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At least 289 records · Page 16

Pre-Transient Characterization of MLOF-1 Test Pin

This study focuses on the pre-transient characterization of U-10Zr test and sibling fuel pins for the THOR-M-LOF test series. Using neutron radiography, element contact profilometry (ECP), precise gamma scan (PGS), and gas assay, sampling, and recharge (GASR) analysis, it was confirmed that the fuel pins were intact and suitable for testing. Key fuel behaviors quantified include axial elongation, diametral strain, fluff structure geometry, axial isotope distribution, and fission gas release. Any deviations from historically expected behaviors were investigated and attributed to factors other than the irradiation behavior of the fuel pin. These pre-transient measurements establish a baseline for future post-transient analysis, which will be used to inform fuel performance models and safety criteria for sodium-cooled fast reactors (SFRs). The results will enhance understanding of transient fuel behavior and expand limited data on LOF scenarios.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Status Report for NEET In-Pile Sensor Irradiation Capabilities at TREAT

The recently-established in-pile sensor development project under the Nuclear Energy Enabling Technology (NEET) program is focused on the development of measurement technologies to enable acquisition of data focused on the performance, properties, and boundary conditions of nuclear fuels/materials under irradiation in nuclear reactors. These sensors can be used in steady state irradiation tests to measure parameters of interest under conditions representing normal nuclear plant operation, or in transient irradiation tests to determine specimen performance under postulated unplanned conditions. In either case, development of suitable sensor technologies must include instrument irradiations to help guide sensor material/design selection, verify instrument performance, and demonstrate deployment viability. The Transient Reactor Test facility (TREAT) while just a part of the overall capabilities portfolio needed to enable this sensor development mission, has unique capabilities and characteristics that make it well poised for a number of instrument irradiation data objectives including sensor survivability screening, response characterization, qualification tests, and in-situ properties measurements. TREAT’s brilliantly-basic design, when married with appropriate irradiation hardware and analytic capabilities, fills a crucial capability gap between out-of-pile testing and irradiation in steady state high power research reactors. Leveraging a recently developed approach to irradiation testing in TREAT, the present work demonstrates that cost-effective and scientifically-impactful instrument irradiation capabilities for TREAT are a well-placed capability in the NEET in-pile sensor development project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Testing of graphite thermal neutron scattering law data in support of treat neutronic analysis

The Transient Reactor Test Facility (TREAT) has historically provided the capability for transient fuel performance and safety testing in the United States. The operation of this facility was suspended in 1994. However, in recent years preparations for the restart of TREAT were initiated. TREAT is a graphite moderated and air cooled reactor. Its core is composed of a 19 x 19 array of fuel assemblies that contain UO{sub 2} (HEU) fuel particles dispersed in a graphite matrix, along with reflector assemblies filled with graphite only. Each fuel assembly is 8 feet tall with a fueled region that is 4 feet tall and upper and lower regions of graphite that are 2 feet tall. In addition, a 2 feet thick radial graphite reflector region is present. Finally, a 5 feet thick concrete biological shield surrounds the core. Previous work showed that the graphite thermal scattering cross sections can introduce variations exceeding 2000 pcm in the predicted eigenvalue of the TREAT core. Recognizing that the graphite in the assembly (and thereby the core) is a porous mixture of crystalline and amorphous carbon (with a lower density in comparison to ideal crystalline graphite), it becomes clear that TREAT specific thermal neutron scattering cross section libraries will be needed (at all temperatures of interest) for use within neutronic codes seeking to simulate TREAT. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Suitability of Type-II Fiber Bragg Gratings in Silica Optical Fiber for Temperature Sensing in TREAT

Optical fibers provide a variety of options for instrumentation in reactor environments. Optical fibers can be used to measure multiple physical phenomena including, temperature, strain, pressure, and fluid level. In addition to the various sensing applications, optical fibers are immune to electromagnetic interference, have a small footprint (~100 μm), and a fast response. The Department of Energy and Idaho National Laboratory have considered optical fibers for use as in-pile instrumentation in the Transient Reactor Test Facility (TREAT). TREAT was designed to test reactor fuels under accident conditions by replicating accident conditions for a variety of reactor transients, such as those associated with a loss of coolant accident (LOCA). This work investigates silica fiber optic temperature sensors with inscribed type-II fiber Bragg gratings (FBGs) under conditions similar to those that would be experienced in a TREAT transient. Separate effects testing was used to test the sensors under high-temperature step transients and under irradiation up to a total fluence similar to that of TREAT. Specifically, this work investigates distributed temperature measurements, using the Optical Frequency Domain Reflectometry (OFDR) sensing technique, using a Luna Innovations Optical Backscatter Reflectometer (OBR) 4600, with silica optical fibers inscribed with type-II fiber Bragg gratings (FBGs). In conclusion, separate effects testing of type-II FBGs in silica optical fiber, to high temperature and to neutron fluences that are an order of magnitude larger than fluences that are anticipated for TREAT tests, demonstrate that type-II FBGs in silica optical fiber hold great promise for high-temperature reactor instrumentation in TREAT.

McCary, Kelly↗

Time Constants for Optical Fiber Temperature Sensing using a Thermal Resistance Approach

Optical fibers are immune to electromagnetic interference, have no moving parts, no applied voltages, no special environmental conditions, and they are extremely small (100's of microns in diameter). The U.S. Department of Energy is interested in extending this technology to extremely high temperature radiation environments for the purposes of developing in-pile instrumentation for the restart of TREAT (Transient Reactor Test Facility). TREAT is an air-cooled, thermal, heterogeneous test reactor facility that was designed to evaluate reactor fuels and structural materials under conditions that simulate various types of transient overpower and under-cooling conditions that may be postulated to occur in a nuclear reactor. Spatially distributed temperature measurements can be made using an Optical Backscatter Reflectometer (such as, model 4600, manufactured by Luna Innovations, Inc.), coupled to single-mode optical fiber that is placed in an in-pile nuclear reactor testing facility. In-pile temperature sensing for the TREAT reactor will need to have a fast response to temperature changes, since the reactor pulse will be on the order of 100 msec. The sensors must also be robust enough to be installed in the testing facility; this requires the fibers to be in a protective capillary tubing that is made of a tough material, such as stainless steel. The temperature response of the fiber will be delayed by the capillary tubing. This paper discusses the characterization and modeling of the thermal time constants of the sensors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Forward and inverse predictive transient models of TREAT using surrogate reactivity models

In this work, we present two novel approaches for predicting the power evolution and control rod height of the Transient Reactor Test Facility (TREAT) to support experiment modeling; specifically, transient analysis of the NASA-sponsored Sirius series of experiments. These approaches utilize steady-state Monte Carlo model, point kinetics model, and surrogate models to predict power evolution and control rod axial position during transient experiments. Both approaches were tested and validated against several Sirius experiments that were performed in the TREAT facility at different power levels. The validation test results show very good agreement with the experimental data, and the models were able to accurately predict the power evolution and the axial control rod position with an average error within 3.0%. Finally, this indicates that these approaches will help the reactor engineering team of the TREAT facility in preparing and predicting the power and temperature of the experiment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

SPC-71260 Rev 0 MARVEL Heat Extraction Subsystem Secondary Coolant Equipment (SCE) Design/Build

A. The Microreactor Applications Research Validation and Evaluation (MARVEL) reactor will offer experimental capabilities that are not currently available at DOE’s national laboratories. Idaho National Laboratory (INL), operated for the U.S. Department of Energy (DOE) by Battelle Energy Alliance, LLC (BEA) (Contractor hereafter) is procuring services for the design, analysis, fabrication, testing and delivery of a Secondary Coolant Equipment system (SCE). This specification contains the requirements for design, analysis, fabrication, testing and delivery of the SCE as described herein. The MARVEL reactor is a microreactor which uses eutectic sodium-potassium alloy (NaK) as a primary coolant. The primary coolant is circulated through four primary loops by natural convection of the coolant. In each loop is a closed well which will accommodate an intermediate heat exchanger (IHX) for extracting heat from the loop. These wells will be referred to in this specification as the “IHX wells.” It is intended for the IHX containment to also be filled with NaK. The MARVEL design team has determined that a Heat Extraction System (HES) using pumped NaK will be used to extract heat from the IHXs and deliver it to a downstream system for power generation or alternate process heat users. This Heat Extraction System will enable MARVEL operations including the ability to test, demonstrate, and address issues related to installation, startup, and operations. In addition, it will allow down-stream utilization of process heat for various uses. The objective of this specification is to develop the final design for the HES Secondary Coolant Equipment system (SCE) that will be used as the core of the HES. This system provides control of the NaK circulation between the MARVEL reactor and the subsequent process heat utilization systems. It does not include design of the Intermediate Heat Exchangers and piping inside the T-REXc pit in which the reactor is located. B. The MARVEL microreactor will be installed in the Transient Reactor Test Facility (TREAT) building in the Transient Reactor Test (TREAT) Micro-Reactor Experiment Cell (T-REXc) C. An INL Subcontractor has developed a conceptual design for this system per SPC-71145, referred to in that specification as the Process Heat Extraction System. SPC-71260 is based on the pumped NaK loop concept developed under SPC-71145. D. The SCE system design and (as option scope) fabrication shall be provided by the awardee of the subcontract (Subcontractor hereafter) pertaining to this Specification. Prior to shipment, the SCE will be fabricated, assembled, and tested at the Subcontractor’s facility. After successful completion of acceptance testing, the SCE and associated equipment will be shipped to the Materials and Fuels Complex (MFC) at the INL (Contractor’s Facility hereafter) to be installed by others in TREAT/T-REXc.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

An advanced experimental design for modified burst testing of nuclear fuel cladding materials during transient loading

Here an advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the “driver tube” concept and allows for high-speed testing with internal pressure impulses of 10–1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360° view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

OPTI-TWIST Test Report, FY2025

OPTI-TWIST is the out-of-pile testing prototype and instrumentation TWIST with analogous components without nuclear material. Fabrication, construction, and testing in FY2025 is described in this test report.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Code Benchmark of Depressurized Conduction Cooldown Transient in the High Temperature Test Facility

This paper presents results from modeling of a depressurized conduction cooldown (DCC) transient at the High Temperature Test Facility (HTTF) as part of the OECD-NEA Thermal Hydraulics Code Validation Benchmark for High-Temperature Gas-Cooled Reactors using HTTF Data . This paper briefly describes the benchmark and the models being used. It then presents a comparison of steady state and transient results based on the Problem 2 Exercise 1A and 1B definitions. We compare block and helium temperature distributions, mass flow distribution, and energy balance in steady state. All models show comparable mass flow distributions and energy balances. The temperatures within the core and outer regions are comparable in all models too, but inner reflector temperatures can vary significantly. Despite that, we find that the models are in good agreement for the full-power steady state. In the DCC, we look at block temperature at the core midplane and RCCS water exit temperature. The INL and ANL models are found to be in excellent agreement with one another on block temperature over time, while the agreement when the KAERI and NRG models are added into consideration is good. Differences in the transient heat removal from the RCCS cause the differences in block temperature over time in these models. The CNL models show similar trends to the INL, ANL, KAERI, and NRG models, but the temperatures are high because the volumes used in calculating the average temperature include the heater rods in the CNL models only. The HUN-REN model shows results that suggest significantly lower heat removal in the RCCS which merit further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

FOR-868 Microreactor Applications Research Validation and Evaluation (MARVEL) Project

The Microreactor Applications Research Validation and Evaluation (MARVEL) reactor will offer experimental capabilities that are not currently available at DOE’s national laboratories. The test bed will perform R&D on the fundamental features, operations, and behaviors of microreactor technologies and help industry partners quickly test, develop, and demonstrate their technologies. The liquid- metal cooled microreactor will produce energy using small amounts of high- assay, low-enriched uranium (HALEU) from available research materials. Its design is primarily based on existing technology and will be built using off-the- shelf components to the extent practical. The reactor will be built inside the Transient Reactor Test (TREAT) Facility Micro-Reactor Experiment Cell (T- REXC), a separate project governed by the requirements of FOR-684, “Transient Reactor Test (TREAT) Facility Micro-Reactor Experiment Cell (T REXC),” and outside the scope of this document.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

MCNP Analysis fo the Aqua-SETH Experiment in TREAT

The versatile Transient Reactor Test (TREAT) Facility’s capability to deliver customized transient nuclear power excursions gives researchers and developers a multitude of specialized testing options. In conjunction with reactorprescribed delivery of neutron flux, experiment design requires specimen packaging to meet TREAT safety containment requirements and experimenter test objectives. These objectives may range from the very simple separate effects tests (generally dry capsule) to the more complex integral scale testing wherein prototypic boundary or specimen reactor environment conditions are created. The historic MARK-III sodium flow loop that achieved thermalhydraulic prototypic test conditions serves as an example of the more complex experiment design package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Report on RIA Relevant Modified Burst Testing of ATF Cladding Materials

The mechanical performance of accident-tolerant fuel (ATF) cladding candidates in light-water reactors (LWRs) must be similar to or better than that of current conventional nuclear fuel claddings to reduce dose to the public and ensure a that the core coolable geometry is maintained during a postulated reactivity-initiated accident (RIA) in light-water reactors (LWRs). During an RIA event, the rapid thermal expansion of nuclear fuel can deform the cladding once the fuel–cladding gap closes. In some cases, the pellet–cladding mechanical interaction (PCMI) can induce mechanical failure in ATF candidates. Thus, the mechanical response of ATF cladding must be investigated by mimicking the conditions of RIA and potentially performing Transient Reactor Test (TREAT) experiments to establish or verify the safety envelope. The work presented in this report investigated the failure behavior of as-received, hydrided, and chromium-coated (Cr-coated) Zircaloy-4 (Zry-4) cladding tube under strain-driven mechanical conditions, mimicking postulated RIA loading conditions. Mechanical testing was performed at 300°C via modified burst test (MBT) equipment with pulse width control previously developed under the Department of Energy’s (DOE’s) Advanced Fuel Campaign (AFC). The mechanical strains were determined using 2D digital image correlation (DIC) techniques. The base Zry-4 acquired by Cameco Inc. was in stress-relieved annealed (SRA) condition. Because of the observed large deformation of the cladding tubes, the failure strain definition was updated for the MBT, which can also be applied to other tube tests where significant bulging (out-of-plane deformation) is present. The failure strain was determined to be affected by the speed of the test or the RIA event. As the RIA-like event duration decreased from 75 to 15 ms, the failure strain decreased 5, 7 and 1% for as-received, hydrided, and Cr-coated specimens, respectively. Fractography on the Cr-coated specimens indicated the presence of two failure mechanisms: (i) crenulation at the outer surface of the coating due to the tensile tractional forces along with coatings grain microstructure and (ii) formation of critical defect at the coating/cladding interface that initiated coating rupture after severe plastic deformation of the Zry- 4 substrate. Based on the MBT results and fractography observations, performing mechanical property testing at high strain rates—in particular on Cr-coated tubes—and semi-integral TREAT experiments are required future efforts to ensure ATF cladding performance during transients. The testing recommended would also inform the development of long-term generalized cladding technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations

As a result of the accident at Fukushima, the United States has made a concerted effort to develop fuel with enhanced accident tolerance for the current fleet of reactors. In 2012 Congress laid out a schedule for the development of accident tolerant fuel (ATF), the goal of which is to test in a commercial reactor by 2022. Idaho National Lab's (INL) Transient Reactor Test Facility (TREAT) is one of the locations that will be performing ATF testing and is scheduled for restart in 2018. TREAT was built to subject fuel to conditions ranging from mild transients to severe reactor accidents. However, typically a number of pre-test calibrations are required before evaluation of fuel can begin, which is costly and time consuming. Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT

The U.S. nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. Here, this work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles' silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

On the origin of permanent performance loss of anion exchange membrane fuel cells: Electrochemical oxidation of phenyl group

The durability of alkaline anion exchange membrane fuel cells (AEMFCs) is a critical requirement for implementing this technology in cost-effective energy conversion. However, the lifetime of current AEMFCs is short, ca. < 1,000 h and the cause of the performance loss over time is poorly understood. Here, we investigate the origin of AEMFC unrecoverable performance loss. We diagnosed the recoverable performance loss induced by carbonation and anode flooding during an extended-term test with transient operating conditions. Multiple extended-term tests under current densities and the electrochemical and analytical analyses suggest that the most prominent permanent performance loss of AEMFCs occurs by phenol formation as a result of electrochemical phenyl oxidation of cathode ionomeric binder. The produced phenol is acidic and neutralizes the quaternary ammonium group to lower the local pH. Finally, the impact of this degradation pathway is discussed to explain the reported lifetime performance of current AEMFCs.

08 HYDROGEN↗

Effects of Coolant Void Condition on Melt Fragmentation using Simulants in Initial Phase of Metal-fueled SFR Severe Accident

Metal fuel has inherent safety characteristics, especially in severe accident conditions. However, the metal-fueled SFR severe accident scenarios need to be categorized since overall severe accident mitigation strategies have not been prepared. In addition, these mitigation strategies should be classified in specific phases. In an initial phase of metal-fueled SFR severe accident scenarios, fuel melting, and cladding failures would occur. After the occurrence of the cladding breach, the molten metal fuel would be ejected into a coolant channel due to fission gas in fuel cavity or rapid evaporation of sodium bond. This is called Ex-pin phenomena. There are few studies considering the Ex-pin phenomena of metal fuel. Related representative studies are transient reactor test facility (TREAT) M series tests. The TREAT M series tests provided fundamental metal fuel performance in transient condition. They informed partial experimental data such as the amount of melt ejected into coolant channel after the cladding failure. However, these experimental data did not examine the melt behavior during the Ex-pin phenomena. In metal-fueled SFR, it is a known fact that the negative reactivity feedback occurs in the initial phase of severe accident. In an unprotected loss-of-flow (ULOF) accident, the void-reactivity feedback may be positive depending on where in the reactor core the sodium voiding occurs. However, molten metal fuel would be better dispersed upward compared to those of an oxide fuel and significant negative reactivity feedback could be induced through such behaviors. Thus, a key safety strategy to mitigate the accident progression at the initial phase of severe accident is to discharge melt upwards out of the core. In this paper, the upward discharge of melt is defined as melt levitation. Such melt levitation could ensure early termination of severe accident. There could be different melting behaviors inside the pin depending on reactor types or the type of fuel. In case of pressurized water reactor (PWR) oxide fuel-water system, the rapid power excursion would initiate melting of fuel at the surface of the pellets. In this paper, the ULOF accident for the metal-fueled SFR is considered as an accident scenario. To induce the melt levitation, the driving force should be clarified. Heo et al. observed a built-up vapor pressure was suggested as one of the driving forces for the upward dispersion of the molten materials. The following factors also have the potential to induce the melt levitation; the pressure induced by fission gas in the cavity or rapid boiling of sodium bond, natural circulation flow of the sodium, the sodium vapor flow generated by the difference between inlet and outlet plenum in the reactor core, the built-up vapor pressure due to the fuel-coolant interaction etc. These factors are balanced each other. They should be examined through the Ex-pin experiments. Before conducting a parametric study for each factor, the general behavior of the molten metal fuel should be investigated in the coolant channel. Since rapid sodium coolant boiling could occur in the initial phase of severe accident scenarios, the molten fuel would be ejected and fragmented in voided channel. Unlike previous studies concerning an un-voided channel, differences in melt behavior in the voided channel are expected. In present study, molten material behavior was evaluated to reveal their characteristics in voided channel. The Ex-pin experiments were performed in both un-voided and voided channel, thus such behavior differences were explicitly observed. In the lab-scale experiments, the un-voided and voided channel were reflected by coolant-filled and air-filled channel, respectively. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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). Further, 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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗