References

General References

• 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.
• 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.

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.

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.

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.

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.

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.