Floyd E. Dunn

Verification and implications of the multiple-pin treatment in the SASSYS-1 liquid-metal reactor systems analysis code

As part of a program to obtain realistic, as opposed to excessively conservative, analysis of reactor transients, a multiple-pin treatment for the analysis of intrasubassembly thermal hydraulics has been included in the SASSYS-1 liquid-metal reactor systems analysis code. This new treatment has made possible a whole new level of verification for the code. The code can now predict the steady-state and transient responses of individual thermocouples within instrumented subassemblies in a reactor rather than just predicting average temperatures for a subassembly. Very good agreement has been achieved between code predictions and the experimental measurements of steady-state and transient temperatures and flow rates in the shutdown heat removal tests in the Experimental Breeder Reactor II (EBR-II). Detailed multiple-pin calculations for blanket subassemblies in the EBR-II demonstrate that the actual steady-state and transient peak temperatures in these subassemblies are significantly lower than those that would be calculated by simpler models.

The SASSYS LMFBR systems analysis code

The main purpose of the SASSYS code is to analyze the consequences of failures in the shutdown heat removal system of a liquid metal cooled fast breeder reactor. The code is also capable of analyzing a wide range of transients, from mild operational transients through severe hypothetical core disruptive accidents. Some important features of the code include a detailed multi-channel core thermal hydraulics treatment with transient flow redistribution between assemblies, a model for boiling of sodium in the fuel assemblies, a general thermal hydraulics treatment for an arbitrary arrangement of components in the primary and intermidiate heat transport loops, and a fast-running steam generator model. The numerical methods mainly use semi-implicit or fully implicit time differencing to provide large time step sizes and rapid calculations.

IMPROVEMENTS TO NEUTRON SLOWING DOWN THEORY FOR FAST REACTORS.

Continuous neutron slowing down theory has proved useful in problems associated with thermal reactors. There are, however, two principal problem areas which inhibit obtaining the full benefits of c...

Critical Heat Flux in TRIGA-Fueled Reactors Cooled by Natural Convection

Abstract An experimental study of low-pressure, natural convection critical heat flux (CHF) has been carried out with full-scale fuel pins, simulating typical Training, Research, Isotopes, and General Atomics (TRIGA) reactor conditions. The test section is an annular upwardly flowing channel formed by a round tube and a simulated fuel pin heater rod with a chopped-cosine power profile, located in the center of the tube. Experiments were performed under the following conditions: inlet water subcooling varying from 10 to 70 K, pressure varying from 110 to 200 kPa, and natural circulation mass flux up to 400 kg/m2·s. CHF was observed, and associated data have been compared with selected CHF correlations. It has been found that the CHF increases as the pressure or mass flux increases, but does not significantly depend on the inlet subcooling. Among the numerous presented CHF data and correlations, few data exist, and no specific correlations have been developed for TRIGA reactor conditions. Because of the lack of specific correlation, the correlations of Bernath, El-Genk et al., Mishima and Ishii, and Block and Wallis have been used to estimate the TRIGA CHF outside of their intended ranges of applicability. These correlations are evaluated against the current experimental data.

Study of Critical Heat Flux in Natural Convection–Cooled TRIGA Reactors with Single Annulus and Rod Bundle Geometries

Abstract A critical heat flux (CHF) experimental study at low pressure and natural convection condition has been conducted. The test apparatus is a natural circulation loop with an upward flow channel, simulating TRIGA (Training, Research, Isotopes, General Atomics) reactors. CHF is studied in three types of geometries: a single-rod annulus, a three-rod bundle in a trefoil tube, and a four-rod bundle in a square tube. The full-scale fuel pin heater rod is electrically heated with a prototypic axial power profile, equipped with thermocouples for CHF detection. Experiments are carried out at the following conditions: inlet subcooling from 10 to 70 K, pressure from 110 to 290 kPa, and mass flux from 0 to 400 kg/m2·s. It is observed that CHF increases as the pressure or mass flux increases but does not significantly depend on the inlet subcooling within the testing range. The current CHF data are compared with a few selected CHF correlations whose application ranges are close to the testing conditions. The relevance of the CHF to the testing parameters is investigated. A modified CHF correlation compatible with TRIGA reactor conditions is proposed based on a previous correlation and current experimental data.

Thermal-Hydraulic Analysis for HEU and LEU Transitional Core Conversion of the MIT Research Reactor

The Massachusetts Institute of Technology Research Reactor (MITR) is evaluating a transitional core conversion strategy for converting from high-enrichment uranium (HEU) to low-enrichment uranium (LEU) fuel. The objective of this study is to analyze steady-state operational safety margins and loss of primary flow (LOF) accidents for the postulated HEU-LEU transitional core configurations. The thermal-hydraulic calculation was performed using the RELAP5 MOD 3.3 code based on 7.40-MW reactor power, which is the limiting safety system settings of the current licensed reactor power of 6 MW. A lumped average and a single hot channel were modeled in each core configuration with radial peaking factors of 2.0 and 1.76 for HEU and LEU fuel elements, respectively. Four natural convection valves and two antisiphon valves were modeled for natural convective heat removal during the LOF transient. Two different hot-channel configurations and full- and side-channel geometries were evaluated because the unique design of the MITR fuel element can form these two types of geometries. RELAP5 calculation results suggest that the transitional core conversion strategy is feasible and that sufficient thermal-hydraulic safety margins can be maintained.

An Enhanced Code for the Safety Analysis of Pool-Type Sodium-Cooled Fast Reactors

Abstract The Super System Code of the Korea Atomic Energy Research Institute (SSC-K) has been developed for the transient analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) system. Recently, a detailed three-dimensional (3-D) core thermal-hydraulic model was developed to describe nonuniformities of radial temperature and flow within a subassembly and to decrease the uncertainties in the reactor safety margins during accident situations. The Shutdown Heat Removal Test-17 (SHRT-17) performed in the Experimental Breeder Reactor-II (EBR-II) and the postulated unscrammed events for the KALIMER conceptual design have been analyzed using a code system that has coupled a detailed 3-D core thermal-hydraulic model with SSC-K. The coupled code predicted behaviors for the experimental trends for the protected loss-of-flow SHRT-17. The KALIMER-150 design was adopted for a plant application of the same code system. Three events, unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), and unprotected loss of heat sink (ULOHS) were analyzed, and the simulation results were compared to those obtained using another code system that has coupled the Safety Analysis Section SYStem (SASSYS)-1 code with the same detailed 3-D core thermal-hydraulic model. The results, calculated with SSC-K coupled with the detailed 3-D core thermal-hydraulic model showed good agreement with the calculated results of the SASSYS-1 coupled code system for the UTOP and ULOF; however, some discrepancies were shown in the results for the ULOHS. These were found to have occurred because of a difference of the modeling for the decay heat removal system and primary coolant inventory. Through these analyses, the coupled code system was validated in order to be available for the safety analysis of a liquid-metal reactor (LMR) plant.