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Tests of effectiveness of chemical immobilization of lanthanide imigration in U-Zr alloy fuel
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Analysis on the Heat Transfer Phenomena Inside the Scaled-Down VHTR RCCS Riser Experiment
Very High Temperature gas-cooled Reactor (VHTR) that is one of GEN-IV reactors has been developed for the purpose of hydrogen production. VHTR is characterized by high operating temperature and stepped-up inherent safety given by passive safety systems such as Reactor Cavity Cooling System (RCCS). In case the active heat removal system losses its capability, RCCS significantly contributes to the inherent safety of VHTR by removing the decay heat transferred from the reactor core to the reactor vessel. Therefore, the performance of RCCS is the key parameter to ensure the safety of the reactor. Several computation and experiment researches on the performance of RCCS have been conducted. There are two types of RCCS depending on its coolant; one is the air-cooled type and the other is water-cooled type. This study focused on the former type. Due to the excessive height of the prototype RCCS, experiment facilities were constructed in reduced scale. Korea Atomic Energy Research Institute (KAERI) constructed the quarter-scale RCCS experiment facility, NACEF and Argonne National Laboratory (ANL) has constructed half-scale experiment facility, NSTF. University of Wisconsin (UW) also has the quarter-scale experiment facility and Texas A and M University has conducted separate effect test on the jet effect at the upper plenum of RCCS. From the reduced scale experiment result, the performance of RCCS such as a heat removal rate of RCCS and a peak temperature of reactor vessel were evaluated. When the reduced-scale result is applied to the prototype, however, a distortion could be brought about due to the reduced height of RCCS in the experiments. Especially, if the heat transfer regime inside the riser which roles as a heat exchanger to discharge the decay heat into the atmosphere is not identical, the riser wall temperature and the consequent reactor vessel temperature of the prototype could be incorrectly predicted. Thus, the convective heat transfer phenomena inside the riser duct both in the reduced scale and in the prototype is necessary to be investigated. In this study, the preliminary analysis on the heat transfer phenomena inside the RCCS riser was conducted. The subjects of the analysis were the prototype RCCS riser of PMR 200, and its scaled-down experiment facilities, NACEF. First, the geometry, flow and thermal conditions of the riser in PMR200 RCCS and NACEF were introduced with a scaling method. And then, the heat transfer regimes and the heat transfer coefficient corresponding to each condition were predicted using correlation and CFD calculation. Finally, planned activities including the experiment was introduced. (authors)
Preliminary Analysis of Reactivity Initiated Accident Separate Effects Mechanical Tests on Chromium-Coated Zirconium Cladding
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Advances in Integral and Separate Effects Experiments for Water Cooled Small Modular Reactors
This paper focuses on the progress being made in water-cooled small modular reactor (SMR) advanced integral effects and separate effects experiments for reactor licensing. SMRs, considered modular in design, are mostly factory-built and then shipped to the reactor site. Of the several types of SMR designs available, water-cooled SMRs are likely to receive regulatory approval faster than others, as most of the technologies involved (e.g., the fuel and coolant technologies) are matured. However, the unique safety systems of SMRs, which depend on specific SMR design features, require integral and separate effects experiments to achieve reactor licensing. Thus, of the many SMR designs being proposed, only a few have successfully undergone licensing and reached the final development and demonstration stage. Many nuclear vendors and newcomer companies are investing millions of dollars to develop integral and separate effects testing facilities for preparing final safety analysis reports to include in licensing applications. Development and analysis of these experimental facilities is costly and takes about four to five years. The unique challenges involved can be reduced when stakeholders synergistically apply lessons learned, knowing the critical role played by advancements in experiments that support the licensing of SMR safety systems. Identification of knowledge/research gaps with the phenomena identification and ranking table (PIRT) and designing experimental facilities focusing on the phenomena of interest (POI) and figures of merit (FOMs) are pivotal to select the critical path to successful design demonstration and licensing application.
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.
Testbeds for test Readiness: Strategic Snapshot
Test bed engineering, monitoring for underground explosive and nuclear testing will be needed if the United States returns to a test posture of nuclear device testing. To maintain and rebuild the capability of test readiness, agile cost-effective test beds are needed and LANL as well as NNSS are the perfect locations to do this. Shallow and deep boreholes should be created to research containment science, diagnostic placement and optimization, coupling and emplacement science, tracer use for fracture and cratering research, and validation of test article performance.
High-Fidelity Modeling of Fuel-To-Coolant Thermomechanical Transport Behaviors Under Transient Conditions
This report summarizes the work completed under NEUP project number 21-24006. The objectives of this project are to advance the high-fidelity modeling capabilities and important phenomena that is important for high-burnup UO 2 and accident tolerant fuels (ATF) during transient conditions. Accurate modeling of the time-dependent phenomena that impact material performance must be used to determine the figures of merit and safety margin. Phenomena such as fuel fragmentation, cladding oxidation, pellet-clad interaction, clad ballooning, and clad rupture are examples that pose challenges to modeling during these transients. This project focused on the development of high-fidelity tightly coupled multiphysics models that can capture the time-dependent material response and associated thermal hydraulic conditions during these events. These models can then be validated against existing separate effects tests and in-pile integral experiments and will be used to model Transient Reactor Test facility (TREAT) loss-of-coolant accidents (LOCA) experiments. To achieve the project objective, we used a combination of NEAMS and NRC codes to model various LOCA test sets for the separate effects and in-pile integral experiments. BlueCRAB tool set, which can accurately predict material response at a sub-fuel pin level, as well as modeling the entire reactor system response to these events. Fuel performance was modeled using BISON (various versions) and FAST (version 1.2.1). BISON and FAST can model on a sub-fuel pin level the fuel performance under transient conditions. Both have simplified thermal hydraulic models that are capable of providing basic coolant boundary conditions. To better capture the thermomechanical interaction between the fuel, clad, and coolant, more sophisticated thermal hydraulic models are necessary.
Inter-laboratory testing of the effect of DNA blocking reagent G2 on DNA extraction from low-biomass clay samples [Inter-laboratory testing of the effect of DNA blocking reagent G2 on DNA extraction from low-biomass clay subsoil]
Here we show that a commercial blocking reagent (G2) based on modified eukaryotic DNA significantly improved DNA extraction efficiency. We subjected G2 to an inter-laboratory testing, where DNA was extracted from the same clay subsoil using the same batch of kits. The inter-laboratory extraction campaign revealed large variation among the participating laboratories, but the reagent increased the number of PCR-amplified16S rRNA genes recovered from biomass naturally present in the soils by one log unit. Furthermore, an extensive sequencing approach demonstrated that the blocking reagent was free of contaminating DNA, and may therefore also be used in metagenomics studies that require direct sequencing.
Operation Tumbler-Snapper
Operation Tumbler-Snapper began on April 1, 1952, when Able, a low-yield nuclear device, detonated 793 feet over the Frenchman Flat area of the Nevada Proving Ground. Able, the first of four airdrops conducted as the Tumbler phase of the Operation, provided the Department of Defense with reliable data on the relationship between height of burst and blast overpressure. Such information was vital to establishing the battlefield use of nuclear weapons. A final set of four tower detonations, the Snapper phase, provided the AEC and Los Alamos with diagnostic data on new weapon designs. Although the test series was nominally divided between the AEC and the DOD, this distinction held little meaning because two of the Tumbler effects tests, Charlie and Dog, employed experimental devices and all four of the Snapper tests involved effects experiments, including military troop maneuvers.
Evaluation of ceria as a surrogate material for UO 2 in experiments on fuel cracking driven by resistive heating
A variety of normal operation and accident scenarios can generate thermal stresses large enough to cause cracking in light-water reactor (LWR) fuel pellets. Cracking of fuel pellets can lead to reduced heat removal, larger centerline temperatures, and localized stress in cladding all of which impact fuel performance. Furthermore, pellet cracking also contributes to a temperature reduction in the pellet since the pellet fragments tend to move towards the heat sink (cladding), and the heat flow remains predominantly radial despite the presence of cracks. It is important to understand the temperature profile on the pellet before and after cracking to improve cracking models in fuel performance codes However, in-reactor observation and measurement of cracking is very challenging owing to the harsh environment and design of fuel rods. Recently, an experimental pellet cracking test stand was developed for separate effects testing of normal operations and accident temperature conditions, using thermal imaging to capture the pellet surface temperature for evaluation of thermal stresses and optical imaging to capture the evolution of cracking in real time. Cracking experiments were initially performed using ceria (CeO 2 ) as a surrogate fuel material, which is useful for developing and demonstrating the experimental approaches but is also valuable in its own right for cracking model development and validation. A combination of induction and resistance heating was used for volumetric heat generation in the pellet creating a thermal gradient. The material properties of CeO 2 and UO 2 are reviewed and compared for use in model development. Simulations of the experiment were performed to evaluate the behavior of the surrogate (CeO 2 ) fuel in BISON. The measured temperature profiles from BISON models match reasonably well with the observed experiments for the ceria pellets before cracking. The findings from this work will help improve confidence in fracture models used for fuel pellets under similar in-reactor conditions.
Fuel Performance Modeling Results for Representative FCRD Irradiation Experiments: Projected Deformation in the Annular AFC-3A U-10Zr Fuel Pins and Comparison to Alternative Designs
The objective of the present study was to predict the outcome of the AFC-3A annular U-10Zr fuel irradiation experiment. Specifically, the study attempted to predict whether the annular fuel will swell inward and fill the annulus, or swell outward resulting in an undesirable cladding deformation. It was predicted that both 55% and 75% annular fuels operating at 350 W/cm for 201 days will swell inward and partially fill the annulus which signifies a positive outcome of the experiment. The driving mechanism for such behavior is the fuel creep under the compressive stress exerted on the fuel by the cladding as a result of the fuel cladding mechanical interaction. Comparison with the solid fuel revealed that the annular fuel is expected to swell less early in life due to the mechanical constraint provided by the cladding. Furthermore, mechanical constraint is expected to yield a marked reduction of the axial elongation of the annular fuel as compared to the solid fuel. The prediction is based on the assumption that the annular fuel is not capable of moving axially relative to the cladding after the two come in contact. Post irradiation examination results would be very useful to assess the validity of this assumption. Specifically, underprediction of the axial growth would point to the fact that some slippage of the fuel relative to the cladding occurs after the onset of the FCMI. As fuel creep plays the major role in the deformation of the annular fuel, the contributions from the thermal and irradiation induced creep to the total creep rate of fuel were examined to provide guidance for possible creep testing experiments of the fuel (separate effect tests). It was found that irradiation induced creep dominates deformation of the fresh fuel at low temperatures. At high temperatures, and in the case of porous fuel, the thermal creep of the fuel becomes dominant and irradiation induced creep can be neglected. Fission gas induced porosity seems to accelerate fuel creep drastically. Recognizing the sensitivity of the fuel creep to the porosity, additional studies exploring this phenomenon and verifying published equations either experimentally or through computation may benefit the understanding of the annular fuel behavior.
TREAT Fuel Motion Summary Report. SETH A-E Experiments
In support of the Accident Tolerant Fuels program, experiment devices are being developed to enable transient testing of pressurized water reactor (PWR) type fuel specimens. In fiscal year 2019 several fueled tests were conducted in the Transient Reactor Test Facility (TREAT) using the Separate Effects Test Holder (SETH). Five different SETH capsules (SETH A-E) were assembled, each holding a 10-pellet rodlet with 4.9% enriched, commercially-produced fresh UO2 pellets in zirconium alloy cladding with pellet dimensions having typical PWR radial dimensions and a total pellet stack length measuring 10.16 cm. The rodlet’s length allows the specimen to be axially positioned in the center of the core with minimal flux variation over the height of the rodlet. The fuel motion monitoring system (FMMS) is located on the north side of the TREAT reactor. Having direct line-of-sight of the core enables the FMMS to collect spatially-resolved data for fast neutrons generated due to fission occurring in experiment specimens. The FMMS contains 360 channels that make up a 10 (horizontal) by 36 (vertical) array of viewing slots; for the SETH A-E test series 96 proton-recoil scintillator (PRS) detectors were installed in the system. The fast neutrons are detected in the PRS detectors from light produced by the recoil of protons interacting with ZnS(Ag) grains in each detector's scintillator button. This report focuses on the FMMS data captured in the SETH A-E experiments, with multiple transients varying in reactivity addition and length. The first three experiments had a total reactor energy of 101 MJ each and were critical in understanding the FMMS field of view, lead shielding, and other TREAT data acquisition systems. SETH-B2 and C had a reactivity addition of almost double the previous transients and served as a demonstration of the detector array behavior under increased reactor energy. The final two SETH experiments (D and E) were aimed at reaching cladding melting temperatures with a reactor energy over 500 MJ for each test. In preparation for the higher energy transients, the FMMS lead shielding was reduced to 0.5 inches from the 6 inches used in SETH A-C. This was expected to saturate the detectors at the peak of the transient in exchange for higher sensitivity at lower energies after the peak, when the motion was expected. SETH-D was clipped at approximately 3 seconds, which was too early for the FMMS to capture any rodlet motion. SETH-E had the same reactivity addition as SETH-D but the transient time was extended to 18 seconds. The FMMS captured the downward motion of the rodlet in SETH-E as the count rate significantly changed for the detectors viewing the specimen. Snapshots taken throughout the transient helped identify the approximate time at which the motion initiated. Further analysis of the individual detector data captured on a per-millisecond time basis resulted in a determination that the approximate time for the onset of the downward motion was 2.84 ± 0.02 seconds. Neutron radiographs confirmed that downward motion occurred for SETH-D and SETH-E. Motion in SETH-D could also have been captured by the FMMS if the transient time had been longer for SETH-D. Ultimately, these final two tests served to demonstrate the FMMS's ability to track fuel motion during transient experiments at TREAT.
FY20 SAM Code Developments and Validations for Transient Safety Analysis of Advanced non-LWRs
The System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report summarizes major progress in SAM code development, capability enhancements, demonstration, and validation to support transient safety analysis of advanced non-LWRs. Rapid developments continued in fiscal year 2020 (FY20) to support various needs of the advanced reactor community, especially the NRC and industry on the licensing safety analysis of advanced reactor designs. Significant code changes were made to provide various capability enhancements, bug fixes, and user friendliness improvements. Major code updates are summarized in Section 1, while four important enhancements are detailed in Sections 2-5, including a multi-dimension flow model; reactivity feedback and decay heat models; control and trip system modeling, and additional fluid and solid thermophysical property models. Code validation activities in FY20 include using test data from the Fast Flux Test Facility (FFTF), the High Temperature Test Facility (HTTF), and several separate effects test facilities for pebble-bed modeling.
UNCERTAINTIES FOR IN-REACTOR TESTING OF TRITIUM LOADED GETTERS WITH TEMPERATURE CONTROL
The fourth experiment in TPBAR Materials Irradiation Separate-Effects Test (TMIST) series (TMIST-4) is planned to experimentally determine the effect of neutron irradiation on the equilibrium vapor pressure of tritium over a tritium loaded Nickel-Plated Zirconium (NPZ) getter in a simulated TPBAR service environment. This pressure will be used to estimate the tritium permeation rate. In the thermal design in the TMIST-4, two evaluations are required. 1) The cladding temperature is calculated to estimate the feasibility of detecting potential irradiation enhancement of the partial pressure for an NPZ getter. This work is required to obtain a precise cladding temperature for measurement of a low tritium partial pressure. This work has been completed and indicated that the required/target cladding temperatures based on the TMIST-4 test plan can be satisfied under the test condition range by modifying the capsule component shapes and inserting a thermally conductive metal sleeve. 2) The evaluation of uncertainties associated with the test design and operation is performed. This includes fabrication tolerances, and power and nuclear operating conditions. These factors will impact the component temperatures inside capsules and resulting the tritium partial pressure. The tritium permeation rate is a function of temperature and tritium partial pressure. The prediction of tritium permeation rates affects the feasibility of obtaining measurement resolutions required to reach reasonable conclusion for potential irradiation enhanced performance. This work will investigate the uncertainties.
Effect of residual chips on the material removal process of the bulk metallic glass studied by in situ scratch testing inside the scanning electron microscope
Research on material removal mechanism is meaningful for precision and ultra-precision manufacturing. In this paper, a novel scratch device was proposed by integrating the parasitic motion principle linear actuator. The device has a compact structure and it can be installed on the stage of the scanning electron microscope (SEM) to carry out in situ scratch testing. Effect of residual chips on the material removal process of the bulk metallic glass (BMG) was studied by in situ scratch testing inside the SEM. The whole removal process of the BMG during the scratch was captured in real time. Formation and growth of lamellar chips on the rake face of the Cube-Corner indenter were observed dynamically. Experimental results indicate that when lots of chips are accumulated on the rake face of the indenter and obstruct forward flow of materials, materials will flow laterally and downward to find new location and direction for formation of new chips. Due to similar material removal processes, in situ scratch testing is potential to be a powerful research tool for studying material removal mechanism of single point diamond turning, single grit grinding, mechanical polishing and grating fabrication.
Mechanistic Multiscale Uncertainty Propagation in Support of Accelerated Fuel Qualification
Taking a nuclear fuel concept through the research, development, and qualification stages has historically taken on the order of 20 to 25 years because of extensive irradiation tests required for a variety of conditions. The concept of accelerated fuel qualification (AFQ) has been proposed to increase the innovation pace for nuclear fuels. The goal of AFQ is not to replace the traditional qualification approach but rather to reduce the total number of experiments required to ensure approval from the regulatory authority. Of the many AFQ approaches being explored, advanced modeling—and, in particular, mechanistic modeling—is in a uniquely cross-cutting position to reduce the number of required integral tests through the inclusion of separate-effects testing, while helping to extrapolate reactor performance during rare events. We make the case that propagation of uncertainty through various computational length scales helps contextualize mechanistic modeling. We will utilize UO 2 fission gas diffusion predictions from the atomistically informed cluster dynamics code Centipede to inform fuel performance rodlet simulations using the BISON finite element code as the metric for showing how multiscale mechanistic uncertainty quantification can help reduce uncertainty in fuel performance. In conclusion, by quantifying uncertainty and its reduction through multiscale modeling, the qualification process may be accelerated through the reduction of costly irradiation experiments.
Direct sensitivity analysis on the parameterization of crystal plasticity models
Various methods for calibrating crystal plasticity finite element (CPFE) models lead to non-unique input parameter values, which subsequently introduce uncertainty in the predicted mechanical response. Sensitivity analysis (SA) conducted on crystal plasticity models is used to identify how variability in these parameters contribute to output uncertainty. Traditional SA on CPFE parameters uses simplified surrogate models to save computational time. However, the accuracy of the surrogate models depends on the quantity of training data used, and any modeling error can propagate into the SA results, potentially affecting their reliability. In this work, the elementary effects test (EET) method, a global SA technique using direct CPFE simulations was employed, and the results obtained were compared with the First Order Second Moment (FOSM) method. ExaConstit, an open-source GPU-enabled CPFE code, was used to perform the simulations and direct SA. The EET method was accurately able to capture the non-linear effects of all the input parameters on the output and is a valuable approach for reliably attributing parameter sensitivities in CPFE models. Based on the results, efficient strategies to perform future parameter calibration and SA are discussed. Additionally, the SA trends observed in different single crystal orientations closely mirrored the activity of the slip systems.