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The references listed on this page represent open-literature citations related to SAS4A/SASSYS-1 code modeling capabilities and developments. It does not include references to modeling and simulation analyses in which SAS4A/SASSYS-1 was used to evaluate safety performance.
== General References ==
== General References ==
* D. R. Ferguson et al., “The SAS4A LMFBR Accident Analysis Code System: A Progress Report,” ''Proceedings of the International Meeting on Fast Reactor Safety and Related Physics'', American Nuclear Society, Chicago, IL, October 5 8, 1976.
* J. E. Cahalan et al., “The Status and Experimental Basis of the SAS4A Accident Analysis Code System,” ''Proceedings of the International Meeting on Fast Reactor Safety Technology'', American Nuclear Society, Seattle, WA, August 19 23, 1979.
* H. U. Wider et al., “Status and Validation of the SAS4A Accident Analysis Code System,” ''Proceedings of the LMFBR Safety Topical Meeting'', European Nuclear Society, Lyon, France, July 19 23, 1982.
* A. M. Tentner et al., “The SAS4A LMFBR Whole Core Accident Analysis Code,” ''Proceedings of the International Topical Meeting on Fast Reactor Safety'', American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* A. M. Tentner et al., “SAS4A: A Computer Model for the Analysis of Hypothetical Core Disruptive Accidents in Liquid Metal Reactors,” ''1987 SCS Eastern Simulation Conference'', Society for Computer Simulation, Orlando, FL, April 6 9, 1987.
* A. M. Tentner et al., “Simulating Unprotected Accidents for Advanced Liquid Metal Reactors Using the SAS4A Accident Analysis Code,” ''1988 SCS Simulators Conference'', Society for Computer Simulation, Orlando, FL, April 18 21, 1988.
* J. E. Cahalan and T. Wei, “Modeling Developments for the SAS4A and SASSYS Computer Codes,” ''Proceedings of the International Fast Reactor Safety Meeting'', American Nuclear Society, Snowbird, UT, August 12-16, 1990.
* J. E. Cahalan et al., “Advanced LMR Safety Analysis Capabilities in the SASSYS-1 and SAS4A Computer Codes,” ''Proceedings of the International Topical Meeting on Advanced Reactors Safety'', American Nuclear Society, Pittsburgh, PA, April 17-21, 1994.


* F. E. Dunn and F. G. Prohammer, “SASSYS Analysis of Degraded Shut Down Heat Removal Performance in LMFBRs,” ASME Paper No. 82 WA/HT 37, 1982.
* F. E. Dunn and F. G. Prohammer, “SASSYS Analysis of Degraded Shut Down Heat Removal Performance in LMFBRs,” ASME Paper No. 82 WA/HT 37, 1982.
* F. E. Dunn and F. G. Prohammer, “The SASSYS LMFBR Systems Analysis Code,” Proceedings of the 10th IMACS World Conference on Systems Simulation and Scientific Computation, Vol. 4, Montreal, Canada, pp. 127-129, August, 1982.
* F. E. Dunn and F. G. Prohammer, “The SASSYS LMFBR Systems Analysis Code,” ''Proceedings of the 10th IMACS World Conference on Systems Simulation and Scientific Computation'', Vol. 4, Montreal, Canada, pp. 127-129, August, 1982.
* F. E. Dunn et al., “The SASSYS 1 LMFBR Systems Analysis Code,” Proceedings of the International Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* F. E. Dunn et al., “The SASSYS 1 LMFBR Systems Analysis Code,” ''Proceedings of the International Topical Meeting on Fast Reactor Safety'', American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* D. K. Warinner and F. E. Dunn, “SASSYS 1 Computer Code Verification with EBR II Test Data,” Proceedings of the International Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* D. K. Warinner and F. E. Dunn, “SASSYS 1 Computer Code Verification with EBR II Test Data,” ''Proceedings of the International Topical Meeting on Fast Reactor Safety'', American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* F. E. Dunn, “The SAS4A/SASSYS l Sodium Boiling Model for LMFBR Whole Core Analysis,” Heat Transfer - Denver 1985, AIChE Symposium Series, No. 245, Vol. 81, 1985.
* F. E. Dunn, “The SAS4A/SASSYS l Sodium Boiling Model for LMFBR Whole Core Analysis,” ''Heat Transfer - Denver 1985'', AIChE Symposium Series, No. 245, Vol. 81, 1985.
* F. E. Dunn et al., “LMR Thermal Hydraulics Calculations in the U.S.,” Proceedings of the International Topical Meeting on Advances in Reactor Physics, Mathematics and Computation, Paris, France, April 27 30, 1987.
* F. E. Dunn et al., “LMR Thermal Hydraulics Calculations in the U.S.,” ''Proceedings of the International Topical Meeting on Advances in Reactor Physics'', Mathematics and Computation, Paris, France, April 27 30, 1987.
* D. J. Hill, “SASSYS Analysis of EBR-II SHRT Experiments,” Trans. Am. Nucl. Soc., 55, 421, 1987.
* D. J. Hill, “SASSYS Analysis of EBR-II SHRT Experiments,” ''Trans. Am. Nucl. Soc.'', 55, 421, 1987.
* F. E. Dunn, “LMR Thermal Hydraulics Calculations in the U.S.,” Nucl. Sci. Eng., 100, 558, 1988.
* F. E. Dunn, “LMR Thermal Hydraulics Calculations in the U.S.,” ''Nucl. Sci. Eng.'', 100, 558, 1988.
* F. E. Dunn and T. Y. C. Wei, “Simulating Operational Transients with the SASSYS 1 LMR Systems Analysis Code,” 1988 SCS Simulators Conference, Society for Computer Simulation, Orlando, FL, April 18-21, 1988.
* F. E. Dunn and T. Y. C. Wei, “Simulating Operational Transients with the SASSYS 1 LMR Systems Analysis Code,” ''1988 SCS Simulators Conference'', Society for Computer Simulation, Orlando, FL, April 18-21, 1988.
* F. E. Dunn and T. Y. C. Wei, “The Role of SASSYS 1 in LMR Safety Analysis,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
* F. E. Dunn and T. Y. C. Wei, “The Role of SASSYS 1 in LMR Safety Analysis,” ''Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors'', American Nuclear Society, Seattle, WA, May 1 5, 1988.
* D. J. Hill, “SASSYS Validation Studies,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
* D. J. Hill, “SASSYS Validation Studies,” ''Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors'', American Nuclear Society, Seattle, WA, May 1 5, 1988.
* F. E. Dunn, “Decay Heat Calculations for Transient Analysis,” Trans. Am. Nucl. Soc., 60, 633, 1989.
* F. E. Dunn, “Decay Heat Calculations for Transient Analysis,” ''Trans. Am. Nucl. Soc.'', 60, 633, 1989.
* J. P. Herzog, “SASSYS Validation with the EBR II Shutdown Heat Removal Tests,” Trans. Am. Nucl. Soc., 60, 730, 1989.
* J. P. Herzog, “SASSYS Validation with the EBR II Shutdown Heat Removal Tests,” ''Trans. Am. Nucl. Soc.'', 60, 730, 1989.
* F. E. Dunn and J. P. Herzog, “Thermal-Hydraulic Impact of Failure of Highly Irradiated Fuel Pins on LMR Passive Safety,” Trans. Am. Nucl. Soc., 62, 673, 1990.
* F. E. Dunn and J. P. Herzog, “Thermal-Hydraulic Impact of Failure of Highly Irradiated Fuel Pins on LMR Passive Safety,” ''Trans. Am. Nucl. Soc.'', 62, 673, 1990.
* F. E. Dunn, “Consequences of Pipe Ruptures in Metal Fueled, Liquid Metal Cooled Reactors,” Proceedings of the International Fast Reactor Safety Meeting, American Nuclear Society, Snowbird, UT, August 12-16, 1990.
* F. E. Dunn, “Consequences of Pipe Ruptures in Metal Fueled, Liquid Metal Cooled Reactors,” ''Proceedings of the International Fast Reactor Safety Meeting'', American Nuclear Society, Snowbird, UT, August 12-16, 1990.
* J. E. Cahalan and T. Wei , “Modeling Developments for the SAS4A and SASSYS Computer Codes,” Proceedings of the International Fast Reactor Safety Meeting, American Nuclear Society, Snowbird, UT, August 12-16, 1990.
* J. E. Cahalan and T. Wei , “Modeling Developments for the SAS4A and SASSYS Computer Codes,” ''Proceedings of the International Fast Reactor Safety Meeting'', American Nuclear Society, Snowbird, UT, August 12-16, 1990.
* P. L. Garner et al., "Development of a Graphical User Interface Allowing Use of the SASSYS 1 LMR Systems Analysis Code as an EBR II Interactive Simulator", Proceedings of International Topical Meeting on Advanced Reactors Safety, American Nuclear Society, Pittsburgh, PA, pp. 282 289, April 17 21, 1994.
* P. L. Garner et al., "Development of a Graphical User Interface Allowing Use of the SASSYS 1 LMR Systems Analysis Code as an EBR II Interactive Simulator", ''Proceedings of International Topical Meeting on Advanced Reactors Safety'', American Nuclear Society, Pittsburgh, PA, pp. 282 289, April 17 21, 1994.
* J. E. Cahalan et al., "Advanced LMR Safety Analysis Capabilities in the SASSYS 1 and SAS4A Computer Codes", Proceedings of the International Topical Meeting on Advanced Reactors Safety, American Nuclear Society, Pittsburgh, PA, April 17 21, 1994.
 
== Core Modeling ==
 
=== Oxide Fuel Models ===
 
* A. M. Tentner, H. U. Wider, and C. H. Bowers, “A Mechanistic Model for Fuel Flow Regimes and Fuel Plateout,” ''Trans. Am. Nucl. Soc.'', 30, 448, 1978.
* H. U. Wider et al., “The PLUT02 Overpower Excursion Code and a Comparison with EPIC,” ''Proceedings of the International Topical Meeting on Fast Reactor Safety Technology'', American Nuclear Society, Seattle, WA, August 19 23, 1979.
* A. M. Tentner and H. U. Wider, “LEVITATE A Mechanistic Model for the Analysis of Fuel and Cladding Dynamics under LOF Conditions for SAS4A,” ''Proceedings of the International Topical Meeting on Fast Reactor Safety Technology'', American Nuclear Society, Seattle, WA, August 19 23, 1979.
* C. H. Bowers et al., “Analysis of TREAT Tests L7 and L8 with SAS3D, LEVITATE and PLUT02,” ''Specialists Workshop on Predictive Analysis of Material Dynamics in LMFBR Safety Experiments'', LA 7938 C, Los Alamos Scientific Laboratory, March, 1979.
* A. M. Tentner and H. U. Wider, “Steel Ablation and Fuel Steel Mixing Modeling in LMFBR Accidents, ''Trans. Am. Nucl. Soc.'', 33, 540, 1979.
* A. M. Tentner and H. U. Wider, “The Influence of Steel Vapor Pressure on Fuel Motion in Voided LMFBR Channels,” ''Trans. Am. Nucl. Soc.'', 34, 512, 1980.
* A. M. Tentner and H. U. Wider, “Pressure Drop Modeling in Variable Area, Multiphase Flow,” ''Multiphase Transport: Fundamentals, Reactor Safety and Applications'', Editor N. Veziroglu, Hemisphere Publishing Co., May, 1980.
* A. M. Tentner and H. U. Wider, “New Aspects in the Analysis of Fuel Dynamics During Loss of Flow Transients,” ''Trans. Am. Nucl. Soc.'', 41, 374, 1982.
* A. M. Tentner and H. U. Wider, “Hydrodynamic and Thermal Modeling of Solid Particles in a Multi Phase, Multi Component Flow,” ''Proceedings of the 3rd Multiphase Flow and Heat Transfer Symposium - Workshop'', Miami Beach, Florida, April, 1983.
* A. M. Tentner and H. U. Wider, “Thermal Hydraulic Modeling for the Analysis of LMFBR Disrupted Core Behavior, ''Nuc. Eng. Des.'', 82, 373, 1984.
* D. J. Hill, “SAS4A Validation and Analysis of In Pile Experiments for Slow Ramp TOP’s,” ''Proceedings of the International Topical Topical Meeting on Fast Reactor Safety'', American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* J. A. Morman et al., “SAS Validation and Analysis of In Pile TUCOP Experiments,” ''Proceedings of the International Topical Meeting on Fast Reactor Safety'', American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* K. J. Miles and Kalimullah, “The Inherent Safety Phenomenon of Fission Gas Induced Axial Extrusion in Oxide and Metal Fueled LMFBRs,” ''Proceedings of the International Topical Meeting on Fast Reactor Safety'', American Nuclear Society, Knoxville, TN, April 21 25, 1985.
* A. M. Tentner and D. J. Hill, “PINACLE A Mechanistic Model for the Analysis of In Pin Fuel Relocation Under LOF and TOP Conditions for SAS4A, “ ''Trans. Am. Nucl. Soc.'', 49, 275, 1985.
* K. J. Miles and D. J. Hill, “DEFORM 4: Fuel Pin Characterization and Transient Response in the SAS4A Accident Analysis Code System,” ''Proceedings of the International Meeting on Science and Technology of Fast Reactor Safety'', British Nuclear Energy Society, Guernsey, UK, May 12 16, 1986.
* A. M. Tentner et al., “Fuel Relocation Modeling in the SAS4A Accident Analysis Code System,” ''Proceedings of the International Meeting on Science and Technology of Fast Reactor Safety'', British Nuclear Energy Society, Guernsey, UK, May 12 16, 1986.
 
=== Metallic Fuel Models ===
 
* Kalimullah, “SSCOMP: Model for Annular Zone Formation in U Pu Zr Fuel Pin,” ''Trans. Am. Nucl. Soc.'', 52, 499, 1986.
* K. J. Miles, “Metal Fuel Modeling for Inherently Safe Reactor Design,” ''Trans. Am. Nucl. Soc.'', 55, 417, 1987.
* K. J. Miles, “Metal Fuel Safety Performance,” ''Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors'', American Nuclear Society, Seattle, WA, May 1 5, 1988.
* A. M. Tentner et al., “Analyzing Unprotected Transients in Metal Fuel Cores with the SAS4A Accident Analysis Code,” ''Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors'', American Nuclear Society, Seattle, WA, May 1 5, 1988.
* A. M. Tentner and Kalimullah, “SAS4A Analysis of the M7 Metal Fuel TREAT Experiment,” ''Trans. Am. Nucl. Soc.'', 60, 419, 1989.
* A. M. Tentner, et al., “Analysis of Metal Fuel Transient Overpower Experiments with the SAS4A Accident Analysis Code,” ''Proceedings of the International Fast Reactor Safety Meeting'', American Nuclear Society, Snowbird, UT, August 12 16, 1990.
* A. M. Tentner, “Validation of the Metal Fuel Version of the SAS4A Accident Analysis Code,” ''Computer Simulation Multiconference'', New Orleans, LA, April (1991).
* T. Sofu and J. M. Kramer, “Implementation, Verification, and Validation of the FPIN2 Metal Fuel Pin Mechanics Model in the SASSYS/SAS4A LMR Transient Analysis Codes,” ''Proceedings of the International Topical Meeting on Advanced Reactors Safety'', American Nuclear Society, Pittsburgh, PA, April 17-21, 1994.
* T. Sofu et al., “SASSYS/SAS4A-FPIN2 Liquid Metal Reactor Transient Analysis Code System for Mechanical Analysis of Metallic Fuel Elements,” ''Nuclear Technology'', 113(3), 268, 1996.
 
=== Boiling Model ===
 
* G. Hoppner et al., “TREAT R5 Loss-of-Flow Experiment in Comparison with SAS Pretest Analysis,” ''Trans. Am.Nucl. Soc.'', 18, 213, 1974.
* L. L. Briggs, “Analysis of the OPERA 15 Two Dimensional Voiding Experiment Using the SAS4A Code,” CONF 841074 2 Rev, Eleventh Meeting of the Liquid Metal Boiling Working Group, Grenoble, France, October, 1984.
* F. E. Dunn, “Validation of the SAS4A Sodium Boiling Model at Low Power,” ,” ''Trans. Am. Nucl. Soc.'', 88, 287, 2003.
 
=== Multiple-Pin Model ===


== Multiple-Pin Model ==
* F. E. Dunn, "Integrated Intra Subassembly Treatment in the SASSYS 1 LMR Systems Analysis Code," ''Proceedings of the Fifth International Topical Meeting on Reactor Thermal Hydraulics'', NURETH 5, Salt Lake City, September, 1992.
* F. E. Dunn, "Verification and Implications of the Multiple Pin Treatment in the SASSYS 1 LMR Systems Analysis Code", ''Proceedings of International Topical Meeting on Advanced Reactors Safety'', American Nuclear Society, Pittsburgh, PA, April 17 21, 1994.
* F. E. Dunn, "Validation of Detailed Thermal Hydraulic Models Used for LMR Safety and for Improvement of Technical Specifications", ''Proceedings of the American Nuclear Society International Topical Meeting on Safety of Operating Reactors'', American Nuclear Society, Seattle (Bellevue), WA, September 17 20, 1995.
* F. E. Dunn, "Verification and Implications of the Multiple Pin Treatment in the SASSYS 1 Liquid Metal Reactor Systems Analysis Code", ''Nucl. Tech.'', 114, 147, 1996.


* F. E. Dunn, "Integrated Intra Subassembly Treatment in the SASSYS 1 LMR Systems Analysis Code," Proceedings of the Fifth International Topical Meeting on Reactor Thermal Hydraulics, NURETH 5, Salt Lake City, September, 1992.
=== Sub-Channel Thermal-Hydraulics Model ===
* F. E. Dunn, "Verification and Implications of the Multiple Pin Treatment in the SASSYS 1 LMR Systems Analysis Code", Proceedings of International Topical Meeting on Advanced Reactors Safety, American Nuclear Society, Pittsburgh, PA, April 17 21, 1994.
* F. E. Dunn, "Validation of Detailed Thermal Hydraulic Models Used for LMR Safety and for Improvement of Technical Specifications", Proceedings of the American Nuclear Society International Topical Meeting on Safety of Operating Reactors, American Nuclear Society, Seattle (Bellevue), WA, September 17 20, 1995.
* F. E. Dunn, "Verification and Implications of the Multiple Pin Treatment in the SASSYS 1 Liquid Metal Reactor Systems Analysis Code", Nucl. Tech., 114, 147, 1996.


== Radial Core Expansion Model ==
* F. E. Dunn, D. Hahn, H. Jeong, K Ha, and J. E. Cahalan, "Whole Core Sub-Channel Analysis for LMR Passive Safety Analysis," ''14th Pacific Basin Nuclear Conference'', Honolulu, Hawaii, March 21-25, 2004.
* F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, "Detailed Sub-Channel Treatment for Whole Core LMR Analysis," ''NUTHOS-6 International Topical Meeting on Nuclear Reactor Thermal Hydraulics'', Operation and Safety, Nara, Japan, October 4-8, 2004.
* F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, "Whole Core Sub-Channel Analysis in LMR Systems Codes, Current Status," ''Trans. Am. Nucl. Soc.'', 92, 427, 2005.
* F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, "Whole Core Sub-Channel Analysis Verification with the EBR-II SHRT-17 Test," ''Proc. ICAPP ’06'', Paper 6364, Reno, NV, June 4-8, 2006.


* R. A. Wigeland, “Effect of a Detailed Radial Core Expansion Reactivity Feedback Model on ATWS Calculations Using SASSYS/SAS4A,” Trans. Am. Nucl Soc., 53, 303, 1986.
=== Radial Core Expansion Model ===
* R. A. Wigeland, “Comparison of the SASSYS/SAS4A Radial Core Expansion Reactivity Feedback Model and the Empirical Correlation for the FFTF,” Trans. Am. Nucl. Soc., 55, 423, 1987.
* R. A. Wigeland and T. J. Moran, “Radial Core Expansion Reactivity Feedback in Advanced LMRs: Uncertainties and Their Effects on Inherent Safety,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
* D. J. Hill and R. A. Wigeland, “Validation of the SASSYS Core Radial Expansion Reactivity Feedback Model,” Trans. Am. Nucl. Soc., 56, 380, 1988.


== Pump Model ==
* R. A. Wigeland, “Effect of a Detailed Radial Core Expansion Reactivity Feedback Model on ATWS Calculations Using SASSYS/SAS4A,” ''Trans. Am. Nucl Soc.'', 53, 303, 1986.
* R. A. Wigeland, “Comparison of the SASSYS/SAS4A Radial Core Expansion Reactivity Feedback Model and the Empirical Correlation for the FFTF,” ''Trans. Am. Nucl. Soc.'', 55, 423, 1987.
* R. A. Wigeland and T. J. Moran, “Radial Core Expansion Reactivity Feedback in Advanced LMRs: Uncertainties and Their Effects on Inherent Safety,” ''Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors'', American Nuclear Society, Seattle, WA, May 1 5, 1988.
* D. J. Hill and R. A. Wigeland, “Validation of the SASSYS Core Radial Expansion Reactivity Feedback Model,” ''Trans. Am. Nucl. Soc.'', 56, 380, 1988.


* F. E. Dunn and D. J. Malloy, “LMR Centrifugal Pump Coastdowns,” Proceedings of the International Topical Meeting on Anticipated and Abnormal Transients in Nuclear Power Plants, American Nuclear Society, Atlanta, GA, April 12 15, 1987.
=== Spatial Kinetics ===


== RVACS/RACS Model ==
* J. E. Cahalan. et al., “Development of a Coupled Dynamics Code with Transport Theory Capability and Application to Accelerator-Driven Systems Transients,” ''Proceedings of the ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium'', American Nuclear Society, Pittsburgh, PA, May 7- 12, 2000.


* F. E. Dunn, “Validation of the RVACS/RACS Model in SASSYS 1,” Trans. Am. Nucl. Soc., 55, 723, 1987.
== Systems Models ==
* F. E. Dunn, “SASSYS 1 Modeling of RVACS/RACS Heat Removal in an LMR,” Trans. Am. Nucl. Soc., 55, 724, 1987.
* F. E. Dunn, “RACS Shutdown Heat Removal in a Modular Sized LMR,” ASME Winter Meeting, Chicago, IL, November 28 December 2, 1988.


== Control System Model ==
=== Pump Model ===


* R. B. Vilim et al., “A Control System Model for the SASSYS 1 Systems Analysis Code,” Trans. Am. Nucl. Soc., 52, 505, 1986.
* F. E. Dunn and D. J. Malloy, “LMR Centrifugal Pump Coastdowns,” ''Proceedings of the International Topical Meeting on Anticipated and Abnormal Transients in Nuclear Power Plants'', American Nuclear Society, Atlanta, GA, April 12 15, 1987.
* R.B. Vilim, “Solution of Generalized Control System Equations at Steady State,” Trans. Am. Nucl. Soc., 54, 171, 1987.
* R. B. Vilim et al., “Generalized Control System Modeling for Liquid Metal Reactors,” Nucl. Sci. Eng., 99, 183, July, 1988.


== Balance-of-Plant Model ==
=== RVACS/RACS Model ===


* L. L. Briggs, “A New Balance of Plant Model for the SASSYS 1 LMR System Analysis Code,” Trans. Am. Nucl. Soc., 60, 709, 1989.
* F. E. Dunn, “Validation of the RVACS/RACS Model in SASSYS 1,” ''Trans. Am. Nucl. Soc.'', 55, 723, 1987.
* P. A. Pizzica, “An Improved Steam Generator Model for the SASSYS Code,” Trans. Am. Nucl. Soc., 60, 712, 1989.
* F. E. Dunn, “SASSYS 1 Modeling of RVACS/RACS Heat Removal in an LMR,” ''Trans. Am. Nucl. Soc.'', 55, 724, 1987.
* J. Y. Ku, “SASSYS 1 Balance of Plant Component Models for an Integrated Plant Response,” Trans. Am. Nucl. Soc., 60, 716, 1989.
* F. E. Dunn, “RACS Shutdown Heat Removal in a Modular Sized LMR,” ''ASME Winter Meeting'', Chicago, IL, November 28 December 2, 1988.


== Spatial Kinetics ==
=== Control System Model ===


* J. E. Cahalan. et al., “Development of a Coupled Dynamics Code with Transport Theory Capability and Application to Accelerator-Driven Systems Transients,” Proceedings of the ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, American Nuclear Society, Pittsburgh, PA, May 7- 12, 2000.
* R. B. Vilim et al., “A Control System Model for the SASSYS 1 Systems Analysis Code,” ''Trans. Am. Nucl. Soc.'', 52, 505, 1986.
* R. B. Vilim, “Solution of Generalized Control System Equations at Steady State,” ''Trans. Am. Nucl. Soc.'', 54, 171, 1987.
* R. B. Vilim et al., “Generalized Control System Modeling for Liquid Metal Reactors,” ''Nucl. Sci. Eng.'', 99, 183, July, 1988.


== Sub-Channel Thermal-Hydraulics Model ==
=== Balance-of-Plant Model ===


* F. E. Dunn, D. Hahn, H. Jeong, K Ha, and J. E. Cahalan, Whole Core Sub-Channel Analysis for LMR Passive Safety Analysis, 14th Pacific Basin Nuclear Conference, Honolulu, Hawaii, March 21-25, 2004.
* L. L. Briggs, “A New Balance of Plant Model for the SASSYS 1 LMR System Analysis Code,” ''Trans. Am. Nucl. Soc.'', 60, 709, 1989.
* F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, Detailed Sub-Channel Treatment for Whole Core LMR Analysis, NUTHOS-6 International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Nara, Japan, October 4-8, 2004.
* P. A. Pizzica, “An Improved Steam Generator Model for the SASSYS Code,” ''Trans. Am. Nucl. Soc.'', 60, 712, 1989.
* F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, Whole Core Sub-Channel Analysis in LMR Systems Codes, Current Status, Trans. Am. Nucl. Soc., 92, 427, 2005.
* J. Y. Ku, “SASSYS 1 Balance of Plant Component Models for an Integrated Plant Response,” ''Trans. Am. Nucl. Soc.'', 60, 716, 1989.
* F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, Whole Core Sub-Channel Analysis Verification with the EBR-II SHRT-17 Test, Proc. ICAPP ’06, Paper 6364, Reno, NV, June 4-8, 2006.

Latest revision as of 17:38, February 13, 2012

The references listed on this page represent open-literature citations related to SAS4A/SASSYS-1 code modeling capabilities and developments. It does not include references to modeling and simulation analyses in which SAS4A/SASSYS-1 was used to evaluate safety performance.

General References

  • D. R. Ferguson et al., “The SAS4A LMFBR Accident Analysis Code System: A Progress Report,” Proceedings of the International Meeting on Fast Reactor Safety and Related Physics, American Nuclear Society, Chicago, IL, October 5 8, 1976.
  • J. E. Cahalan et al., “The Status and Experimental Basis of the SAS4A Accident Analysis Code System,” Proceedings of the International Meeting on Fast Reactor Safety Technology, American Nuclear Society, Seattle, WA, August 19 23, 1979.
  • H. U. Wider et al., “Status and Validation of the SAS4A Accident Analysis Code System,” Proceedings of the LMFBR Safety Topical Meeting, European Nuclear Society, Lyon, France, July 19 23, 1982.
  • A. M. Tentner et al., “The SAS4A LMFBR Whole Core Accident Analysis Code,” Proceedings of the International Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
  • A. M. Tentner et al., “SAS4A: A Computer Model for the Analysis of Hypothetical Core Disruptive Accidents in Liquid Metal Reactors,” 1987 SCS Eastern Simulation Conference, Society for Computer Simulation, Orlando, FL, April 6 9, 1987.
  • A. M. Tentner et al., “Simulating Unprotected Accidents for Advanced Liquid Metal Reactors Using the SAS4A Accident Analysis Code,” 1988 SCS Simulators Conference, Society for Computer Simulation, Orlando, FL, April 18 21, 1988.
  • J. E. Cahalan and T. Wei, “Modeling Developments for the SAS4A and SASSYS Computer Codes,” Proceedings of the International Fast Reactor Safety Meeting, American Nuclear Society, Snowbird, UT, August 12-16, 1990.
  • J. E. Cahalan et al., “Advanced LMR Safety Analysis Capabilities in the SASSYS-1 and SAS4A Computer Codes,” Proceedings of the International Topical Meeting on Advanced Reactors Safety, American Nuclear Society, Pittsburgh, PA, April 17-21, 1994.
  • F. E. Dunn and F. G. Prohammer, “SASSYS Analysis of Degraded Shut Down Heat Removal Performance in LMFBRs,” ASME Paper No. 82 WA/HT 37, 1982.
  • F. E. Dunn and F. G. Prohammer, “The SASSYS LMFBR Systems Analysis Code,” Proceedings of the 10th IMACS World Conference on Systems Simulation and Scientific Computation, Vol. 4, Montreal, Canada, pp. 127-129, August, 1982.
  • F. E. Dunn et al., “The SASSYS 1 LMFBR Systems Analysis Code,” Proceedings of the International Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
  • D. K. Warinner and F. E. Dunn, “SASSYS 1 Computer Code Verification with EBR II Test Data,” Proceedings of the International Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
  • F. E. Dunn, “The SAS4A/SASSYS l Sodium Boiling Model for LMFBR Whole Core Analysis,” Heat Transfer - Denver 1985, AIChE Symposium Series, No. 245, Vol. 81, 1985.
  • F. E. Dunn et al., “LMR Thermal Hydraulics Calculations in the U.S.,” Proceedings of the International Topical Meeting on Advances in Reactor Physics, Mathematics and Computation, Paris, France, April 27 30, 1987.
  • D. J. Hill, “SASSYS Analysis of EBR-II SHRT Experiments,” Trans. Am. Nucl. Soc., 55, 421, 1987.
  • F. E. Dunn, “LMR Thermal Hydraulics Calculations in the U.S.,” Nucl. Sci. Eng., 100, 558, 1988.
  • F. E. Dunn and T. Y. C. Wei, “Simulating Operational Transients with the SASSYS 1 LMR Systems Analysis Code,” 1988 SCS Simulators Conference, Society for Computer Simulation, Orlando, FL, April 18-21, 1988.
  • F. E. Dunn and T. Y. C. Wei, “The Role of SASSYS 1 in LMR Safety Analysis,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
  • D. J. Hill, “SASSYS Validation Studies,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
  • F. E. Dunn, “Decay Heat Calculations for Transient Analysis,” Trans. Am. Nucl. Soc., 60, 633, 1989.
  • J. P. Herzog, “SASSYS Validation with the EBR II Shutdown Heat Removal Tests,” Trans. Am. Nucl. Soc., 60, 730, 1989.
  • F. E. Dunn and J. P. Herzog, “Thermal-Hydraulic Impact of Failure of Highly Irradiated Fuel Pins on LMR Passive Safety,” Trans. Am. Nucl. Soc., 62, 673, 1990.
  • F. E. Dunn, “Consequences of Pipe Ruptures in Metal Fueled, Liquid Metal Cooled Reactors,” Proceedings of the International Fast Reactor Safety Meeting, American Nuclear Society, Snowbird, UT, August 12-16, 1990.
  • J. E. Cahalan and T. Wei , “Modeling Developments for the SAS4A and SASSYS Computer Codes,” Proceedings of the International Fast Reactor Safety Meeting, American Nuclear Society, Snowbird, UT, August 12-16, 1990.
  • P. L. Garner et al., "Development of a Graphical User Interface Allowing Use of the SASSYS 1 LMR Systems Analysis Code as an EBR II Interactive Simulator", Proceedings of International Topical Meeting on Advanced Reactors Safety, American Nuclear Society, Pittsburgh, PA, pp. 282 289, April 17 21, 1994.

Core Modeling

Oxide Fuel Models

  • A. M. Tentner, H. U. Wider, and C. H. Bowers, “A Mechanistic Model for Fuel Flow Regimes and Fuel Plateout,” Trans. Am. Nucl. Soc., 30, 448, 1978.
  • H. U. Wider et al., “The PLUT02 Overpower Excursion Code and a Comparison with EPIC,” Proceedings of the International Topical Meeting on Fast Reactor Safety Technology, American Nuclear Society, Seattle, WA, August 19 23, 1979.
  • A. M. Tentner and H. U. Wider, “LEVITATE A Mechanistic Model for the Analysis of Fuel and Cladding Dynamics under LOF Conditions for SAS4A,” Proceedings of the International Topical Meeting on Fast Reactor Safety Technology, American Nuclear Society, Seattle, WA, August 19 23, 1979.
  • C. H. Bowers et al., “Analysis of TREAT Tests L7 and L8 with SAS3D, LEVITATE and PLUT02,” Specialists Workshop on Predictive Analysis of Material Dynamics in LMFBR Safety Experiments, LA 7938 C, Los Alamos Scientific Laboratory, March, 1979.
  • A. M. Tentner and H. U. Wider, “Steel Ablation and Fuel Steel Mixing Modeling in LMFBR Accidents, Trans. Am. Nucl. Soc., 33, 540, 1979.
  • A. M. Tentner and H. U. Wider, “The Influence of Steel Vapor Pressure on Fuel Motion in Voided LMFBR Channels,” Trans. Am. Nucl. Soc., 34, 512, 1980.
  • A. M. Tentner and H. U. Wider, “Pressure Drop Modeling in Variable Area, Multiphase Flow,” Multiphase Transport: Fundamentals, Reactor Safety and Applications, Editor N. Veziroglu, Hemisphere Publishing Co., May, 1980.
  • A. M. Tentner and H. U. Wider, “New Aspects in the Analysis of Fuel Dynamics During Loss of Flow Transients,” Trans. Am. Nucl. Soc., 41, 374, 1982.
  • A. M. Tentner and H. U. Wider, “Hydrodynamic and Thermal Modeling of Solid Particles in a Multi Phase, Multi Component Flow,” Proceedings of the 3rd Multiphase Flow and Heat Transfer Symposium - Workshop, Miami Beach, Florida, April, 1983.
  • A. M. Tentner and H. U. Wider, “Thermal Hydraulic Modeling for the Analysis of LMFBR Disrupted Core Behavior, Nuc. Eng. Des., 82, 373, 1984.
  • D. J. Hill, “SAS4A Validation and Analysis of In Pile Experiments for Slow Ramp TOP’s,” Proceedings of the International Topical Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
  • J. A. Morman et al., “SAS Validation and Analysis of In Pile TUCOP Experiments,” Proceedings of the International Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
  • K. J. Miles and Kalimullah, “The Inherent Safety Phenomenon of Fission Gas Induced Axial Extrusion in Oxide and Metal Fueled LMFBRs,” Proceedings of the International Topical Meeting on Fast Reactor Safety, American Nuclear Society, Knoxville, TN, April 21 25, 1985.
  • A. M. Tentner and D. J. Hill, “PINACLE A Mechanistic Model for the Analysis of In Pin Fuel Relocation Under LOF and TOP Conditions for SAS4A, “ Trans. Am. Nucl. Soc., 49, 275, 1985.
  • K. J. Miles and D. J. Hill, “DEFORM 4: Fuel Pin Characterization and Transient Response in the SAS4A Accident Analysis Code System,” Proceedings of the International Meeting on Science and Technology of Fast Reactor Safety, British Nuclear Energy Society, Guernsey, UK, May 12 16, 1986.
  • A. M. Tentner et al., “Fuel Relocation Modeling in the SAS4A Accident Analysis Code System,” Proceedings of the International Meeting on Science and Technology of Fast Reactor Safety, British Nuclear Energy Society, Guernsey, UK, May 12 16, 1986.

Metallic Fuel Models

  • Kalimullah, “SSCOMP: Model for Annular Zone Formation in U Pu Zr Fuel Pin,” Trans. Am. Nucl. Soc., 52, 499, 1986.
  • K. J. Miles, “Metal Fuel Modeling for Inherently Safe Reactor Design,” Trans. Am. Nucl. Soc., 55, 417, 1987.
  • K. J. Miles, “Metal Fuel Safety Performance,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
  • A. M. Tentner et al., “Analyzing Unprotected Transients in Metal Fuel Cores with the SAS4A Accident Analysis Code,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
  • A. M. Tentner and Kalimullah, “SAS4A Analysis of the M7 Metal Fuel TREAT Experiment,” Trans. Am. Nucl. Soc., 60, 419, 1989.
  • A. M. Tentner, et al., “Analysis of Metal Fuel Transient Overpower Experiments with the SAS4A Accident Analysis Code,” Proceedings of the International Fast Reactor Safety Meeting, American Nuclear Society, Snowbird, UT, August 12 16, 1990.
  • A. M. Tentner, “Validation of the Metal Fuel Version of the SAS4A Accident Analysis Code,” Computer Simulation Multiconference, New Orleans, LA, April (1991).
  • T. Sofu and J. M. Kramer, “Implementation, Verification, and Validation of the FPIN2 Metal Fuel Pin Mechanics Model in the SASSYS/SAS4A LMR Transient Analysis Codes,” Proceedings of the International Topical Meeting on Advanced Reactors Safety, American Nuclear Society, Pittsburgh, PA, April 17-21, 1994.
  • T. Sofu et al., “SASSYS/SAS4A-FPIN2 Liquid Metal Reactor Transient Analysis Code System for Mechanical Analysis of Metallic Fuel Elements,” Nuclear Technology, 113(3), 268, 1996.

Boiling Model

  • G. Hoppner et al., “TREAT R5 Loss-of-Flow Experiment in Comparison with SAS Pretest Analysis,” Trans. Am.Nucl. Soc., 18, 213, 1974.
  • L. L. Briggs, “Analysis of the OPERA 15 Two Dimensional Voiding Experiment Using the SAS4A Code,” CONF 841074 2 Rev, Eleventh Meeting of the Liquid Metal Boiling Working Group, Grenoble, France, October, 1984.
  • F. E. Dunn, “Validation of the SAS4A Sodium Boiling Model at Low Power,” ,” Trans. Am. Nucl. Soc., 88, 287, 2003.

Multiple-Pin Model

  • F. E. Dunn, "Integrated Intra Subassembly Treatment in the SASSYS 1 LMR Systems Analysis Code," Proceedings of the Fifth International Topical Meeting on Reactor Thermal Hydraulics, NURETH 5, Salt Lake City, September, 1992.
  • F. E. Dunn, "Verification and Implications of the Multiple Pin Treatment in the SASSYS 1 LMR Systems Analysis Code", Proceedings of International Topical Meeting on Advanced Reactors Safety, American Nuclear Society, Pittsburgh, PA, April 17 21, 1994.
  • F. E. Dunn, "Validation of Detailed Thermal Hydraulic Models Used for LMR Safety and for Improvement of Technical Specifications", Proceedings of the American Nuclear Society International Topical Meeting on Safety of Operating Reactors, American Nuclear Society, Seattle (Bellevue), WA, September 17 20, 1995.
  • F. E. Dunn, "Verification and Implications of the Multiple Pin Treatment in the SASSYS 1 Liquid Metal Reactor Systems Analysis Code", Nucl. Tech., 114, 147, 1996.

Sub-Channel Thermal-Hydraulics Model

  • F. E. Dunn, D. Hahn, H. Jeong, K Ha, and J. E. Cahalan, "Whole Core Sub-Channel Analysis for LMR Passive Safety Analysis," 14th Pacific Basin Nuclear Conference, Honolulu, Hawaii, March 21-25, 2004.
  • F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, "Detailed Sub-Channel Treatment for Whole Core LMR Analysis," NUTHOS-6 International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Nara, Japan, October 4-8, 2004.
  • F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, "Whole Core Sub-Channel Analysis in LMR Systems Codes, Current Status," Trans. Am. Nucl. Soc., 92, 427, 2005.
  • F. E. Dunn, J. E. Cahalan, D. Hahn, and H. Jeong, "Whole Core Sub-Channel Analysis Verification with the EBR-II SHRT-17 Test," Proc. ICAPP ’06, Paper 6364, Reno, NV, June 4-8, 2006.

Radial Core Expansion Model

  • R. A. Wigeland, “Effect of a Detailed Radial Core Expansion Reactivity Feedback Model on ATWS Calculations Using SASSYS/SAS4A,” Trans. Am. Nucl Soc., 53, 303, 1986.
  • R. A. Wigeland, “Comparison of the SASSYS/SAS4A Radial Core Expansion Reactivity Feedback Model and the Empirical Correlation for the FFTF,” Trans. Am. Nucl. Soc., 55, 423, 1987.
  • R. A. Wigeland and T. J. Moran, “Radial Core Expansion Reactivity Feedback in Advanced LMRs: Uncertainties and Their Effects on Inherent Safety,” Proceedings of the International Topical Meeting on Safety of Next Generation Power Reactors, American Nuclear Society, Seattle, WA, May 1 5, 1988.
  • D. J. Hill and R. A. Wigeland, “Validation of the SASSYS Core Radial Expansion Reactivity Feedback Model,” Trans. Am. Nucl. Soc., 56, 380, 1988.

Spatial Kinetics

  • J. E. Cahalan. et al., “Development of a Coupled Dynamics Code with Transport Theory Capability and Application to Accelerator-Driven Systems Transients,” Proceedings of the ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, American Nuclear Society, Pittsburgh, PA, May 7- 12, 2000.

Systems Models

Pump Model

  • F. E. Dunn and D. J. Malloy, “LMR Centrifugal Pump Coastdowns,” Proceedings of the International Topical Meeting on Anticipated and Abnormal Transients in Nuclear Power Plants, American Nuclear Society, Atlanta, GA, April 12 15, 1987.

RVACS/RACS Model

  • F. E. Dunn, “Validation of the RVACS/RACS Model in SASSYS 1,” Trans. Am. Nucl. Soc., 55, 723, 1987.
  • F. E. Dunn, “SASSYS 1 Modeling of RVACS/RACS Heat Removal in an LMR,” Trans. Am. Nucl. Soc., 55, 724, 1987.
  • F. E. Dunn, “RACS Shutdown Heat Removal in a Modular Sized LMR,” ASME Winter Meeting, Chicago, IL, November 28 December 2, 1988.

Control System Model

  • R. B. Vilim et al., “A Control System Model for the SASSYS 1 Systems Analysis Code,” Trans. Am. Nucl. Soc., 52, 505, 1986.
  • R. B. Vilim, “Solution of Generalized Control System Equations at Steady State,” Trans. Am. Nucl. Soc., 54, 171, 1987.
  • R. B. Vilim et al., “Generalized Control System Modeling for Liquid Metal Reactors,” Nucl. Sci. Eng., 99, 183, July, 1988.

Balance-of-Plant Model

  • L. L. Briggs, “A New Balance of Plant Model for the SASSYS 1 LMR System Analysis Code,” Trans. Am. Nucl. Soc., 60, 709, 1989.
  • P. A. Pizzica, “An Improved Steam Generator Model for the SASSYS Code,” Trans. Am. Nucl. Soc., 60, 712, 1989.
  • J. Y. Ku, “SASSYS 1 Balance of Plant Component Models for an Integrated Plant Response,” Trans. Am. Nucl. Soc., 60, 716, 1989.
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