The manual for SAS4A/SASSYS-1 version 3.1.2 describes code capabilities, user input, and major code modules. It is divided into 16 chapters as described below. The general citation for the SAS4A/SASSYS-1 manual is

:T. H. Fanning, ed., ''The SAS4A/SASSYS-1 Safety Analysis Code System'', ANL/NE-12/4, Nuclear Engineering Division, Argonne National Laboratory, January 31, 2012.

== Introduction ==

SAS4A/SASSYS-1 provide a detailed, multiple-channel thermal/hydraulic treatment of the reactor core. Each channel represents a fuel pin, its cladding, the associated coolant, and a fraction of the subassembly duct wall. Other positioning hardware, such as wire wraps or grid spacers, is usually lumped into the structure field with the duct wall. Within a channel, the flow is assumed to be one-dimensional in the axial direction, and the temperature field in the fuel, cladding, coolant, and structure is assumed to be two-dimensional in the radial and axial directions. Usually, a channel represents an average fuel element in a subassembly or a group of subassemblies. A channel may also represent pins in blanket or control subassemblies. Alternately, a single channel may also be used to represent the hottest pin in an assembly, or any other subset of a subassembly. The axial extent of a channel covers the entire length of a subassembly, including the core, the axial blankets, the fission gas plenum and the spaces above and below the pin/cladding geometry. Different channels may be used to account for radial and azimuthal design geometry, power, coolant flow, and burnup variations within the reactor core. From ten to thirty channels normally provide sufficient discretization, depending on the core design. Significantly more channels can be used if necessary.

([[Media:SAS_Ch01_-_Introduction_(ANL-NE-12-4).pdf|PDF]])

== User's Guide ==

Chapter 2 contains a user's guide for SAS4A/SASSYS-1 applications, including a complete description of the standard input file.

([[Media:SAS_Ch02_-_User_Guide_(ANL-NE-12-4).pdf|PDF]])

== Steady-State and Transient Thermal Hydraulics in Core Assemblies ==

Chapter 3 contains the description of the formulation for the SAS4A/SASSYS-1 pin heat transfer and single-phase coolant thermal/hydraulics model. A major new addition to this model is the capability to treat channel‐to-channel heat transfer due to conduction and convection at all axial locations between the channel inlet and outlet, permitting a consistent multiple-pin subassembly treatment. Also, the subassembly-to-subassembly heat transfer model has been improved, and axial conduction in the coolant has been added. These modeling additions have been proven in validation analyses of EBR-II Shutdown Heat Removal Tests, and are required for accurate predictions of intra-subassembly flow and temperature variations in EBR-II during transients from normal to shutdown operating conditions.

([[Media:SAS_Ch03_-_Core_Channel_Thermal_Hydraulics_(ANL-NE-12-4).pdf|PDF]])

== Reactor Point Kinetics, Decay Heat, and Reactivity Feedback ==

Chapter 4 contains the description of the formulation for the SAS4A/SASSYS‐1 reactor points kinetics and reactivity feedback models. The new addition to this module is an option for an EBR-II-specific reactivity feedback model that is being validated with analysis of reactor operating data and used for predictive calculations of margins in design basis analyses. This module provides the reactor power level to the core thermal/hydraulics models for determination of the heating rate in the fuel, and receives core materials temperature and geometry information to calculate the reactivity feedbacks employed in the solution of the point kinetics equations.

([[Media:SAS_Ch04_-_Reactivity_and_Kinetics_(ANL-NE-12-4).pdf|PDF]])

== Primary and Intermediate Loop Thermal Hydraulics Module ==

Chapter 5 presents a full description of the formulation for the PRIMAR-4 sodium loops thermal/hydraulic model. This model provides boundary coolant pressure and flow conditions for the core channel models, including transient heat losses through normal and emergency heat removal systems and the transient performance of pumps. The major new addition to PRIMAR-4 is the option for multiple core inlet and outlet coolant plena, permitting exact representation of the actual EBR-II coolant systems geometry.

([[Media:SAS_Ch05_-_PRIMAR-4_(ANL-NE-12-4).pdf |PDF]])

== Control System ==

The plant control and protection system model described in Chapter 6 is unchanged from prior versions of SASSYS-1, except for the addition of an option to allow dynamic allocation of model data storage at execution time.

([[Media:SAS_Ch06_-_Control_System_(ANL-NE-12-4).pdf|PDF]])

== Balance of Plant Thermal/Hydraulics Models ==

The balance-of-plant (BOP) model described in Chapter 7 is new for SAS4A/SASSYS-1. It was implemented to permit 1) improved simulation of EBR-II design basis transients, 2) whole-plant analysis of IFR designs for optimization of advanced reactor control system strategies, and 3) core temperature margin assessments in unprotected accident sequences (i.e. beyond design basis accidents (BDBA) and anticipated transients without scram (ATWS)). In these latter sequences, core response depends strongly upon the performance of the balance-of‐plant, because the core neutronic and thermal/hydraulic behavior is determined by the availability of heat sinks outside the core. The BOP model couples to PRIMAR-4 at the steam generator.

([[Media:SAS_Ch07_-_Balance_of_Plant_(ANL-NE-12-4).pdf|PDF]])

== DEFORM-4: Steady-State and Transient Pre-Failure Pin Behavior ==

Chapter 8 provides a description of the DEFORM-4 fuel element behavior model for stainless steel‐clad oxide fuel, which is unchanged from prior versions of SAS4A/SASSYS-1.

([[Media:SAS_Ch08_-_DEFORM4_Pre-Failure_Pin_Modeling.pdf |PDF]])

== DEFORM-5: Metallic Fuel Cladding Transient Behavior Model ==

Chapter 9 contains the description of the DEFORM‐5 model, which treats the transient behavior of stainless steel and advanced (HT-9) cladding for metal fuel elements. This is a new model for SAS4A/SASSYS-1, and is aimed at predicting margin to cladding failure, and timing and location of failure in limiting transients. It includes physical phenomena unique to metallic fuel, such as fuel/cladding chemical interactions.

([[Media:SAS_Ch09_-_DEFORM5_Metallic_Fuel_Cladding_Model.pdf|PDF]])

== SSCOMP: Pre-Transient Characterization of Metallic Fuel Pins ==

The SSCOMP model described in Chapter 10 has been revised to reflect newly available metal fuel material properties evaluations recorded in the IFR Material Properties Handbook. An efficient correlation technique has been implemented in all SAS4A/SASSYS-1 material properties routines that accurately generates the data from the IFR Handbook for use in all the modules of the code. It is planned to revise the material migration capability in SSCOMP for ternary fuel, to add models for fission gas generation and release, swelling, and all other phenomena needed to describe the transition from cold, clean, unirradiated conditions to hot irradiated conditions.

([[Media:SAS_Ch10_-_SSCOMP_Pre-Transient_Metallic_Fuel_(ANL-NE-12-4).pdf|PDF]])

== FPIN2: Pre-Failure Metal Fuel Pin Behavior Model ==

Chapter 11 contains the description of a major new addition to the SAS4A/SASSYS-1 cods, the FPIN2 metal fuel pin mechanics model. FPIN2 is a validated model for metal fuel pin transient behavior. Unlike DEFORM-5, which treats only the cladding response, FPIN2 provides a finite-element solution of the fuel and cladding mechanics equations for the elastic/plastic response, including fission gas pressurization and migration, molten cavity formation and growth, and fuel/cladding chemical interaction and cladding thinning. The interface between SAS4A/SASSYS-1 and FPIN2 has been designed to permit stand-alone execution of FPIN2 for direct verification, and to replace the FPIN2 thermal/hydraulics calculation with the SAS4A/SASSYS-1 counterparts for coupled calculations. The application for this model is design basis analysis of driver and experimental fuel elements in EBR-II for the purpose of margin-to-failure assessments.

([[Media:SAS_Ch11_-_FPIN2_Pre-Failure_Metal_Fuel_Model_(ANL-NE-12-4).pdf|PDF]])

== Sodium Voiding Model ==

The TSBOIL module for liquid metal coolant boiling and two-phase thermal/hydraulics calculations has been retained intact from previous versions of SAS4A/SASSYS-1, with the addition of a set of modifications to describe the sudden release of noncondensible fission gas from a cladding rupture in the upper fission gas plenum of metal fuel elements and the subsequent plenum blow-down and liquid coolant expulsion. This option has been used to assess the safety implications of long-term fuel element irradiations in EBR-II.

([[Media:SAS_Ch12_-_Sodium_Voiding_(ANL-NE-12-4).pdf|PDF]])

== Cladding Motion Model — CLAP ==

The CLAP model described in Chapter 13 is relevant only to oxide fuel, and have remained unchanged since the previous documentation.

([[Media:SAS_Ch13_-_CLAP_Cladding_Model_(ANL-NE-12-4).pdf|PDF]])

== Non-Voided Channel Fuel Motion Analysis — PLUTO2 ==

The PLUTO2 model described in Chapter 14 is relevant only to oxide fuel, and have remained unchanged since the previous documentation.

([[Media:SAS_Ch14_-_PLUTO2_Non-Voided_Fuel_Motion_(ANL-NE-12-4).pdf|PDF]])

== PINACLE: In-Pin Pre-Failure Molten Fuel Relocation Module ==

The PINACLE model described in Chapter 15 and the LEVITATE model described in Chapter 16 have been upgraded for applications to metallic fuel. The model enhancements added to PINACLE and LEVITATE for metal fuel include fuel/cladding and fuel/structure chemical interactions and fission gas generation and migration with fuel swelling. Preliminary analyses of TREAT M-Series in-pile metal fuel tests have been completed, and applications to severe accident sequences in metal-fueled IFR cores have been completed and documented.

([[Media:SAS_Ch15_-_PINACLE_In-Pin_Fuel_Motion_(ANL-NE-12-4).pdf|PDF]])

== LEVITATE: Voided Channel Fuel Motion Analysis ==

See the [[#PINACLE: In-Pin Pre-Failure Molten Fuel Relocation Module|PINACLE]] chapter description.

([[Media:SAS_Ch16_-_LEVITATE_Voided_Fuel_Motion_(ANL-NE-12-4).pdf|PDF]])

PDF Code Manual

2012-05-29T13:12:54Z

PDF Code Manual

2012-05-03T13:52:16Z

2012-05-03T13:36:55Z

Fanning: Added link for Intro

== The SAS4A/SASSYS-1 Safety Analysis System (ANL/NE-12/4) ==

The manual for SAS4A/SASSYS-1 describes code capabilities, user input, and major code modules. It is divided into 16 chapters.

=== Introduction ===

SAS4A/SASSYS-1 provide a detailed, multiple-channel thermal/hydraulic treatment of the reactor core. Each channel represents a fuel pin, its cladding, the associated coolant, and a fraction of the subassembly duct wall. Other positioning hardware, such as wire wraps or grid spacers, is usually lumped into the structure field with the duct wall. Within a channel, the flow is assumed to be one-dimensional in the axial direction, and the temperature field in the fuel, cladding, coolant, and structure is assumed to be two-dimensional in the radial and axial directions. Usually, a channel represents an average fuel element in a subassembly or a group of subassemblies. A channel may also represent pins in blanket or control subassemblies. Alternately, a single channel may also be used to represent the hottest pin in an assembly, or any other subset of a subassembly. The axial extent of a channel covers the entire length of a subassembly, including the core, the axial blankets, the fission gas plenum and the spaces above and below the pin/cladding geometry. Different channels may be used to account for radial and azimuthal design geometry, power, coolant flow, and burnup variations within the reactor core. From ten to thirty channels normally provide sufficient discretization, depending on the core design. Significantly more channels can be used if necessary.

([[Media:SAS_Ch01_-_Introduction_(ANL-NE-12-4).pdf|PDF]])

=== User's Guide ===

Chapter 2 contains a user's guide for SAS4A/SASSYS-1 applications, including a complete description of the standard input file.

=== Steady-State and Transient Thermal Hydraulics in Core Assemblies ===

Chapter 3 contains the description of the formulation for the SAS4A/SASSYS-1 pin heat transfer and single-phase coolant thermal/hydraulics model. A major new addition to this model is the capability to treat channel‐to-channel heat transfer due to conduction and convection at all axial locations between the channel inlet and outlet, permitting a consistent multiple-pin subassembly treatment. Also, the subassembly-to-subassembly heat transfer model has been improved, and axial conduction in the coolant has been added. These modeling additions have been proven in validation analyses of EBR-II Shutdown Heat Removal Tests, and are required for accurate predictions of intra-subassembly flow and temperature variations in EBR-II during transients from normal to shutdown operating conditions.

=== Reactor Point Kinetics, Decay Heat, and Reactivity Feedback ===

Chapter 4 contains the description of the formulation for the SAS4A/SASSYS‐1 reactor points kinetics and reactivity feedback models. The new addition to this module is an option for an EBR-II-specific reactivity feedback model that is being validated with analysis of reactor operating data and used for predictive calculations of margins in design basis analyses. This module provides the reactor power level to the core thermal/hydraulics models for determination of the heating rate in the fuel, and receives core materials temperature and geometry information to calculate the reactivity feedbacks employed in the solution of the point kinetics equations.

=== Primary and Intermediate Loop Thermal Hydraulics Module ===

Chapter 5 presents a full description of the formulation for the PRIMAR-4 sodium loops thermal/hydraulic model. This model provides boundary coolant pressure and flow conditions for the core channel models, including transient heat losses through normal and emergency heat removal systems and the transient performance of pumps. The major new addition to PRIMAR-4 is the option for multiple core inlet and outlet coolant plena, permitting exact representation of the actual EBR-II coolant systems geometry.

=== Control System ===

The plant control and protection system model described in Chapter 6 is unchanged from prior versions of SASSYS-1, except for the addition of an option to allow dynamic allocation of model data storage at execution time.

=== Balance of Plant Thermal/Hydraulics Models ===

The balance-of-plant (BOP) model described in Chapter 7 is new for SAS4A/SASSYS-1. It was implemented to permit 1) improved simulation of EBR-II design basis transients, 2) whole-plant analysis of IFR designs for optimization of advanced reactor control system strategies, and 3) core temperature margin assessments in unprotected accident sequences (i.e. beyond design basis accidents (BDBA) and anticipated transients without scram (ATWS)). In these latter sequences, core response depends strongly upon the performance of the balance-of‐plant, because the core neutronic and thermal/hydraulic behavior is determined by the availability of heat sinks outside the core. The BOP model couples to PRIMAR-4 at the steam generator.

=== DEFORM-4: Steady-State and Transient Pre-Failure Pin Behavior ===

Chapter 8 provides a description of the DEFORM-4 fuel element behavior model for stainless steel‐clad oxide fuel, which is unchanged from prior versions of SAS4A/SASSYS-1.

=== DEFORM-5: Metallic Fuel Cladding Transient Behavior Model ===

Chapter 9 contains the description of the DEFORM‐5 model, which treats the transient behavior of stainless steel and advanced (HT-9) cladding for metal fuel elements. This is a new model for SAS4A/SASSYS-1, and is aimed at predicting margin to cladding failure, and timing and location of failure in limiting transients. It includes physical phenomena unique to metallic fuel, such as fuel/cladding chemical interactions.

=== SSCOMP: Pre-Transient Characterization of Metallic Fuel Pins ===

The SSCOMP model described in Chapter 10 has been revised to reflect newly available metal fuel material properties evaluations recorded in the IFR Material Properties Handbook. An efficient correlation technique has been implemented in all SAS4A/SASSYS-1 material properties routines that accurately generates the data from the IFR Handbook for use in all the modules of the code. It is planned to revise the material migration capability in SSCOMP for ternary fuel, to add models for fission gas generation and release, swelling, and all other phenomena needed to describe the transition from cold, clean, unirradiated conditions to hot irradiated conditions.

=== FPIN2: Pre-Failure Metal Fuel Pin Behavior Model ===

Chapter 11 contains the description of a major new addition to the SAS4A/SASSYS-1 cods, the FPIN2 metal fuel pin mechanics model. FPIN2 is a validated model for metal fuel pin transient behavior. Unlike DEFORM-5, which treats only the cladding response, FPIN2 provides a finite-element solution of the fuel and cladding mechanics equations for the elastic/plastic response, including fission gas pressurization and migration, molten cavity formation and growth, and fuel/cladding chemical interaction and cladding thinning. The interface between SAS4A/SASSYS-1 and FPIN2 has been designed to permit stand-alone execution of FPIN2 for direct verification, and to replace the FPIN2 thermal/hydraulics calculation with the SAS4A/SASSYS-1 counterparts for coupled calculations. The application for this model is design basis analysis of driver and experimental fuel elements in EBR-II for the purpose of margin-to-failure assessments.

=== Sodium Voiding Model ===

The TSBOIL module for liquid metal coolant boiling and two-phase thermal/hydraulics calculations has been retained intact from previous versions of SAS4A/SASSYS-1, with the addition of a set of modifications to describe the sudden release of noncondensible fission gas from a cladding rupture in the upper fission gas plenum of metal fuel elements and the subsequent plenum blow-down and liquid coolant expulsion. This option has been used to assess the safety implications of long-term fuel element irradiations in EBR-II.

=== Cladding Motion Model — CLAP ===

The CLAP model described in Chapter 13 is relevant only to oxide fuel, and have remained unchanged since the previous documentation.

=== Non-Voided Channel Fuel Motion Analysis — PLUTO2 ===

The PLUTO2 model described in Chapter 14 is relevant only to oxide fuel, and have remained unchanged since the previous documentation.

=== PINACLE: In-Pin Pre-Failure Molten Fuel Relocation Module ===

The PINACLE model described in Chapter 15 and the LEVITATE model described in Chapter 16 have been upgraded for applications to metallic fuel. The model enhancements added to PINACLE and LEVITATE for metal fuel include fuel/cladding and fuel/structure chemical interactions and fission gas generation and migration with fuel swelling. Preliminary analyses of TREAT M-Series in-pile metal fuel tests have been completed, and applications to severe accident sequences in metal-fueled IFR cores have been completed and documented.

=== LEVITATE: Voided Channel Fuel Motion Analysis ===

See [[#PINACLE: In-Pin Pre-Failure Molten Fuel Relocation Module]]

File:SAS Ch01 - Introduction (ANL-NE-12-4).pdf

2012-05-03T13:33:19Z

Fanning:

PDF Code Manual

2012-05-03T13:12:27Z

Fanning: Chapter descriptions have been added

== The SAS4A/SASSYS-1 Safety Analysis System (ANL/NE-12/4) ==

The manual for SAS4A/SASSYS-1 describes code capabilities, user input, and major code modules. It is divided into 16 chapters.

=== Introduction ===

SAS4A/SASSYS-1 provide a detailed, multiple-channel thermal/hydraulic treatment of the reactor core. Each channel represents a fuel pin, its cladding, the associated coolant, and a fraction of the subassembly duct wall. Other positioning hardware, such as wire wraps or grid spacers, is usually lumped into the structure field with the duct wall. Within a channel, the flow is assumed to be one-dimensional in the axial direction, and the temperature field in the fuel, cladding, coolant, and structure is assumed to be two-dimensional in the radial and axial directions. Usually, a channel represents an average fuel element in a subassembly or a group of subassemblies. A channel may also represent pins in blanket or control subassemblies. Alternately, a single channel may also be used to represent the hottest pin in an assembly, or any other subset of a subassembly. The axial extent of a channel covers the entire length of a subassembly, including the core, the axial blankets, the fission gas plenum and the spaces above and below the pin/cladding geometry. Different channels may be used to account for radial and azimuthal design geometry, power, coolant flow, and burnup variations within the reactor core. From ten to thirty channels normally provide sufficient discretization, depending on the core design. Significantly more channels can be used if necessary.

=== User's Guide ===

Chapter 2 contains a user's guide for SAS4A/SASSYS-1 applications, including a complete description of the standard input file.

=== Steady-State and Transient Thermal Hydraulics in Core Assemblies ===

Chapter 3 contains the description of the formulation for the SAS4A/SASSYS-1 pin heat transfer and single-phase coolant thermal/hydraulics model. A major new addition to this model is the capability to treat channel‐to-channel heat transfer due to conduction and convection at all axial locations between the channel inlet and outlet, permitting a consistent multiple-pin subassembly treatment. Also, the subassembly-to-subassembly heat transfer model has been improved, and axial conduction in the coolant has been added. These modeling additions have been proven in validation analyses of EBR-II Shutdown Heat Removal Tests, and are required for accurate predictions of intra-subassembly flow and temperature variations in EBR-II during transients from normal to shutdown operating conditions.

=== Reactor Point Kinetics, Decay Heat, and Reactivity Feedback ===

Chapter 4 contains the description of the formulation for the SAS4A/SASSYS‐1 reactor points kinetics and reactivity feedback models. The new addition to this module is an option for an EBR-II-specific reactivity feedback model that is being validated with analysis of reactor operating data and used for predictive calculations of margins in design basis analyses. This module provides the reactor power level to the core thermal/hydraulics models for determination of the heating rate in the fuel, and receives core materials temperature and geometry information to calculate the reactivity feedbacks employed in the solution of the point kinetics equations.

=== Primary and Intermediate Loop Thermal Hydraulics Module ===

Chapter 5 presents a full description of the formulation for the PRIMAR-4 sodium loops thermal/hydraulic model. This model provides boundary coolant pressure and flow conditions for the core channel models, including transient heat losses through normal and emergency heat removal systems and the transient performance of pumps. The major new addition to PRIMAR-4 is the option for multiple core inlet and outlet coolant plena, permitting exact representation of the actual EBR-II coolant systems geometry.

=== Control System ===

The plant control and protection system model described in Chapter 6 is unchanged from prior versions of SASSYS-1, except for the addition of an option to allow dynamic allocation of model data storage at execution time.

=== Balance of Plant Thermal/Hydraulics Models ===

The balance-of-plant (BOP) model described in Chapter 7 is new for SAS4A/SASSYS-1. It was implemented to permit 1) improved simulation of EBR-II design basis transients, 2) whole-plant analysis of IFR designs for optimization of advanced reactor control system strategies, and 3) core temperature margin assessments in unprotected accident sequences (i.e. beyond design basis accidents (BDBA) and anticipated transients without scram (ATWS)). In these latter sequences, core response depends strongly upon the performance of the balance-of‐plant, because the core neutronic and thermal/hydraulic behavior is determined by the availability of heat sinks outside the core. The BOP model couples to PRIMAR-4 at the steam generator.

=== DEFORM-4: Steady-State and Transient Pre-Failure Pin Behavior ===

Chapter 8 provides a description of the DEFORM-4 fuel element behavior model for stainless steel‐clad oxide fuel, which is unchanged from prior versions of SAS4A/SASSYS-1.

=== DEFORM-5: Metallic Fuel Cladding Transient Behavior Model ===

Chapter 9 contains the description of the DEFORM‐5 model, which treats the transient behavior of stainless steel and advanced (HT-9) cladding for metal fuel elements. This is a new model for SAS4A/SASSYS-1, and is aimed at predicting margin to cladding failure, and timing and location of failure in limiting transients. It includes physical phenomena unique to metallic fuel, such as fuel/cladding chemical interactions.

=== SSCOMP: Pre-Transient Characterization of Metallic Fuel Pins ===

The SSCOMP model described in Chapter 10 has been revised to reflect newly available metal fuel material properties evaluations recorded in the IFR Material Properties Handbook. An efficient correlation technique has been implemented in all SAS4A/SASSYS-1 material properties routines that accurately generates the data from the IFR Handbook for use in all the modules of the code. It is planned to revise the material migration capability in SSCOMP for ternary fuel, to add models for fission gas generation and release, swelling, and all other phenomena needed to describe the transition from cold, clean, unirradiated conditions to hot irradiated conditions.

=== FPIN2: Pre-Failure Metal Fuel Pin Behavior Model ===

Chapter 11 contains the description of a major new addition to the SAS4A/SASSYS-1 cods, the FPIN2 metal fuel pin mechanics model. FPIN2 is a validated model for metal fuel pin transient behavior. Unlike DEFORM-5, which treats only the cladding response, FPIN2 provides a finite-element solution of the fuel and cladding mechanics equations for the elastic/plastic response, including fission gas pressurization and migration, molten cavity formation and growth, and fuel/cladding chemical interaction and cladding thinning. The interface between SAS4A/SASSYS-1 and FPIN2 has been designed to permit stand-alone execution of FPIN2 for direct verification, and to replace the FPIN2 thermal/hydraulics calculation with the SAS4A/SASSYS-1 counterparts for coupled calculations. The application for this model is design basis analysis of driver and experimental fuel elements in EBR-II for the purpose of margin-to-failure assessments.

=== Sodium Voiding Model ===

The TSBOIL module for liquid metal coolant boiling and two-phase thermal/hydraulics calculations has been retained intact from previous versions of SAS4A/SASSYS-1, with the addition of a set of modifications to describe the sudden release of noncondensible fission gas from a cladding rupture in the upper fission gas plenum of metal fuel elements and the subsequent plenum blow-down and liquid coolant expulsion. This option has been used to assess the safety implications of long-term fuel element irradiations in EBR-II.

=== Cladding Motion Model — CLAP ===

The CLAP model described in Chapter 13 is relevant only to oxide fuel, and have remained unchanged since the previous documentation.

=== Non-Voided Channel Fuel Motion Analysis — PLUTO2 ===

The PLUTO2 model described in Chapter 14 is relevant only to oxide fuel, and have remained unchanged since the previous documentation.

=== PINACLE: In-Pin Pre-Failure Molten Fuel Relocation Module ===

The PINACLE model described in Chapter 15 and the LEVITATE model described in Chapter 16 have been upgraded for applications to metallic fuel. The model enhancements added to PINACLE and LEVITATE for metal fuel include fuel/cladding and fuel/structure chemical interactions and fission gas generation and migration with fuel swelling. Preliminary analyses of TREAT M-Series in-pile metal fuel tests have been completed, and applications to severe accident sequences in metal-fueled IFR cores have been completed and documented.

=== LEVITATE: Voided Channel Fuel Motion Analysis ===

See [[#PINACLE: In-Pin Pre-Failure Molten Fuel Relocation Module]]

PDF Code Manual

2012-05-03T12:47:11Z

Fanning: Outline of code manual

== The SAS4A/SASSYS-1 Safety Analysis System (ANL/NE-12/4) ==

The manual for SAS4A/SASSYS-1 describes code capabilities, user input, and major code modules. It is divided into 16 chapters.

# '''Introduction'''
# '''User's Guide'''
# '''Steady-State and Transient Thermal Hydraulics in Core Assemblies'''
# '''Reactor Point Kinetics, Decay Heat, and Reactivity Feedback'''
# '''Primary and Intermediate Loop Thermal Hydraulics Module'''
# '''Control System'''
# '''Balance of Plant Thermal/Hydraulics Models'''
# '''DEFORM-4: Steady-State and Transient Pre-Failure Pin Behavior'''
# '''DEFORM-5: Metallic Fuel Cladding Transient Behavior Model'''
# '''SSCOMP: Pre-Transient Characterization of Metallic Fuel Pins'''
# '''FPIN2: Pre-Failure Metal Fuel Pin Behavior Model'''
# '''Sodium Voiding Model'''
# '''Cladding Motion Model — CLAP'''
# '''Non-Voided Channel Fuel Motion Analysis — PLUTO2'''
# '''PINACLE: In-Pin Pre-Failure Molten Fuel Relocation Module'''
# '''LEVITATE: Voided Channel Fuel Motion Analysis'''

SAS

2012-05-03T12:31:12Z

Fanning:

== SAS: Reactor Safety Analysis System ==

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives.

Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models in SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.

== Availability ==

Version 3.1.2 of the SAS4A/SASSYS-1 safety analysis system is available from the Energy Science and Technology Software Center. To view the code package description and to submit a request, please visit [http://www.osti.gov/estsc/details.jsp?rcdid=4805 ESTSC]. The distribution includes code manuals, executables for Mac and PC systems, and two sample problems. Export control restrictions may apply.

The manual is also available as a separate download.

== Resources ==

* [[Code Manual]]

* [[References]]

* [[Questions and Answers]]

* [[Preparing Input]]

* [[Execution]]

* [[Post-Processing Results]]

{{Template:Standard Footer}}

MediaWiki:Sidebar

2012-04-20T03:42:01Z

Fanning:

SAS

2012-03-12T19:34:19Z

Fanning:

== SAS: Reactor Safety Analysis System ==

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives.

Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models in SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.

== Availability ==

Version 3.1.2 of the SAS4A/SASSYS-1 safety analysis system is available from the Energy Science and Technology Software Center. To view the code package description and to submit a request, please visit [http://www.osti.gov/estsc/details.jsp?rcdid=4805 ESTSC]. The distribution includes code manuals, executables for Mac and PC systems, and two sample problems. Export control restrictions may apply.

== Resources ==

* [[References]]

* [[Questions and Answers]]

* [[Preparing Input]]

* [[Execution]]

* [[Post-Processing Results]]

{{Template:Standard Footer}}

References

2012-02-13T17:38:15Z

Fanning:

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.