Description Annulus Pressurization Load 6A Adequacy Evaluation NEDO-21291 Group Notch Mode of the RSCS for Cooper 15.4 (June 1976)
NEDO-26453 Oyster Creek Station, FSAR Amendment 10 1.5 "Summary Memorandum on Excursion Analysis 4.3, 15.0 Uncertainties," Dresden Nuclear Power Station, Unit 3, Plant Design Analysis Report Amendment 3 Hatch Nuclear Plant, Unit 1, PSAR 15.5 Amendment 10, Appendix L Millstone Nuclear Power Station, PSAR 6.3 Amendment 14 Pilgrim Nuclear Power Station, 6.3 PSAR, Amendment 14 APED-4827 F. J. Moody, Maximum Two-Phase Vessel Blowdown from 6.2 Pipes, Topical Report, General Electric Company, (1965)
FSAR Rev. 63 Page 5 of 8
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION NEDC-32071P Diefenderfer, S. B., and D. C. Pappone, Susquehanna 5.4, 6.3 Steam Electric Station Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, General Electric Nuclear Engineering Department NEDC-32161P December 1993, (General Electric Report), Power Uprate 5.2, 5.3, 10.2, Engineering Report for Susquehanna Steam Electric Station 11.1, 15.6 Units 1 and 2 NEDC-32281P SAFER/GESTR-LOCA Analysis Basis Documentation for 6.3 Susquehanna Steam Electric Station Units 1 and 2, September 1993 NEDE-24131 Basis for 8x8 Retrofit Fuel Thermal Analysis Application, 4.4 General Electric Company, September 1978, Proprietary NEDE-10813 PDA-Pipe Dynamic Analysis Program for Pipe Rupture 3.6 Movement, Proprietary Filing NEDE-21544P R. J. Ernst and M. G. Ward, Mark II Pressure Suppression 6.2 Containment Systems: An Analytical Model of the Pool Swell Phenomenon, General Electric Co., December 1976 NEDE-23785-1-PA The GESTR-LOCA and SAFER Models for the Evaluation of 6.3 the Loss-of-Coolant, Accident, Volume III, SAFER/GESTR Application Methodology, Revision 1, October 1984 NEDE-24057-P Assessment of Reactor Internals Vibration in BWR/4 and 3.9 BWR/5 Plants, (Class III) and NEDO-24057 (Class I),
November 1977 NEDE-24154-P F. Odar, Safety Evaluation for General Electric Topical 5.2, 15.1, 15.2, Report: Qualification of One-Dimensional Core Transient 15.5 Model for Boiling Water Reactors, also NEDO-24154, Vols. I, II, III, dated June 1980 NEDM-23842 R. C. Stirn and J. F. Klapproth, Continuous Control Rod 15.4 Withdrawal Transient in the Startup Range, April 18, 1978 NEDO-10722A Core Flow Distribution in a Modern Boiling Water Reactor as 4.4 Measured in Moniticello, August 1976 NEDO-21143-1 Radiological Accident Evaluation - The CONOC03 15.4, 15.6, 15.7 Computer Code FSAR Rev. 63 Page 6 of 8
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION NEDO-21231 C. J. Paone, Bank Position Withdrawal Sequence, 15.4 September 1976 NEDO-21506 Stability and Dynamic Performance of the General Electric 4.4 Boiling Water Reactor, January 1977 NEDO-21985 Functional Capability Criteria for Essential Mark II Piping, 78 3.9 NED174 (Class I), September 1978 CHASTE05 Letter, A. J. Levine (GE) to D. F. Ross (NRC) dated 6.3 January 27, 1977, General Electric (GE) Loss-of-Coolant Accident (LOCA) Analysis Model Revisions - Core Heatup Code CHASTE05 REFLOOD05 Letter, A. J. Levine (GE) to D. B. Vassallo (NRC) dated 6.3 March 14, 1977, Request for Approval for Use of Loss-of-Coolant Accident (LOCA) Evaluations Model Code REFLOOD05 NEDO-20566A General Electric Company Analytical Model for Loss-of- 6.3 Coolant Analysis in Accordance with 10CFR50, Appendix K, September 1986 NEDE-32417P GE 12 Compliance with Amendment 22 of NEDE-24011-P-A 3.9 (GESTAR 11), GE Nuclear Energy, December 1994 NEDE-22290-A Supplement 1, Safety Evaluation of the General Electric 4.2 Hybrid I Control Rod Assembly for the BWR 4/5 C Lattice, General Electric Company, July 1985 NEDE-22290-A Supplement 3, Safety Evaluation of the General Electric 4.2 Duralife 230 Control Rod Assembly, General Electric Company, May 1988 NEDE-31758P-A GE Marathon Control Rod Assembly GE Nuclear Energy, 4.1, 4.2 October 1991 NEDO-31960-A "BWR Owners' Group Long-Term Stability Solutions 3.1, 7.1 Licensing Methodology" NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions 3.1, 7.1 Supplement 1 Licensing Methodology (Supplement 1)"
NEDO-32465-A Reactor Stability Detect and Suppress Solution 3.1, 4.1, 7.1 Licensing Basis Methodology for Reload Applications" FSAR Rev. 63 Page 7 of 8
SSES-FSAR NIMS Rev. 58 TABLE 1.6-1 GENERAL ELECTRIC TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION NEDC-32410P-A Nuclear Measurement Analysis and Control Power Range 3.1, 7.6 Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip Function, October 1995.
NECD-32410P-A Nuclear (Measurement Analysis and Control Power Range 3.1, 7.6 Supplement 1 Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip Function, Supplement 1, November 1997.
FSAR Rev. 63 Page 8 of 8
SSES-FSAR TABLE 1.6*2 BECHTEL TOPICAL REPORTS (1) BC-TOP-1, Rev. 1 (12-72),
3.8 Containment Building Liner Plate Design (2) BC-TOP-4-A, Rev. 3 (11-74)
Seismic Analysis of Structures and Equipment for Nuclear 3.7B Power Plants (3) BC-TOP-9-A, Rev, 2 (9-74) 3.6 Design of Structures for Missile Impact (4) BN-TOP-1, Rev. 1 (11-72)
Testing Criteria for Integrated Leak Rate Testing of Primary 3.9 Containment Structures for Nuclear Power Plants (5) BN-TOP-2, Rev. 2 (5-74) 3.6 Design for Pipe Break Effects (6) BP-TOP-1, Rev. 2 (1-75) 3.7B Seismic Analysis Piping Systems (Except as noted on 3.9 Table 3.9-15)
(7) BN-TOP-4, Rev. 1 (10-77)
Subcompartment Pressure and Temperature Transient 6B Analysis. (This report was approved by the NRC in February, 1979.
Rev. 54, 10/99 Page 1 of 1
SSES-FSAR Table Rev. 64 TABLE 1.6-4 OTHER TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION AE-RTL-788 Void Measurements in the Region of Subcooled and Low 4.4 Quality Boiling (April 1966)
ANL-5621 Boiling Density in Vertical Rectangular Multichannel 4.4 Sections with Natural Circulation (November 1956)
ANL-6385 Power-to-Void Transfer Functions 4.4 (July 1961)
BHR/DER 70-1 Radiological Surveillance Studies at a Boiling Water 11.1 Nuclear Power Reactor (March 1970)
BMI-1163 Vapor Formation and Behavior in Boiling Heat Transfer 4.4 (February 1957)
CF 59-6-47 (ORNL) Removal of Fission Product Gases From Reactor Off Gas 11.3 Streams by Adsorption (June 11, 1959)
ST1-372-38 Kinetic Studies of Heterogeneous Water Reactors 4.4 (April 1966)
TID-4500 Relap 3 - A Computer Program for Reactor Blowdown 3.6 Analysis IN-1321 (June 3970)
UCRL-50451 Improving Availability and Readiness of Field Equipment 18.3 Through Periodic Inspection, p. 10 (July 16, 1968)
WAPD-BT-19 A Method of Predicting Steady-Boiling Vapor Fractions in 4.4 Reactor Coolant Channels (June 1960)
NE-092-001A SSES Power Uprate Licensing Topical Report, and NRC 10.2, 15.6 letter dated November 30, 1993, from Thomas E. Murley to Robert G. Byram (PP&L).
Subject:
Licensing Topical Report for Power Uprate with Increased Core Flow, Rev. 0, Susquehanna Steam Electric Station, Units 1 and 2 (PLA-3788) (TAC NOS. M83426 and M83427) with enclosed Safety Evaluation Report PL-NF-89-005-A Qualification of Transient Analysis Methods for BWR 15.1 Design and Analysis, Issue Date: July 1992.
FSAR Rev. 71 Page 1 of 3
SSES-FSAR Table Rev. 64 TABLE 1.6-4 OTHER TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION XN-NF-80-19(P)(A) Volume 1, Supplement 3, Advanced Nuclear Fuels 15.1 Methodology for Boiling Water Reactors-Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology, EXXON Nuclear Company, Richland, WA 99352, November 1990 XN-NF-80-19(P)(A) Exxon Nuclear Methodology for Boiling Water Reactors; 15.0, 15.4 Neutronic Methods for Design and Analysis, Volume 1, and Volume 1 Supplements 1 and 2, March 1983.
XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, 15.0, 15.3 XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, February 1987 XN-NF-84-105(P)(A) Volume 1 Supplement 4, XCOBRA-T: A Computer Code 15.3 for BWR Transient Thermal-Hydraulic Core Analysis, Void Fraction Model Comparison to Experimental Data, June 1988 XN-NF-80-19(P)(A) Volume 3, Revision 2, Exxon Nuclear Methodology for 4.1, 15.0 Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Siemens Power Corporation, January 1987 XN-NF-79-59(P)(A) Methodology for Calculation of Pressure Drop in BWR 4.1 Fuel Assemblies, November 1983 EMF-CC-074(P)(A) Volume 1, 2 and 4 STAIF - A Computer Program for 4.1, 4.4 BWR Stability Analysis in the Frequency Domain ANF-89-98(P)(A) Rev. 1 and Rev. 1 Supplement 1, Generic Mechanical 4.2 Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995 CENPD-400-P-A "Generic Topical Report for the ABB Option III Oscillation 3.1, 4.4, 7.6 Power Range Monitor (OPRM)"
EMF-93-177 (P)(A), Mechanical Design for BWR Fuel Channels Siemens 4.2 SUPPLEMENTS Power Corporation, August 2005 1&2 EMF-2209 (P)(A) SPCB Critical Power Correlation, September 2003 4.1, 4.4 EMF-2158(P)(A) Siemens Power Corporation Methodology For Boiling 4.1, 4.3, 4.4, Water Reactors - Evaluation and Validation of 15.4, 15.5 CASMO-4/Microburn-B2 Rev. 0, October 1999 XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for 5.2, 15.0, 15.1, Boiling Water Reactors: Application of the ENC 15.2, 15.4, 15.5 Methodology to BWR Reloads, Exxon Nuclear Company, June 1986 FSAR Rev. 71 Page 2 of 3
SSES-FSAR Table Rev. 64 TABLE 1.6-4 OTHER TOPICAL REPORTS REPORT REFERENCED IN TITLE NUMBER FSAR SECTION EMF-2361(P)(A) EXEM BWR-2000 ECCS Evaluation Model, Framatome 6.3 ANP, May 2001 ANF-1358(P)(A) The Loss of Feedwater Heating Transient in Boiling 15.1 Water Reactors, Advanced Nuclear Fuels Corporation, September 2005 XN-NF-85-74(P)(A) RODEX2A(BWR) Fuel Thermal-Mechanical Evaluation 4.1 Model, Exxon-Nuclear Company, Inc,. August 1986 BAW-10247(P)(A) & Realistic Thermal Mechanical Fuel Rod Methodology for 4.1, 4.2, 4.3, 6.3 SUPPLEMENT 2P-A Boiling Water Reactors ANP-10335(P)(A) ACE/ATRIUM 11 Critical Power Correlation 4.1, 4.4 ANP-10307(P)(A) AREVA MCPR Safety Limit Methodology for Boiling 4.1, 4.4 Water Reactors ANP-10300P-A AURORA-B: An Evaluation Model for Boiling Water 5.2, 15.0, 15.1, Reactors; Application to Transient and Accident 15.2, 15.3, 15.5, Scenarios, January 2018 15.8 ANP-10332P-A AURORA-B: An Evaluation Model for Boiling Water 6.3 Reactors; Application to Loss of Coolant Accident Scenarios.
XN-NF-82-07(P)(A), Exxon Nuclear Company ECCS Cladding Swelling and 6.3 Rev. 1 Rupture Model, November 1982 ANP-10333P-A AURORA-B: An Evaluation Model for Boiling Water 15.4 Reactors; Application to Control Rod Drop Accident (CRDA), March 2018 FSAR Rev. 71 Page 3 of 3