State University TRIGA Reactor Research Articles

Computational Optimization of 133m Xe Production in the Washington State University TRIGA Reactor

Here we present a new method of irradiating 132Xe capsules with neutrons to produce 133mXe gas standards that are used for radiation detector calibration at radioxenon measurement laboratories in support of the Comprehensive Test Ban Treaty (CTBT). This method is designed to maximize the production of 133mXe compared to 133Xe, both of which are competing products from the 132Xe(n, g) reaction. The 133mXe is produced at a much higher fraction for high-energy neutron absorptions in 132Xe (~50% for fast neutrons versus ~11% for thermal neutrons). We performed “spectral tuning” of the Washington State University (WSU) TRIGA reactor neutron spectrum inside the 132Xe ampules to maximize the number of fast neutrons and minimize the number of thermal neutrons available for 132Xe absorption. Spectral tuning analysis, done with Monte Carlo simulations, provided valuable insights into a future final design for a 132Xe irradiation capsule. With no spectral tuning, the fractional yield of 133mXe in the WSU reactor was ~11.7%. By surrounding the 132Xe capsule with a 0.5-cm-thick layer of tungsten and a 2.83-cm layer of europium (III) oxide and placing it in the reactor’s cadmium rotator tube next to the fuel elements, the fractional yield of 133mXe can be increased to 24.6%, a 111% increase in yield. Thus, by improving the fractional yield of 133mXe through spectral tuning, the CTBT will have better quality gas standards to use for radioxenon detector calibration to assist in the CTBT’s mission.

Computationally Optimized Irradiation Chamber Design for Production of 135Xe in the Washington State University TRIGA Reactor

This work summarizes the radiation transport–based design for a new D2O-moderated ex-core irradiation facility in the Washington State University (WSU) TRIGA reactor for optimization of 135Xe sources used for calibration and quality control testing of Xe gas detection equipment in support of the Comprehensive Test Ban Treaty (CTBT). Three-dimensional (3-D) particle transport analysis characterizing the WSU reactor core using MCNP6.2 (3-D Monte Carlo) and PENTRAN (3-D deterministic parallel SN) form the basis for the computational optimization. Excellent agreement between MCNP6.2 and PENTRAN predictions is observed. A fundamental fuel bundle depletion analysis is applied to enable a more accurate prediction of neutron flux and neutron spectrum distribution, which drives production rates of 135Xe and 133Xe. The results of various model simulations were used to inform recommendations for the final irradiation chamber design, which has been optimized for safe placement in the reactor tank prior to startup and will allow for insertion and rotation of xenon “bean” samples using existing WSU irradiation equipment, while remaining within operational parameters. The irradiation chamber is expected to produce samples that will remain viable for use in CTBT standards applications for durations 70% to 80% longer than samples produced using current procedures. Thus, this design is expected to improve CTBT-related calibrations and performance testing and to support the continued stability of the CTBT monitoring network.

Modified Fuchs-Nordheim model for use in reactor pulse measurements

Experimental reactor physics is constantly advancing new methods to determine parameters of interest in reactor systems. The development of analytical and semi-analytical solutions to characterize complex systems relies on the ability to isolate specific phenomena that may be represented by governing equations. Two values in particular, the delayed neutron fraction, β, and the prompt neutron lifetime, l, are directly related to the performance of a pulse-type reactor during a reactor transient. A great deal of research has been performed to determine these quantities with limited success, especially at high power conditions. The Fuchs-Nordheim model is capable of qualitatively predicting reactor power during the transient, but assumptions limit its ability to determine reactor kinetics parameters during the transient.A modification to the FN model is presented and compared with the FN model along with experimental data from the Oregon State University TRIGA reactor to independently measure of the β/l ratio of the OSTR experimentally with quantifiable uncertainty.

Fission products measured from highly-enriched uranium irradiated under 10B4C in a research reactor

Prior work has demonstrated the use of a natural B4C capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method to critical assemblies for performing nuclear data measurements at near 235U fission-energy neutron spectrum. Previous fission product measurements showed that the neutron spectrum achievable with natural B4C was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a B4C capsule 96 % enriched in 10B resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. Fission product yields measured following an irradiation of a sample with this new method and subsequent radiochemical separations are presented here.

Cumulative fission yields of short-lived isotopes under natural-abundance-boron-carbide-moderated neutron spectrum

This work expands the availability of energy-resolved short-lived cumulative fission product yields for 235,238,233U, 239Pu, and 237Np subjected to a 2$ pulse in the Washington State University TRIGA reactor. A boron carbide capsule tailored the neutron spectrum, creating a spectrum with an average energy of 700 keV, similar to fast reactor spectra and approaching that of 235U fission. Unique gamma spectra were collected from 4 min to 1 week after fission using high purity germanium detectors. Measured cumulative fission product yields generally agree well with published fast pooled fission product yield values from ENDF/B-VII, though a bias was noted for 239Pu.

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

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 Energy's 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.

Design and Validation of Optimized Feedforward with Robust Feedback Control of a Nuclear Reactor

Design applications for robust feedback and optimized feedforward control, with confirming results from experiments conducted on the Pennsylvania State University TRIGA reactor, are presented. The combination of feedforward and feedback control techniques complement each other in that robust control offers guaranteed closed-loop stability in the presence of uncertainties, and optimized feedforward offers an approach to achieving performance that is sometimes limited by overly conservative robust feedback control. The design approach taken in this work combines these techniques by first designing robust feedback control. Alternative methods for specifying a low-order linear model and uncertainty specifications, while seeking as much performance as possible, are discussed and evaluated. To achieve desired performance characteristics, the optimized feedforward control is then computed by using the nominal nonlinear plant model that incorporates the robust feedback control.

Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor’s power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified.

A New Fuel Management Plan for the Pennsylvania State University TRIGA Reactor Including Supporting Experiments and Calculations

The Pennsylvania State University Breazeale (TRIGA) Reactor (PSBR) has operated for 25 yr (440 MWd) using a mixed 12 wt% ZrHx-U and 8.5 wt% ZrHx-U fuel configuration (both enriched to 20 wt% 235U, and x, the ratio of H to Zr, is nominally 1.65). In this configuration, the most reactive 12 wt% ZrHx-U fuel is always in the B-ring. The B-ring is the innermost hexagonal ring, incorporating 6 fuel elements, and the C-ring is the next outward ring, having 12 fuel elements. PSBR experience during pulsing and steady-state operation indicates that with these configurations the maximum fuel temperatures should be reduced in order to extend the useful life of the 12 wt% ZrHx-U fuel. This is because during the past 10 yr, the fuel temperatures of the new fuel have been significantly higher than the original fuel. The instrumented fuel element (I-15) loaded into the core ~10 yr ago and the most recent batch of fresh 12 wt% ZrHx-U fuel elements (six total, including I-16 and I-17) measured temperatures more than 100°C higher than any previous instrumented fuel element. Subsequent pulsing of I-15 increased its measured fuel temperature to where it began to approach the limiting safety system setting. Recent pulsing of I-16 and I-17 caused their steady-state fuel temperatures to decrease slightly, but they remain high. The new fuel management plan reduces these fuel temperatures by replacing the used 12 wt% ZrHx-U fuel in the C-ring with fresh 12 wt% ZrHx-U fuel. The 12 wt% ZrHx-U fuel in the B-ring is replaced with 8.5 wt% ZrHx-U fuel. Experiments have been performed to verify the predicted core parameters for the new plan. The lifetime of the new 12 wt% ZrHx-U fuel should now be limited by its maximum allowed burnup, which has not occurred so far.