Wade R. Marcum

Steady-State Thermal-Hydraulic Analysis of the Oregon State University TRIGA Reactor Using RELAP5-3D

Abstract Oregon State University has recently conducted a complete core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors Program. The goals of the thermal-hydraulic steady-state analysis were to calculate natural-circulation flow rates, coolant temperatures, and fuel temperatures as a function of core power, as well as peak values of fuel temperature, cladding temperature, surface heat flux, critical heat flux ratio, and temperature profiles in the hot channel for both the highly enriched uranium and low-enriched uranium cores. RELAP5-3D Version 2.4.2 was used for all computational modeling during the thermal-hydraulic analysis. This is a lumped parameter code forcing engineering assumptions to be made during the analysis. A single-hot-channel model’s results are compared to results produced from more refined two- and eight-channel models in order to identify variations in thermal-hydraulic characteristics as a function of spatial refinement.

Neutronic Analysis of the Oregon State TRIGA Reactor in Support of Conversion from HEU Fuel to LEU Fuel

Abstract In support of the conversion of the Oregon State TRIGA Reactor (OSTR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel, a comprehensive neutronic analysis utilizing MCNP5 was performed on the HEU and LEU core configurations. The initial 1974 HEU core provided an opportunity for verification of the MCNP5 baseline model; all fuel elements in the initial core were congruent in geometry and material composition, having no burnup. In addition, a substantial database of core parameters was documented during the initial HEU core start-up. This verification study examined control rod worth, core excess reactivity, burnup, core power, power per element, temperature coefficient of reactivity, void coefficient of reactivity, moderator coefficient of reactivity, axial and radial power profiles, prompt-neutron lifetime, effective delayed-neutron fraction, power defect, and xenon poisoning. Fuel material composition and core loadings are presented. The excellent comparison between the numerical results and the experimental data of the initial HEU core established an objective, credible baseline model and methodology, which were then extended to the LEU core neutronic analysis. Comparison between the numerically calculated core physics values for the new LEU core and data collected during start-up provided a complete verification that the MCNP5 models developed for both the HEU and LEU cores were representative of the OSTR.

A Comparison of Pulsing Characteristics of the Oregon State University TRIGA Reactor with FLIP and LEU Fuel

Abstract The Oregon State TRIGA Reactor (OSTR) was converted from highly enriched uranium (HEU) Fuel Life Improvement Program (FLIP) fuel to low-enriched uranium (LEU) fuel in October 2008. This effort was driven by the U.S. Department of Energys Reduced Enrichment for Research and Test Reactor program. The new LEU fuel is 30/20 U-Zr1.6H (30% uranium in the fuel matrix, 19.75 wt% enriched) in contrast to the FLIP fuel having U-Zr1.6H (8.5% uranium in the fuel matrix, 70 wt% enriched). This new fuel composition provides the best match in performance of the available mixture ratios when compared to the previous FLIP fuel. To support conversion, a complete assessment and reevaluation of the OSTR Safety Analysis Report was performed. This evaluation included steady-state thermal-hydraulic and neutronics characterizations of the HEU and LEU cores as well as a transient behavior (pulse) analysis of both core types. This paper presents a summary of the methods used and results produced during the pulse analysis identifying power, temperature, and reactivity during pulsed operation for the FLIP and new LEU fuel. This analysis was performed using RELAP5-3D version 2.4.2 and point reactor kinetics simulation software; these two methods are found to agree very well. We discuss the differences between the two fuels and the impact of pulse behavior on the safety limits for the converted reactor.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

The Advanced Test Reactor (ATR) is conducting scoping studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low-enriched uranium (LEU) composition, through the Reduced Enrichment for Research and Test Reactors Program, within the Global Threat Reduction Initiative. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary reactor physics scoping and feasibility analysis of TRIGA fuel within the current ATR fuel element envelope and compares it to the functional requirements delineated by the Naval Reactors Program, which includes >4.8×1014 fissions/s·g−1 of 235U in test positions, a fast–to–thermal neutron flux ratio that has a <5% deviation from its current value, a desired steady cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other design parameters outside those put forth by the Naval Reactors Program that are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time. The result of this study demonstrates potential promise for implementation of TRIGA fuel in the ATR from a reactor physics perspective; discussion of observations and limitations are provided herein.

On the Steady Mechanical Response of a Heterogeneous Fuel Plate

Abstract An exact solution for a heterogeneous, discontinuous wide beam is developed herein having three unique boundary condition (BC) sets. These BC sets were chosen based on their relevance to plate-type–fueled reactors. This solution, with respective BC sets, contributes new insights into the field of engineering mechanics as applied within the nuclear engineering discipline. Herein, the exact solution is validated under a set of test cases. A comparison is then made against other relations that have been developed to provide similar engineering insights. Last, the solution is applied toward a presently relevant engineering problem regarding the design of a monolithic plate-type fuel. The outcome of this work provides a solution form for the computation of out-of-plane deflection of a heterogeneous, discontinuous wide beam that can be easily applied in engineering mechanics and is flexible in use.

A New Molybdenum Production Element for Implementation in TRIGA Reactors: Thermal-Hydraulic Characterization

Abstract Oregon State University (OSU) has developed and patented a technology that produces 99Mo within a standard TRIGA reactor core and does not negatively impact safety bases for the operations of such reactor designs. This new technology, referred to as the “molybdenum element,” is intended on being demonstrated within the OSU TRIGA Reactor (OSTR) with figures of merit including 99Mo yield and operation. A comprehensive design and thermal-hydraulic analysis has been conducted to characterize the safety-related traits of the molybdenum element to facilitate a license amendment through the U.S. Nuclear Regulatory Commission to insert such an experiment in the OSTR. This study details the thermal-hydraulic characteristics of the molybdenum element exhibited within the OSTR under the three sets of conditions necessary to demonstrate the element’s safety. The study leverages the lumped-parameter code RELAP5-3D Version 2.4.2 for conduct of the primary body of this work. The first condition analyzes the molybdenum element’s response under steady-state, full-power operation; the second condition assumes that the inner region of the annular molybdenum element is blocked while remaining at full power; and the last condition considers several loss-of-coolant-accident scenarios. Key thermal-hydraulic parameters that may impact the safety of the OSTR are identified, presented, and discussed herein. The result of this study provides objective evidence through use of RELAP5-3D that the molybdenum element remains in a safe state during the steady and abnormal conditions considered.

Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization

Abstract The development, characterization, and qualification testing of nuclear fuel at Idaho National Laboratory’s Advanced Test Reactor (ATR) requires extensive design and analysis activities prior to the insertion of an irradiation experiment in-pile. Significant effort is made in the design and development phase of all in-pile experiments to ensure that the maximum feasible impacts of all necessary experimental requirements are satisfied. The advancement of fuel, cladding, and in-reactor materials technology in recent years has introduced complexities associated with the design and construct of in-pile experiments necessitating deeper understanding of boundary conditions and increasingly comprehensive observations resulting from the experiment. Each unique experiment must be assessed for neutronics response, thermal/hydraulic/hydrodynamic performance, and structural integrity. This is accomplished either analytically, computationally, or experimentally, or some combination thereof, prior to insertion into the ATR. The various effects are interrelated to various degrees, such as the case with the experiment temperature affecting the thermal cross section of the fuel or the increased temperature of the experiment’s materials reducing the mechanical strength of the assemblies. Additionally, the feedback between the experiment’s response to a reactor transient could alter the neutron flux profile of the reactor during the transient. Each experiment must therefore undergo a barrage of analyses to assure the ATR operational safety review committee that the insertion and irradiation of the experiment will not detrimentally affect the safe operational envelope of the reactor. In many cases, the nuclear fuel being tested can be double-encapsulated to ensure safety margins are adequately addressed, whereas failed fuel would be encased in a protective capsule. In other cases, the experiments can be inserted in a self-contained loop that passes through the reactor core, remaining isolated from the primary coolant. In the case of research reactor fuel, however, the fuel plates must be tested in direct contact with the reactor coolant, and being fuel designed for high neutron fluxes, they are inherently power-dense plates. The combination of plate geometry, high-power density, and direct contact with primary coolant creates a scenario where the neutronic/thermomechanic/hydrodynamic characteristics of the fuel plates are tightly coupled, necessitating as complete characterization as possible to support the safety and programmatic assessments, thus enabling a successful experiment. This paper explores the efforts of the U.S. High-Performance Research Reactor program to thermomechanically/hydromechanically characterize the program’s wide variety of experiments, which range from stacks of miniplate capsules to full-sized, geometrically representative curved plates. Special attention is given to instances where the combination of experimental characterization and analytical assessment has reduced uncertainties of the safety margins, allowing experiments to be irradiated that would otherwise not have passed the rigorous qualification process for irradiation in the ATR. In some cases, the combined processes have exposed flow and heat transfer characteristics that would have been missed using historical methods, which allows for more accurate and representative postirradiation assessments.