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The xenon walk-away is a postulated uncontrolled criticality that could occur if operators are forced to evacuate the control room in certain circumstances. This poster shows how previously held assumptions are not entirely valid and that less restrictions should be placed on operators and engineers than initially assumed.
As part of the advanced manufacturing (AM) focus of the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) Program, new in-house capabilities to fabricate and test novel active and passive in-pile sensors are being developed at the Idaho National Laboratory (INL). These advanced-manufactured sensors, such as peak temperature monitors and strain gauges, must be miniaturized and be able to provide real-time feedback, all while withstanding the coupled extremes of temperature and radiation typically observed in the environment of a nuclear test reactor. Fabricating durable and robust in-pile sensors with high fidelity requires a fundamental understanding of the optimal process parameters required in the AM technique, as well as the effect of deviations from optimal process parameters on the sensor-substrate adhesion and the ultimate robustness of the sensor. This report presents the use of a laser-based, non-contact approach for characterizing the adhesion of sensors manufactured using aerosol jet printing (AJP) that comprise of silver nanoparticle inks deposited and sintered on austenitic stainless steel (SS) 316L substrates. The effect of substrate surface roughness, surface energy, ink sintering time and sintering temperature on the printed sensor topography and the sensor-substrate adhesion has been systematically investigated. The sensor-substrate interfacial adhesion strength was determined from the critical laser pulse energy required to ablate/ detach the printed sensor from the substrate. Measurements of the sensor topography using optical profilometry revealed that post-deposition sintering at 400°C for 60 minutes resulted in a reduced sensor thickness, while plasma treating the substrate prior to printing yielded sensors with uniform thickness profiles transverse to the print direction. Sintering temperature and time were found to be the dominant parameters that affected sensor-substrate adhesion, followed by substrate surface roughness. The laser-based and locally-destructive approach for adhesion strength measurements shows promise in overcoming the challenges and limitations of current standardized peel tests and can be developed into a post-fabrication process control protocol for evaluating the durability of AM printed sensors in a high throughput fashion. Future studies will investigate the use of laser-induced spallation techniques as well as non-destructive laser ultrasonic methods for sensor-substrate adhesion characterization on sensors manufactured using AJP and other AM techniques such as plasma jet printing (PJP).
Calculations are presented for the time evolution of 240 Pu from the proximity of the outer saddle point until the fission fragments are well separated, using the time-dependent density functional theory extended to superfluid systems. We have tested three families of nuclear energy density functionals and found that all functionals exhibit a similar dynamics: The collective motion is highly dissipative and with little trace of inertial dynamics, due to the one-body dissipation mechanism alone. This finding justifies the validity of using the overdamped collective motion approach and to some extent the main assumptions in statistical models of fission. This conclusion is robust with respect to the nuclear energy density functional used. The configurations and interactions left out of the present theory framework only increase the role of the dissipative couplings. An unexpected finding is varying the pairing strength within a quite large range has only minor effects on the dynamics. We find notable differences in the excitation energy sharing between the fission fragments in the cases of spontaneous and induced fission. With increasing initial excitation energy of the fissioning nucleus, more excitation energy is deposited in the heavy fragment, in agreement with experimental data on average neutron multiplicities.
Craters produced by chemical and nuclear explosives at the Nevada Test Site were used to train astronauts before their lunar missions. The craters have characteristics suitable for reconnaissance-type field investigations. The Schooner test produced a crater about 300 m across and excavated more than 72 m of stratigraphic section deposited in a fairly regular fashion so that systematic observations yield systematic results. Other features common on the moon, such as secondary craters and glass-coated rocks, are present at Schooner crater. Smaller explosive tests on Buckboard Mesa excavated rocks from three horizontal alteration zones within basalt flows so that the original sequence of the zones could be determined. One crater illustrated the characteristics of craters formed across vertical boundaries between rock units. Although the exercises at the Nevada Test Site were only a small part of the training of the astronauts, voice transcripts of Apollo missions 14, 16, and 17 show that the exercises contributed to astronaut performance on the moon.
The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.
Using the newly restarted Transient Reactor Test (TREAT) facility at Idaho National Laboratories, fresh (unirradiated) U 3 Si 2 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 U 3 Si 2 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 528J/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 U 3 Si 2 fuel and unirradiated SiC-SiC claddings can likely retain a “rod like” geometry for transient energy depositions up to 528J/g.
The Advanced Test Reactor (ATR), and complimentary zero-power ATR Critical (ATRC) reactor, located at Idaho National Labs (INL), are undergoing conversion from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). Both have a variety of testing locations that can receive large variations in flux due to its unique serpentine design, consisting of five lobes surrounding nine flux traps (see Figure 1). Initial criticality and power distribution throughout the core are controlled by core-external outer shim control cylinders (OSCCs). Distinct test loops allow for testing at specific temperatures, pressures, and irradiation conditions such as flux and fission density. The ATR is one of the key nuclear engineering research and testing facilities within the DOE National Laboratory Complex, and the ATRC supports its operation [1]. Currently, the Office of Material Management and Minimization (M3) within the National Nuclear Security Administration of the DOE is working to convert the remaining research reactors, including the ATR, from 93% HEU fuel to 19.75% LEU fuel (LEU) to support non-proliferation [2]. Extensive materials testing at INL and internationally has demonstrated that a high-density uranium molybdenum (U 10Mo) alloy can meet the performance requirements of the remaining high powered research reactors. The current LEU fuel element design is named the LOWE element. However, there are many technical challenges to address before the conversion to LEU can be successful, including the accurate characterization of the reactor core physics with LEU fuel. To ensure safe operation of the ATR, reactor engineers prepare a CSAP (Core Safety Assurance Package) before each cycle. The purpose of the CSAP is to verify the reactor performance calculation used to determine if the selected fuel loading meets operational, experimental, and safety criteria. Many of the criteria in the CSAP are limits on reactivity insertion in various accident scenarios.
The US High Performance Research Reactor Project has begun a new phase of fuel testing and post irradiation examination (PIE) of plate-type low-enriched uranium (LEU) U-10Mo Monolithic fuel. The testing campaigns are intended to first confirm that the fuel performance behavior of commercially produced fuel matches that of the lab-scale fabricated fuel tested during development, followed by qualification of the U-10Mo monolithic fuel system through the US Nuclear Regulatory Commission. Planned tests include miniature-plate, full-size plate, and fuel element irradiations, during which fuel test specimens fabricated with qualified processes will be irradiated under conditions that bound the NRC-regulated reactors in order to demonstrate that the fuel meets the overarching fuel performance requirements of 1) maintaining mechanical integrity 2) maintaining geometric stability and 3) exhibiting stable and predictable behavior. The MP-1 test, with the objective of confirming fuel performance behavior, is currently undergoing irradiation in the Advanced Test Reactor and PIE in INL hot cells. Status of upcoming qualification and licensing test design activities and highlights of initial as-fabricated characterization and testing results from the MP-1 test will be provided.
X-ray radiography offers the ability to noninvasively inspect the internal structure and material composition of objects. Dual-energy radiography allows material discrimination but is limited in quantitative material measurements due to the broad nature of the interrogating x-ray spectrum. Spectral radiography detectors provide an observation of not just the number of x rays that are transmitted through a sample but also the spectrum. Here, experimental validations of a method to use spectral x-ray radiography to accurately quantify the uranium mass in a powder are presented. An accurate system response model was developed, which includes a model of the incident x-ray source spectrum and a pixel-wise detector response that describes how the transmitted x rays were observed. Further, a calibration to account for the effect of partial x-ray attenuation was developed. As a result, the method is demonstrated to be capable of estimating the uranium mass in a variety of uranium oxide powders to a bias of –0.01±0.62%.
This document presents a test plan developed to assess the fundamental effectiveness of two widely accepted processes for drying spent nuclear fuel: vacuum drying and helium forced-gas drying. The experimental activities planned for Task 5 follow from the “Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan.” The objectives are to evaluate the response of the oxide corrosion layer on aluminum clad spent nuclear fuel (ASNF), processed at temperatures above 100°C, to understand and compare the behavior of oxide corrosion layer compositions resulting from a range of credible procedural variations during the transition of ASNF from wet storage to dry storage.
This is a brief summary of recent and past Ceramic HEPA Filter research and development activities under an NNSA NSR&D funded project and describes on-going plans for FY19 work as presently planned. Prior work has evaluated concepts for a ceramic HEPA filter including ceramic support materials, filter media including electrospun filter media, and metal housing designs to accommodate the conservative temperature performance target of up to 500°C. This has been supported by modelling efforts discussed with nuclear facility system engineers and their management. The temperature and time-attemperature are based on conservative, bounding temperatures from fire analysis presented as an EFCOG best practice paper for nuclear facility safety analysis. The overall goal is to improve safety of DOE nuclear facilities during fire scenarios by developing improved filtration technology including increasing temperature performance while maintaining acceptable filter pressure drop. Ceramic filtration technology can provide robust, passive mitigation against radiological releases. The goal of the ceramic HEPA filter is to provide a HEPA filter that provides increased temperature performance in an equivalent form factor (size and shape) that is similar to existing conventional HEPA filters used in facilities. Once development is complete and qualified for use in facilities, ceramic HEPA filters could simplify and reduce safety and support system costs, as well as lower life-cycle costs. Installation, monitoring, and maintenance of active fire suppression systems, as well as other related systems, are a considerable on-going cost to nuclear facilities. Reducing the level of dependence on Safety Class and Safety Significant fire alarm and fire suppression systems could allow significant cost savings. At present we are continuing research and development efforts towards the goal of a ceramic filter design, including improved ceramic filter media, that could reduce total operation cost for facilities utilizing HEPA filters.
This document presents a test plan developed to assess the fundamental effectiveness of two widely accepted processes for drying spent nuclear fuel: vacuum drying and helium forced-gas drying. The experimental activities planned for Task 5 follow from the “Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan.”[Connolly, INL/EXT-17-43908] The objectives are to evaluate the response of the oxide corrosion layer on aluminum clad spent nuclear fuel (ASNF), processed at temperatures above 100°C, to understand and compare the behavior of oxide corrosion layer compositions resulting from a range of credible procedural variations during the transition of ASNF from wet storage to dry storage.
This is a brief summary of recent and past Ceramic HEPA Filter research and development activities under an NNSA NSR&D funded project and describes on-going plans for FY19 work as presently planned. Prior work has evaluated concepts for a ceramic HEPA filter including ceramic support materials, filter media including electrospun filter media, and metal housing designs to accommodate the conservative temperature performance target of up to 500°C. This has been supported by modelling efforts discussed with nuclear facility system engineers and their management. The temperature and time-at-temperature are based on conservative, bounding temperatures from fire analysis presented as an EFCOG best practice paper for nuclear facility safety analysis. The overall goal is to improve safety of DOE nuclear facilities during fire scenarios by developing improved filtration technology including increasing temperature performance while maintaining acceptable filter pressure drop. Ceramic filtration technology can provide robust, passive mitigation against radiological releases. The goal of the ceramic HEPA filter is to provide a HEPA filter that provides increased temperature performance in an equivalent form factor (size and shape) that is similar to existing conventional HEPA filters used in facilities. Once development is complete and qualified for use in facilities, ceramic HEPA filters could simplify and reduce safety and support system costs, as well as lower life-cycle costs. Installation, monitoring, and maintenance of active fire suppression systems, as well as other related systems, are a considerable on-going cost to nuclear facilities. Reducing the level of dependence on Safety Class and Safety Significant fire alarm and fire suppression systems could allow significant cost savings, while maintaining safety. At present we are continuing research and development efforts towards the goal of a ceramic filter design, including improved ceramic filter media, that could reduce total operation cost for facilities utilizing HEPA filters.
Performance evaluation for NRX-A2 reactor instrumentation kit
Several plasma-sprayed and slurry-coated insulating coating systems were evaluated for structural stability in a low-pressure hot hydrogen environment at a maximum heat flux of 19.6 million watts/sq meter. The heat was provided by an electric-arc plasma generator. The coating systems consisted of a number of thin layers of metal oxides and/or metals. The materials included molybdenum, nichrome, tungsten, alumina, zirconia, and chromia. The study indicates potential usefulness in this environment for some coatings, and points up the need for improved coating application techniques.
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