Transient Reactor Test Facility Research Articles

Analysis of Fission Matrix Databases with Nonuniform Fuel Temperature Profiles for Static and Transient Calculations

Monte Carlo codes are the most accurate way to solve the neutronics in a reactor core but can be computationally expensive, especially for when feedback effects are considered or for transient calculations. In this paper, we use the fission matrix (FM) method to perform static and transient calculations with point-kinetics equations and a quasi-static model for an adiabatic transient with feedback. This was applied to a three-dimensional model of the Transient Reactor Test Facility (TREAT) experimental reactor using the Monte Carlo code Serpent for reference calculations and to generate fission matrix databases (FMDBs). In previous works, FMDBs were generated with uniform fuel temperature profiles. Here, we analyze the use of FMDBs with nonuniform temperature profiles, for static and transient calculations. For static calculations, comparison between Serpent and the FM method using nonuniform and uniform FMDBs showed maximum differences in multiplication factors of 57.0 and 77.9 pcm, respectively. For the fission source distribution, comparisons showed maximum root-mean-square differences of 1.10% and 4.89% for nonuniform and uniform FMDBs, respectively. Similar results were obtained when using homogenized databases. Therefore, using nonuniform FMDBs produces a better approximation than uniform databases. For transient calculations, comparisons between both database sets showed differences of 0.9% and −1.8% for the peak and final total power.

Multi-Scale Characterization of Porosity and Cracks in Silicon Carbide Cladding after Transient Reactor Test Facility Irradiation

Silicon carbide (SiC) ceramic matrix composite (CMC) cladding is currently being pursued as one of the leading candidates for accident-tolerant fuel (ATF) cladding for light water reactor applications. The morphology of fabrication defects, including the size and shape of voids, is one of the key challenges that impacts cladding performance and guarantees reactor safety. Therefore, quantification of defects’ size, location, distribution, and leak paths is critical to determining SiC CMC in-core performance. This research aims to provide quantitative insight into the defect’s distribution under multi-scale characterization at different length scales before and after different Transient Reactor Test Facility (TREAT) irradiation tests. A non-destructive multi-scale evaluation of irradiated SiC will help to assess critical microstructural defects from production and/or experimental testing to better understand and predict overall cladding performance. X-ray computed tomography (XCT), a non-destructive, data-rich characterization technique, is combined with lower length scale electronic microscopic characterization, which provides microscale morphology and structural characterization. This paper discusses a fully automatic workflow to detect and analyze SiC-SiC defects using image processing techniques on 3D X-ray images. Following the XCT data analysis, advanced characterizations from focused ion beam (FIB) and transmission electron microscopy (TEM) were conducted to verify the findings from the XCT data, especially quantitative results from local nano-scale TEM 3D tomography data, which were utilized to complement the 3D XCT results. In this work, three SiC samples (two irradiated and one unirradiated) provided by General Atomics are investigated. The irradiated samples were irradiated in a way that was expected to induce cracking, and indeed, the automated workflow developed in this work was able to successfully identify and characterize the defects formation in the irradiated samples while detecting no observed cracking in the unirradiated sample. These results demonstrate the value of automated XCT tools to better understand the damage and damage propagation in SiC-SiC structures for nuclear applications.

A non-fueled nuclear-heated rod for in-pile transient boiling studies

Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (Bnat ∼ 2.05 wt%) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-α thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. These results give confidence that the BNHR design has successfully met experiment objectives.

Resumption of water capsule reactivity-initiated accident testing at TREAT

A series of integral reactivity-initiated accident commissioning experiments were completed in a new static water capsule in the Transient Reactor Test Facility (i.e., TREAT), marking the first such tests in the United States in more than 40 years. The test campaign included a verification test followed by five tests in the Static Environment Rodlet Transient Test Apparatus capsule. The capsule’s initial conditions varied from room temperature and pressure up to 200 °C and 2.5 MPa, with energy depositions varying between ∼ 500–1,100 J/gUO2. The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. Post-transient examinations such as gamma emission spectroscopy, profilometry, and microscopy were performed to document the end state of the fuel rods. The results from the experiments show that the fuel rodlets behaved very similar to historical tests under similar energy depositions. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

Post-irradiation examination of the Sirius-1 nuclear thermal propulsion fuel test

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high-temperature gas-cooled fission reactor to provide thermal energy to heat the coolant, which also serves as the propellant. The fuel system consists of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, which has been irradiated in Idaho National Laboratory's (INL) Transient Reactor Test Facility (TREAT). This enabled evaluation of these materials under representative nuclear heating rates (∼95 K/s) and peak surface temperatures (∼2527 K). These surface conditions were indicative of achievement of the target peak internal temperature of approximately 2850 K. These tests named Sirius-1 (UN–W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced minor, stable cracks in the fuel specimen and spalling of surface material. Volatilized uranium ‘soot’ was found deposited on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and the production of a laminar microstructure on the edge of the fuel system. Overall, the specimen was stable and remained within a coolable geometry upon conclusion of multiple thermal cycles to prototypical operating temperatures.

Characterization of micro-burnup treat irradiated U-22.5 at.% Zr and U-52.8 at.% Zr foils by transmission electron microscopy and X-ray diffraction

The uranium-zirconium (U-Zr) fuel system has been studied for decades, yet there remains a data gap of micro-burnup (BU) studies where the microstructure undergoes an initial transformation from as-fabricated to the true starting microstructure present in reactor. This study focusses on two U-Zr specimens, U-(22.5 and 52.8) at.% Zr, irradiated to <0.001 at.% BU (∼4.33E+15 F/cc) at 660 °C in the Transient Reactor Test Facility (TREAT). The samples were rapidly gas quenched (4–5 °C/min) to maintain the in-pile structure for post irradiation characterization and modeling input. The bulk features of the U-22.5 at.% Zr specimen were retained, suggesting an impact (beyond crystallographic texture) of the as-fabricated system. It was also found that Zr solubility in α-U increases to ∼18 at.%Zr at 660 °C, and subsequently influences phase fractions and grain boundary density. The U-52.8 at.% Zr alloy fully transformed into an equiaxed granular microstructure immediately upon reactor insertion, implying single phase γ-U-Zr at 660 °C. This work also identifies the retention of a face-centered-cubic Zr rind on the surface of the sample, suggesting that the mechanism of Zr-rind decomposition is not a trivial phase transformation at temperature, but rather likely a diffusion-based process.

Microstructural and phase changes in alpha uranium investigated via in-situ studies and molecular dynamics

A deeper knowledge of thermally induced microstructural and phase evolution in nuclear metallic fuel can be obtained using novel in-situ microscopic analyses. Such studies can provide information on the dynamics of phase transitions which is not possible with conventional postmortem characterization (post-irradiation examination). In this work, the behavior of alpha uranium (α-U) was investigated via in-situ heating tests in a transmission electron microscope. The main objective is to understand the microstructural and phase changes, such as defect annihilation and β phase formation and retention, observed in reactor in-pile transient studies at the Transient Reactor Test facility. Indeed, defect migration and rearrangement were observed within the α phase starting at 673 K; α→β phase transition was observed between 773 K and 1,073 K during the heating ramp (which is in the temperature window reported for α→β transition temperatures). Recrystallization and formation of nano grains was observed at high temperatures (over 1,073 K). Such recrystallization was possibly related to the formation of the γ phase. Finally, it was indeed observed that the β phase (but not γ phase) was retained at room temperature upon rapid cooling. Molecular dynamics studies support these experimental results and shows that the γ phase of pure uranium cannot be retained at room temperature if not stabilized with the addition of an alloying element.

EBR-II MOX Fuel Characterization Enabling ARES Phase I Testing

Pretransient characterization was performed for the Experimental Breeder Reactor II (EBR-II) mixed-oxide (MOX) fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and the United States. Continued collaboration under the Advanced Reactor Experiments for Sodium Fast Reactor Fuels project will investigate the transient performance of these rods in the Transient Reactor Test facility at Idaho National Laboratory in the MOXTOP-THOR experiment. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of 14.3 at. % (~134.4 GWd/t) in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success of this effort. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination. The methodology development and analysis activities utilized in this paper enable current experiment design work and provide the pathway through which measured data of this type can be further evaluated.

TREAT testing of additively manufactured SiC canisters loaded with high density TRISO fuel for the Transformational Challenge Reactor project

Specimens composed of a novel nuclear fuel architecture underwent testing in the Transient Reactor Test Facility (TREAT) to investigate their behavior in thermomechanical states only possible with this type of testing. Uranium mononitride fuel kernels, with coatings typical of Tristructural Isotropic (TRISO) systems, were loaded into additively manufacturing silicon carbide (SiC) canisters and treated by chemical vapor infiltration so that the TRISO particles were held in a fully ceramic SiC matrix. A metallic heat sink capsule was used in TREAT to test fresh fuel specimens under progressively higher energy depositions. Thermomechanical modeling and post transient examinations, including a unique application of x-ray computed tomography, showed that more energetic transients led to stronger temperature gradients, higher stresses, and increased level of fracturing in the specimens. While some specimens were subjected to challenging conditions, the level of fracturing observed did not show a measurable dimensional change or loss of specimen geometry. These results indicate that this fuel architecture has resilience to catastrophic failure under power pulse transients. Future experiments are recommended for specimens which have been previously irradiated in order to investigate the effects of fuel burnup and neutron fluence on transient fuel performance.

Multiphysics Simulation of the NASA SIRIUS-CAL Fuel Experiment in the Transient Test Reactor Using Griffin

After approximately 50 years, NASA is restarting efforts to develop nuclear thermal propulsion (NTP) for interplanetary missions. Building upon nuclear engine tests performed from the late 1950s to the early 1970s, the present research and testing focuses on advanced materials and fabrication methods. A number of transient tests have been performed to evaluate materials performance under high-temperature, high-flux conditions, with several more experiments in the pipeline for future testing. The measured data obtained from those tests are being used to validate the Griffin reactor multiphysics code for this particular type of application. Griffin was developed at Idaho National Laboratory (INL) using the MOOSE framework. This article describes the simulation results of the SIRIUS-CAL calibration experiment in the Transient Reactor Test Facility (TREAT). SIRIUS-CAL was the first transient test conducted on NASA fuels, and although the test was performed with a relatively low core peak power, the test specimen survived a temperature exceeding 900 K. Griffin simulations of the experiment successfully matched the reactor’s power transient after calibrating the initial control rod position to match the initial reactor period. The thermal-hydraulics model largely matches the time-dependent response of a thermocouple located within the experiment specimen to within the uncertainty estimate. However, the uncertainty range is significant and must be reduced in the future.

Dry in-pile fracture test (DRIFT) for separate-effects validation of ceramic fuel fracture models

Fracture is an important component of nuclear fuel behavior, and significant efforts have been invested into developing fuel performance models that are capable of accurately representing fracture. Usable data on the process of fracture propagation in nuclear fuel under realistic service conditions are very limited. To address this need, a series of separate-effects experiments were developed and performed at Idaho National Laboratory's Transient Reactor Test (TREAT) facility. These experiments employ a heat sink to radially remove heat from the fuel in a manner that approximates the effect of coolant in an operating light-water reactor (LWR). The test holder for these experiments is known as the Dry In-pile Fracture Test (DRIFT). A series of experiments employing DRIFT and TREAT were performed to provide data on the extent and nature of fracture in fresh fuel at various points during a ramp to full power. Novel aspects of these experiments include the way they employ a heat sink to replicate steady-state LWR conditions, as well as the use of fiber optic sensors for in-reactor thermal instrumentation. Details on the development of this experiment, experimental conditions, and resulting data (including in situ thermal measurements and post-irradiation imaging of fracture) are provided in this paper. LWR-equivalent powers ranging from 10 to 25 kW/m were tested using this apparatus. Cracking was visible at all power levels, with increasing cracking extent as the power level increased, although there was little difference in the cracking between the two highest-power tests, which had LWR-equivalent powers of 20 and 25 kW/m.

BISON automation to predict transient power history based off prescribed cladding temperature

Since the restart of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) in 2017, advanced testing using various modules within the minimal activation retrievable capsule holder system has been underway with other types being considered for future work due to customer interest. In this work, one such test module—the Temperature Heat-sink Overpower Response module (THOR)—was simulated within BISON with UO2 fuel with zircaloy cladding and U-10Zr fuel with HT9 cladding, and a steel-alloy heat sink with a desired customer prescribed maximum cladding temperature profile. The THOR module is an instrumented capsule within which fuel rods can be irradiated in a liquid–sodium-filled hole in a thick-wall metal heat sink so that, along with TREAT’s unique transient power-shaping capability, experimenters can simulate numerous transient irradiation conditions. For transient testing, specimen temperature history is a primary driver in experiment design, which is primarily driven by the balance of power input to the nuclear heated specimen and the heat-transfer boundary condition associated with a particular experiment design. A simple iterative method was implemented in Python to write a BISON linear heat generation rate (LHGR) power history to produce desired customer-prescribed cladding-temperature values. The BISON LHGR power profile will later be used to write the control rod positions within TREAT to yield the customer-prescribed temperature values. Not all customer prescribed temperature values were achieved when desired temperatures were decreasing due to system limitations of THOR. However, the iterative method shown allows for temporal power histories to be derived, such that all BISON-produced temperatures are within 5% of customer values when increasing or maintaining a constant cladding temperature. The iterative method used allows for the dimensions, materials, and material properties of the THOR module to be customer specific, as well as different customer-prescribed parameters, such as fuel temperature, to be used in place of maximum cladding temperature. This tool allows for different designs and transient scenarios to be explored without manual fitting, allowing for rapid development for customer needs.

Post-transient examination of performance of uranium silicide fuel and silicon-carbide composite cladding under reactivity-initiated accident conditions

• First transient testing of accident-tolerant fuel designs under transient nuclear heating reactivity-initiated accident conditions. • Claddings experienced dimensional changes but maintained primary rod-like geometries. • Elemental species diffusion from the U 3 Si 2 fuel to the zircaloy cladding was observed. • No elemental species diffusion from the U 3 Si 2 fuel to the silicon-carbide cladding was observed. • Through wall thickness cracking observed in the silicon-carbide cladding resulting from pellet cladding interaction from thermal expansion of the pellet. New fuels are being developed for light water nuclear reactors with the initial goal of improving accident tolerance. More recently, these new fuel systems are also being recognized for their potential to support higher fuel burnups and thus improved economics for reactor plants through improved fuel utilization. Fuel safety testing of these new fuels systems is being conducted under transient irradiation conditions in the Transient Reactor Test facility at Idaho National Laboratory. A test campaign including two rodlets of U 3 Si 2 fuel in Zircaloy-4 cladding and two rodlets of U 3 Si 2 fuel in Silicon-Carbide Composite cladding has been performed to begin the development of fuel-safety criteria in design-basis reactivity-initiated accident conditions. Post-Transient-Irradiation Examinations have been performed and found that the accident tolerant materials maintained their primary rod-like geometry. Elemental species diffusion was found to occur in the Zircaloy rodlets indicating rapid diffusion kinetics at the temperatures achieved. No elemental species diffusion was found in the Silicon-Carbide rodlets. However, through-wall thickness cracks were observed due to displacement loading of the cladding from thermal expansion of the fuel pellets. This work presents the first experimental evidence of the performance of these fuel systems under transient nuclear heating and representative reactivity-initiated accident energy depositions.

Validation of the Griffin application for TREAT transient modeling and simulation

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is a graphite-moderated, air-cooled facility that was restarted in 2018 and is specifically built to conduct transient reactor tests. The Department of Energy focused modeling and simulation efforts between 2015 and 2019 on obtaining multiphysics solutions using the Griffin reactor physics application. This research has: (1) developed a consistent analysis method that automatically maps the neutronics data generated with Serpent onto the Griffin model, (2) improved the predictive capability of the traditional MCNP and Point Kinetics models that are commonly used in TREAT modeling and the preparation of engineering calculations, (3) determined the average recoverable energy in TREAT fuel per fission event, and (4) calculated a dynamic energy coupling factor that demonstrates the dependency of the energy deposition in the experiments during transients. Three experiments conducted in 2018 form the basis for the validation of Griffin for the transient modeling of TREAT. The first two experiments are 1.5% and 2.6% Δk/k transient prescription experiments with flux wires placed in the M8CAL test rig. The third experiment is a 0.6% Δk/k with the Minimal Activation Retrievable Capsule Holder (MARCH) test rig and the Separate Effect Test Holder (SETH) capsule, a light-water reactor rodlet in dry conditions. The core behavior predicted by Griffin for the transient prescription transients agrees well with the measured core power, peak power, and reactor feedback. The exception is the total energy deposition, which is underpredicted. The core behavior for the MARCH-SETH transient is significantly influenced by delayed neutrons since the reactivity insertion is delayed critical (i.e. ρ<βeff). The Griffin model experiences a stronger temperature feedback, leading to a lower peak power and total energy deposition. Increasing the number of temperature tabulation points significantly improves the solution for the MARCH-SETH transient. In addition, updating the 1.5% Δk/k transient model to ENDF/B-VIII.r0 data resolves the discrepancy in the prediction of the power tail and the total energy deposition. This is attributed to the new graphite S(α,β) data, which leads to lower values of the temperature coefficient of reactivity. The predicted energy coupling factors in both wire and rodlet experiments is the paramount parameter for TREAT modeling and simulation. In both cases, the Griffin predictions compare very well with measured values and exhibit consistent behavior, which is not observed in MCNP predictions. This work shows that Griffin is able to resolve the time-dependent effects from the control rod movement and the core heat-up in both the core and experiments.

Sensitivity analysis of in-pile CHF experiments in the TREAT facility: Characterization of impacts of fuel system thermal properties

Understanding fuel-to-coolant heat transfer mechanisms under transient heating irradiation conditions is an important aspect of pressurized-water reactor (PWR) safety. The work here aims to elucidate the potential impacts of critical heat flux (CHF) events on nuclear fuel systems, utilizing the heat transfer time constant (HTTC) as the fundamental basis. A set of thermophysical material and heat transfer coolant properties was established to represent the HTTCs of two Transient Reactor Test (TREAT) facility experimental fuel/cladding designs. Pool and flow boiling conditions were used to model these fuel systems within TREAT experimental capabilities, and a power pulse was used to represent a transient insertion of energy. Using Sobol sensitivity analysis and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code, an investigation was conducted to identify the most important HTTC parameters pertaining to the output of the peak outer cladding temperature (POCT), maximum fuel centerline temperature, and the time of CHF.The maximum fuel centerline temperature was shown to be highly dependent on the UO2 volumetric heat capacity. Regarding the POCTs, DNB was determined to have a dominant effect on the heat transfer mechanisms of these fuel systems. Further, the HTTC significantly impacted POCT predictions with the UO2 volumetric heat capacity, the film boiling heat transfer coefficient, and the CHF multiplier being most influential regardless of coolant flow conditions. Gap variations were shown to influence the POCT in cases that experienced the DNB event. In non-DNB cases, the HTTC had no significant impact on POCT. For the time of CHF, variations in gap thickness, and the CHF multiplier were overwhelmingly dominant.

Dynamical System Scaling Application to Zircaloy Cladding Thermal Response During Reactivity-Initiated Accident Experiment

New fuel design and development currently require 20 to 25 years to be qualified for use by the nuclear power industry. The thermal-hydraulics community has taken advantage of scaling theory to design reduced-scale experiments that correctly preserve dominant key phenomena while quantifying distorted phenomena. These techniques can be leveraged in the design and analysis of fuel performance experiments to help reduce the timeline associated with fuel design and development. This study uses the Dynamical System Scaling (DSS) method to analyze cladding temperature data from the recent SETH-C experiment in the Transient Reactor Test Facility (TREAT) and accompanying BISON simulations to assess dynamic distortions occurring throughout the fast power excursion transient. The DSS analysis revealed that on the cooldown from peak cladding temperature, the fuel radial power profile is the most sensitive modeling parameter, with a heterogeneous radial peaking factor corresponding to the lowest distortion compared to a uniform energy deposition. For the heatup to PCT, the heterogeneous radial power profile corresponded to the shortest process action. Last, for the heatup to PCT, the gap conductance model sensitivity was quantified using process actionsm and showed that the default light water reactor gap conductance model corresponded to the longest process action.

Sensitivity analysis of in-pile critical heat flux experiments in TREAT for characterization of RIA power-transient effects

A reactivity-initiated accident (RIA) is one type of postulated design basis accident (DBA) that can cause a departure from nucleate boiling (DNB) event in pressurized-water reactors (PWRs). A DNB occurrence and its consequences depend on the thermophysical properties of the fuel components and coolant, characteristics of the transient energy insertion into the fuel rod, and the onset of the critical heat flux (CHF) phenomenon. To leverage the restart of the Transient Reactor Test (TREAT) Facility, an effort is currently underway to better understand the cladding-to-coolant heat transfer mechanisms and the CHF phenomenon under fast-transient irradiation conditions. This paper characterizes the impact of power transients on the thermal-hydraulic behavior of a TREAT Facility reactor heater rodlet CHF experiment to provide the priority of parameters that need to be investigated for an improved CHF model. Sobol sensitivity analysis methods and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code were used to identify key input parameters on the uncertainty in the prediction of peak outer- and inner-surface temperatures of the heater tube, as well as the time of the DNB event. A series of sensitivity analyses revealed the total energy deposition on the tube and the transient effects of power pulse had large impacts on the maximum temperatures. The CHF multiplier had the largest impact on the time occurrence of CHF. The overall results show the energy deposition rate in the tube is the most influencing factor to the manifestation of CHF and the resulting thermal-hydraulic behaviors of the tube. The multiplier for the CHF, which is interpreted as the predicted CHF value, has the largest Sobol indices for the time of the CHF in all cases, since it directly determines the occurrence of CHF. It is inferred that the uncertainties in the thermal-hydraulic behaviors of fuels increase with respect to the key parameters as the power pulse becomes broader, and an accurate estimation of the energy deposition rate is required to reduce the uncertainty in the evaluation of the integrity of fuel if the CHF is expected to occur near the peak power. The outputs are expected to provide rigorous interpretation of ongoing in-pile CHF experiments in the TREAT Facility reactor regarding thermal-hydraulic behavior of the fuel system aiming for a new transient in-pile CHF model.

Computational fluid dynamics assessment of the Helium-3 Enhanced Negative Reactivity Insertion (HENRI) system for the Transient REActor Test (TREAT) facility

The Transient REActor Test facility (TREAT) is an air-cooled, graphite moderated nuclear reactor that is used for irradiation testing of nuclear fuels and materials. The shutdown of the Halden reactor in Norway lead to an increased interest of an indigenous nuclear facility that can perform Reactivity Initiated Accident (RIA) tests for Light Water Reactors (LWRs). Idaho National Laboratory (INL) has been leading the effort of making minor modifications to the TREAT facility to enhance its testing capabilities so it can mimic the neutronic boundary conditions representative of a RIA for a LWR. The power burst of the TREAT reactor needs to be shortened from 89 to 40 msec. to properly irradiate the fuel sample under RIA conditions. INL has proposed to use helium-3, a strong neutron absorber, as a form of negative reactivity to shorten the pulse of the reactor. In support of the development of a transient control rod for the TREAT, a Computational Fluid Dynamics (CFD) model is needed to confidently simulate the density evolution of helium-3 in this system.

Transient testing of uranium silicide fuel in zircaloy and silicon carbide cladding

Using the newly restarted Transient Reactor Test (TREAT) facility at Idaho National Laboratories, fresh (unirradiated) U3Si2 fuel pellets were tested under reactivity-initiated accident conditions in dry helium capsules. These tests were performed with total energy depositions sufficient to induce melting of the U3Si2 fuel pellets based on available thermophysical property data. The instrumented experiments utilized in-pile thermocouples and infrared pyrometers to measure the cladding temperature during the transient, which are in reasonable agreement with predicted peak cladding temperatures using a simple BISON fuel performance model. This same model is then used to predict expected cladding stresses during the transients in areas where pellet cladding interaction is expected. It was shown that for transient energy depositions up to 528 J/g rod like geometry of the test articles was maintained. While it is acknowledged that more confirmatory testing is required in more prototypic environments, these first of their kind experiments have demonstrated that unirradiated U3Si2 fuel and unirradiated SiC-SiC claddings can likely retain a “rod like” geometry for transient energy depositions up to 528 J/g.

FIPD: The SFR metallic fuels irradiation & physics database

The DOE Advanced Reactor Technology (ART) program has supported efforts to recover and preserve metallic fuel data generated throughout the US sodium-cooled fast reactor (SFR) program. Those efforts have been focused on establishing databases of the experimental data that were mainly generated during the Integral Fast Reactor (IFR) program including data generated at Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and Transient Reactor Test Facility (TREAT) reactors, as well as out of pile data. The data is essential for future licensing activities of metallic fuel based advanced fast reactors. This paper describes the development of the SFR Metallic Fuels Irradiation & Physics Database (FIPD) and covers the scientific knowledge available in the database. The architecture of the FIPD is described by showing the available reactor operation data, fabrication, and post-irradiation examination (PIE) data, and other documents. The applications of the fuel database are also discussed.