Idaho National Laboratory Research Articles

Mechanical properties of irradiated U[sbnd]Mo alloy fuel

This paper presents results of studies characterizing the mechanical properties of irradiated U10Mo fuel in support of low-enriched fuel conversion of several United States high-performance research reactors (USHPRR), specifically the monolithic base fuel qualification. Specimens were tested from ten test fuel plates irradiated at the Advanced Test Reactor (ATR) from the RERTR-12 and AFIP-6 Mk. II irradiation campaigns. Tests were conducted at Idaho National Laboratory (INL) in the Hot Fuel Examination Facility (HFEF) inert atmosphere Main Cell. The monolithic fuel plates consist of a U10Mo fuel foil meat covered with a Zr diffusion barrier layer fabricated by co-rolling, and clad in 6061 Al using a hot isostatic press (HIP) bonding process. Initial unirradiated specimens exhibited nominal (fresh) fuel meat thickness ranging from 0.25 mm to 0.64 mm, and irradiated (as tested) fuel plate average burnup ranged from approximately 0.36 × 1021 to 6.2 × 1021 fissions/cm3. After sectioning individual specimens from the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre-dissolution measurement of width and post-dissolution measurement of thickness were conducted on test specimens to facilitate accurate analysis of bend test results. The final thickness values varied from 0.3 to 0.8 mm. Four-point bend testing was conducted on the HFEF Remote Load Frame (RLF) using custom-designed test fixtures and calibrated force and displacement sensors. All specimens exhibited dominantly elastic behavior and failed in a brittle manner. Analysis of test results showed decreasing elastic modulus and bending failure strength with increasing fission density. The trend of modulus decrease intersected with unirradiated modulus at zero fission density. The trend of decreasing bending strength with increased fission density is apparent, but has significantly higher uncertainty levels than the modulus results over the fission density range evaluated. A linear trend fit of the calculated bend strengths versus fission density does not coincide with unirradiated tensile strength, which is significantly higher. Porosities measured on the test specimen material are used with several model equations to predict irradiated modulus values. Over the range of available porosity information, comparison to the experimental modulus values are favorable. Other factors and their potential influence on modulus and strength calculations are discussed.

The Potential of Nuclear Reactor Technology to Treat Produced/Brackish Water for Oil & Gas Applications

The Potential of Nuclear Reactor Technology to Treat Produced/Brackish Water for Oil & Gas Applications

Scaling Trends of Electric Vehicle Performance: Driving Range, Fuel Economy, Peak Power Output, and Temperature Effect

This study investigated scaling trends of commercially available light-duty battery electric vehicles (BEVs) ranging from model year 2011 to 2018. The motivation of this study is to characterize the status of BEV technology with respect to BEV performance parameters to better understand the limitations and potentials of BEV. The raw data was extracted from three main sources: INL (Idaho National Laboratory) website, EPA (Environmental Protection Agency) Fuel Economy website, and the websites BEV manufacturers and internet in general. Excellent scaling trends were found between the EPA driving range per full charge of a battery and the battery capacity normalized by vehicle weight. In addition, a relatively strong correlation was found between EPA city fuel economy and vehicle curb weight, while a weak correlation was found between EPA highway fuel economy and vehicle curb weight. An inverse power correlation was found between 0–60 mph acceleration time and peak power output from battery divided by vehicle curb weight for 10 BEVs investigated at INL. Tests done on the environmentally controlled chamber chassis dynamometer at INL show that fuel economy drops by 19 ± 5% for the summer driving condition with air conditioner on and 47 ± 7% for the winter driving condition.

Analysis of Krypton-85 Legacy Waste Forms: Part II

During a removal of legacy materials from one hot cell at the Idaho National Laboratory in 2010, five metal capsules and some loose zeolite material were identified as krypton (Kr) immobilization test specimens produced in the late 1970s under the Airborne Waste Management Program (AWMP). This AWMP research and development effort examined the encapsulation of 85Kr within a collapsed zeolite structure for use as a potential waste form for long-term storage. The recovered capsules appeared to have been placed to the side and remained untouched after the AWMP was halted in the mid-1980s. These reclaimed capsules and loose material presented a unique opportunity to study a potential 85Kr waste form after three half-lives had elapsed. The first phase of this study included two parts: The first was to assess the physical condition of the capsule walls, and the second was to examine the Kr-containing material within the capsules. The first part of this study was previously reported and noted that substantial corrosion was observed throughout each capsule wall of the two previously breached capsules that were examined. One of these capsules had been hot isostatic pressed (HIPed) and one was not HIPed. The second part of the study examined the materials contained in the two previously breached capsules. There appears to be a relatively uniform distribution of Kr and rubidium throughout the pellets examined. The chemical composition of the pellets appears to be consistent with 5A molecular sieves. The material contained within the HIPed capsules showed ~1 at. % lead (Pb). The origin of the Pb is currently indeterminate. X-ray diffraction analysis shows a significant shift from the 5A structure, most likely due to the Kr encapsulation/sintering process that occurred when the samples were made. Calculations based on the energy dispersive spectrometry elemental analysis show a residual Kr level within the pellets that is within a factor of 2 of the reported Kr capacities for this type of processed material. This provides a clear indication that a significant fraction of the Kr initially encapsulated in the material remained within the waste form even following a significant breach of the capsule wall. As a result, it would appear that this Kr immobilization method, even in non-HIPed form, is very promising as a waste form for long-term storage. The successful analysis of these two breached capsules forms a solid basis for the future analysis of the remaining unbreached capsules, which offers the opportunity to provide an even more complete understanding of the long-term Kr retention performance of this promising waste form.

Response of Distributed Fiber Optic Temperature Sensors to High-Temperature Step Transients

Experiments were performed to test fiber optic temperature sensors under thermal conditions similar to those that would be experienced in the transient reactor test facility (TREAT) at the Idaho National Laboratory. Specifically, four single-mode optical fiber sensors were dropped into a pre-heated furnace to simulate a step temperature transient, for steps from room temperature to indicated furnace temperatures, ranging from 700 °C to 1100 °C. The four fibers were a standard SMF-28 fiber, a fiber with inscribed type-I fiber Bragg gratings (FBGs), and two fibers with inscribed type-II FBGs. The two type-II FBG fibers were different in the initial backscatter signal of the gratings; one fiber had gratings with greater backscatter reflection than the other. The type-I FBG sensor and standard SMF-28 sensor measured temperature for temperature transients with a final temperature of 700 °C. The low reflection type-II FBG sensor measured temperature transients with final temperatures up to 950 °C. The high-reflection type-II FBG sensor successfully measured temperature for short times with final temperatures up to 1050 °C. For long times at 1050 °C, there was a drift in the temperature sensing. We conclude that the high-reflection type-II FBG sensor would perform acceptably for application in the thermal environment of TREAT, or for other step temperature transients from room temperature to final temperatures up to 1000 °C.

Deuterium retention in neutron-irradiated single-crystal tungsten

Six single crystal tungsten specimens were neutron irradiated to a dose of 0.1 displacements per atom (dpa) at three different irradiation temperatures (633 K, 963 K, and 1073 K) at the High Flux Isotope Reactor in Oak Ridge National Laboratory under the US-Japan PHENIX project. A pair of neutron-irradiated tungsten specimens was exposed to deuterium (D) plasma to D ion fluence of 5.0 × 1025 m−2 at three different exposure temperatures (673 K, 873 K, and 973 K) at the Tritium Plasma Experiment in Idaho National Laboratory. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (Tritium Migration Analysis Program, TMAP) were used to understand D behavior in neutron-irradiated tungsten. A broad D desorption spectrum from the plasma-exposure temperature up to 1173 K was observed. Total D retention up to 1.9 × 1021 m−2 and near-surface D concentrations up to 1.7 × 10−3 D/W were experimentally measured from the 0.1 dpa neutron-irradiated single crystal tungsten. Trap density up to 2.0 × 10−3 Trap/W and detrapping energy ranging from 1.80 to 2.60 eV were obtained from the TMAP modeling.

Fuel – clad chemical interaction evaluation of the TREAT reactor conceptual low-enriched-uranium fuel element

The Transient Reactor Test (TREAT) facility resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The TREAT reactor is currently undergoing design and engineering studies for its conversion from a high enriched uranium (HEU) to a low-enriched uranium (LEU) core. The conceptual design of the LEU fuel element identified two main design differences compared with the HEU fuel element; namely, it will contain four times more fissionable material in its graphite matrix and distinct nuclear-grade Zirconium alloy, as Zircaloy-3 was used in the HEU fuel assembly and is not commercially available currently. These design changes may impact the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operation conditions and, therefore, was evaluated through a combination of experimental testing and thermodynamic modeling in order to determine implications for the fuel assembly. In this study, two potential cladding material types, Zircaloy-4 or Zr-1Nb alloys, were evaluated, and it was found for both material types that the extent of interaction and specific chemical reactions are minimal and no detrimental effect on the overall cladding properties is observed. The resulting interaction layer of 3–6 μm was measured after a 2-week exposure at 820 °C. The thermodynamic analysis was extended to temperatures beyond the TREAT reactor operation and accident conditions in order to give some insight that may be of interest for other reactor systems as the High Temperature Gas Reactors (operation above 1000 °C) and for Nuclear Reactor Severe Accident phenomenology study where the UO2 fuel could reach temperatures over 2800 °C and melt.

Carbon-14 content in tree and soil samples at the Idaho National Laboratory nuclear site

Idaho National Laboratory (INL) is a nuclear research facility located in southeastern Idaho, USA; over the course of its operational history, INL has operated 52 reactors and 1 reactor fuel reprocessing facility. To determine the extent to which previous nuclear operations at INL have impacted local environmental carbon-14 (14C) concentrations, tree and soil samples from the INL desert and surrounding areas were collected, combusted, and analyzed by Accelerator Mass Spectrometry (AMS). Transport models of the plumes from the Advanced Test Reactor Facility (ATR) and the Idaho Chemical Processing Plant (ICPP) suggest the historic annual integrated plume distributions from each of these source terms was most likely in the northeast-southwest direction, with very little ground contact from stack emissions in the immediate vicinity of the facilities and maximum estimated ground contact arising ∼400–1000 m to the northeast/southwest from each facility. 14C data from annual growth ring data from trees located immediately adjacent to the ATR, INL’s Central Facilities Area (CFA), and Mud Lake, Idaho (ML) are in statistical agreement with regional nuclear weapons testing fallout backgrounds. Surface soil samples taken near a low level radioactive disposal facility and downwind from the ICPP show percent modern carbon (pMC) values ranging from 28 ± 2 to 92 ± 4, suggesting a mixture of aged and modern background carbon containing materials. Taken together, these data suggest that 14C in the INL region is predominantly derived from a mixture of aged and modern background sources (e.g. nuclear weapons testing fallout), with insignificant contributions from INL source terms.

A new approach to probabilistic lava flow hazard assessments, applied to the Idaho National Laboratory, eastern Snake River Plain, Idaho, USA

A new approach to probabilistic lava flow hazard assessments, applied to the Idaho National Laboratory, eastern Snake River Plain, Idaho, USA

Establishing Outcrop Time Series at Various Depths in a Nonlinear Soil Column: An Iterative Approach

Seismic soil-structure interaction (SSI) analysis of nuclear facilities is an important consideration during design and retrofit. SSI tools used in the nuclear industry are currently based on an equivalent linear (EL) approach. Procedures for developing input ground motion for EL approaches are well established. However, the procedures for establishing input ground motion for nonlinear soil-structure interaction (NLSSI) analysis of nuclear facilities are not well established. A collaborative research group at Idaho National Laboratory has recently developed analytical methods and numerical tools for using NLSSI analysis for nuclear facility seismic calculations. NLSSI analysis for a nuclear facility allows for calculation of seismic wave motion through a near-field soil domain using either (a) vertically propagating shear and compressive waves, which is the current industry practice, or (b) a three-dimensional nonvertical wave field. This technical note presents an iterative procedure for establishing outcrop motion at a depth in the soil column for NLSSI analysis that uses vertically propagating shear waves. The approach presented in this technical note starts with a known ground motion at the surface that is deconvolved to a depth, and then the obtained motion is convolved up to a different desired location of input for the NLSSI model. To demonstrate the validity of the approach, a finite element soil column that is representative of a nuclear facility site in the United States is used to produce compatible outcrop seismic time series for reduced nonlinear soil mesh depths. The developed approach for reducing the nonlinear soil column model depth is a two-step iterative method. The first step is establishing an outcrop time series at the lowest depth considered that produces the top-of-soil response spectrum of an actual recorded ground motion. The second step is providing compatible outcrop time series at a shallower depth based on the information from the first step. A comparison of the 5% damped response spectrum from the resulting acceleration time series based on the iterated outcrop motions and the original acceleration time series is conducted. The study shows that the proposed iterative approach produced comparable results within 1% range of the original recorded time series results when sufficient iterations were performed.

Findings From an Operator-In-The-Loop Study on System Overview Displays in a Modernized Nuclear Power Plant

The second Operator Study of System Overviews (OSSO-2) was conducted with a three-person reactor operator crew in the Human Systems Simulation Laboratory (HSSL) at Idaho National Laboratory (INL) in August, 2017. The study supported control room modernization at a nuclear power plant and featured a benchmark comparison of three variants of a turbine control system (TCS): the existing analog TCS, a proposed standalone digital TCS with two displays, and the digital TCS with the addition of a third display depicting a system overview screen. TCS prototypes were developed at INL to allow evaluation of operator performance and preferences during realistic turbine scenarios in the full-scope simulator. The study revealed that completion of turbine startup was several minutes faster with the digital TCS variants than with the conventional analog TCS. Eye tracking revealed visual fixations were more widely distributed in the overview vs. standalone TCS condition, suggesting the overview screen may have cued reactor operators to verify values across the boards. Reviewing key plant parameters showed smoother transitions during load following for the digital vs. analog TCS. The study provides evidence that the new digital TCS could be used successfully by operators without extensive training or rewriting of the operating procedures, suggesting high usability for the digital TCS design. Further advantages were realized through the addition of the system overview screen to provide crews with at-a-glance indicators of key turbine parameters.

Improving bioenergy feedstock quality of high moisture short rotation woody crops using air classification

Short rotation woody crops have many advantages as perennial bioenergy feedstocks, including high biomass yields, high carbohydrate and low ash contents, and marginal land utilization. Through short rotation coppicing management, these biomass resources can be harvested year round. The challenge of year round harvesting is feedstock quality variability due to leaf content during periods of non-senescence. The low quality leaf fraction results in higher ash and moisture contents and lower carbohydrate content. Mechanical techniques, such as air classification, provide an economically feasible process to separate heterogeneous biomass samples based on particle density, size, and shape. In this work high moisture (>45%) hybrid poplar and shrub willow short rotation crops were air classified using a series of fan speeds for anatomical fractionation of the material. Air classification using an air velocity of ∼4.7 m/s removed a majority of the leaf material while retaining 88% and 87% of the hybrid poplar and shrub willow, respectively. At this velocity, the ash content was reduced from 2.34% to 1.67% for hybrid poplar and 2.60% to 2.14% for shrub willow. Concurrently, the carbohydrate content increased from 56.32% to 60.62% and from 54.03% to 55.99% for these same materials. As drying is a cost intensive step for processing high moisture biomass materials, the cost benefits (∼$3/Mg dry biomass) for removing low quality, high moisture materials prior to drying were also demonstrated.©Battelle Energy Alliance, LLC, contract manager for Idaho National Laboratory.

Solar Driven Thermo-Electrochemical Hybrid Sulfur Process for Hydrogen Production

A hydrogen economy requires primary sources to provide the power required to produce hydrogen from water, fossil fuels, biomass or similar compounds. Water-splitting processes are particularly attractive, since water is abundant around the planet, the unique products from water splitting are hydrogen and oxygen and a variety of primary energy sources can be used to produce the hydrogen. External electric power can be used to produce hydrogen by conventional water electrolysis [1], but it firstly requires the production of electricity to realize the electrochemical splitting of the water molecule. A more efficient and potentially more cost effective approach is to use thermal power to drive a thermochemical water splitting process. Such cycles break the hydrogen-oxygen bounds through a series of heat-driven chemical reactions, with recirculation of intermediate substances of different type. By this approach the external heat which drives the chemical reactions is converted into chemical energy of the hydrogen produced from water splitting. Among the various water splitting processes, the thermo-electrochemical Hybrid Sulfur (HyS) is one of the most appealing cycles. It has only fluid reactants and is comprised of only two global reaction steps: a low temperature electrochemical exothermic section, operating at temperatures on the order of 100 °C, and a high temperature thermal section operating at max temperatures of about 800 °C. In the electrochemical section SO2 and H2O are combined together to produce electrochemically H2 at the electrolyzer cathode and H2SO4 at the electrolyzer anode. Sulfuric acid is recycled inside the plant, concentrated and decomposed, in the high temperature thermal section, into SO2, O2 and H2O. Oxygen is then separated from SO2 and water and extracted as byproduct of the process. Solar driven HyS process is being studied by the Greenway Energy and the University of South Carolina within the DOE HydroGEN program (part of the DOE Energy Material Network initiative) [2], partnering with the Idaho National Laboratory, the Savannah River National Laboratory and the National Renewable Energy Laboratory. The project mostly focuses on the high temperature sulfuric acid decomposition section, developing new optimized catalyst formulations (i.e. achieving higher conversion, lower performance lifetime degradation and lower costs), to be integrated in novel solar reactor systems. A new bimetallic catalyst formulation is being developed and synthesized, with the catalyst deposited on TiO2 support. Compared with the current state of the art sulfuric acid decomposition catalysts, developed by Idaho National Laboratory during the DOE Nuclear Hydrogen Initiative [3], the proposed catalyst shows the potential for higher SO3 conversion yields and very limited sintering effects. Test results, under sulfuric acid environment and operating temperatures on the order of 800 °C, will be shown and discussed. The new catalyst will be placed in a novel solar reactor, being developed within the HydroGEN project. The proposed concept is based on a solar cavity receiver system, realizing the sulfuric acid decomposition and internal heat recovery in a single unit, directly heated by the incident solar radiation. This system, along with the actual H2SO4 decomposition into SO2, allows: (1) efficient internal heat recovery from the decomposition products and (2) connections with the metallic interfaced HyS equipment (e.g. tubing, valves etc) at low temperatures. Preliminary results from detailed transport model analysis will be shown, demonstrating effective decomposition and internal heat recovery. The integration of the high temperature decomposition section with the low temperature electrochemical section of the HyS process has also been studied from a process modeling point of view. A new SO2 electrolysis concept is proposed and discussed, with a vapor SO2 mixture feeding the electrolyzer. Preliminary results show the potential to achieve high efficiencies, with actual voltage values lower than the traditional target of 600 mV and current densities higher than the target of 500 mA/cm2 [4]. A new process flowsheet, initially developed within the current HydroGEN project, will be presented and discussed, achieving thermochemical efficiencies ≥ 35%.

Investigations of the Structural and Electrochemical Properties of Overheated Li-Ion Batteries and Its Effects in Single Cells Vs. Multi-Cells Packs

Abuse of electrochemical cells and batteries can have a wide range of effects depending on the properties of the battery as well as the nature of the damage caused. From a safety perspective, simple passive monitoring is often unable to identify the onset of failure until it is too late to intervene. Once a significant voltage change or self-heating of a cell is identified, a catastrophic runaway is often inevitable. The ultimate weakness of traditional monitoring techniques, typically based on voltage and temperature, is that they detect only the symptoms of cell failure. Once a significant change has been registered, irreversible internal changes have likely already occurred within the cell. Compounding the issue is the fact that cells in a battery are comprised of all the elements needed to release their stored energy, meaning that once a catastrophic failure has initiated no amount of intervention can keep the reaction from proceeding to its conclusion. The prevention of catastrophic failure requires detection of internal faults well before they have developed to the point of no return. This makes techniques that can monitor the internal health of a cell essential to provide dependable protection from catastrophic failure. Safety issues are exacerbated by the use of multi-cell modules. Monitoring a pack of multiple cells with traditional techniques can be misleading, a battery pack that otherwise appears healthy could have a single problematic cell within the pack. Abusive conditions may also be unintentionally created, as voltage imbalance can create overcharge/overdischarge conditions, while the poor thermal design of a module can generate thermally abusive conditions. However, monitoring the changes to the impedance would aid in identifying ailing cells before a dangerous and costly failure. This work examines the application of Electrochemical Impedance Spectroscopy (EIS) and Differential Capacity calculations (dQ/dV) as tools for determining the state of stability (SOS) of an electrochemical cell and multi-cell battery pack. The cells used for this study were commercial 10 Ah NMC cells subjected to thermal abuse coupled with EIS monitoring. This aims to not only provide a deeper understanding of how abused cells and batteries fail but also form the technical basis of a tool that could ultimately be used to interrogate cells of unknown stability and even monitor active cells for early signs of damage or failure. Fast impedance monitoring hardware previously developed at Idaho National Laboratory is used to provide not only monitoring after an abusive battery test but also look for changes in the cell while abusive conditions are applied. Differential capacity calculations are explored both before tests and after exposing single cells to different temperature ranges to examine any noticeable changes that may be monitored during charge and discharge operations. Lastly, X-ray diffraction measurements are performed on the cathode/anode before and after the abuse test to understand the effect of overheat in the structural changes. Figure 1 and Figure 2 presents the EIS and dQ/dV measurements for the NMC single cells heated from 50°C-150°C. Markers within the data can be identified and applied to cross-examine batteries of unknown stability as well as provide the basis for an active diagnostic method as part of a battery management system. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525. Figure 1

Numerical Modeling of a Novel Solar Driven Sulfuric Acid Decomposition Reactor

An efficient and cost-effective approach to produce hydrogen sees the direct use of thermal power to drive thermochemical water splitting processes. Such cycles break the hydrogen-oxygen bounds through a series of heat-driven chemical reactions, with recirculation of intermediate substances of different type. Among the various water splitting processes, the thermo-electrochemical Hybrid Sulfur (HyS) is one of the most appealing cycles. It has only fluid reactants and is comprised of only two global reaction steps: a low temperature electrochemical exothermic section, operating at temperatures on the order of 100 °C, and a high temperature endothermic thermal section operating at max temperatures of about 800 °C. In the electrochemical section SO2 and H2O are combined to produce electrochemically H2 at the electrolyzer cathode and H2SO4 at the electrolyzer anode. Sulfuric acid is recycled inside the plant, concentrated and decomposed, in the high temperature thermal section, into SO2, O2 and H2O. The endothermic decomposition of sulfuric acid takes place in two main reactions: H2SO4 → SO3 + H2O (1) SO3 → SO2 + 0.5 O2 (2) The required thermal power is provided by external source and by internal heat recovery from the SO2-O2 stream product of the decomposition reactions. Solar driven HyS process is being studied by the Greenway Energy and the University of South Carolina within the DOE HydroGEN program (part of the DOE Energy Material Network initiative) [1], partnering with the Idaho National Laboratory, the Savannah River National Laboratory and the National Renewable Energy Laboratory. The project focuses on the high temperature sulfuric acid decomposition section. One of the main tasks of the project is to develop and fabricate an effective and low cost high temperature solar reactor for the decomposition of the sulfuric acid into sulfur dioxide. Previous work was carried out within the DOE-NHI Initiative and resulted in identifying a reactor concept based on the bayonet heat exchanger as the baseline reactor configuration [2]. The bayonet reactor was tested both from a modeling point of view and from an experimental point of view, demonstrating the potential to achieve almost the reaction equilibrium conditions. At temperatures of about 800 °C and pressures of 14 bar, an SO2 production of approximately 19 mol fraction % was achieved, which is essentially the equilibrium value at the same operating conditions [3,2]. However, a solar driven bayonet concept presents several drawbacks, mainly relative to the need for an additional intermediate heat exchanger to transfer the solar power to the reactive sulfur mixture. The removal of the intermediate solar heat exchanger will result in increased energetic and exergetic efficiencies and decreased capital and operating costs. A novel reactor concept is proposed, described and discussed, based on a direct solar heated cavity receiver system. It realizes the solar driven sulfuric acid decomposition and internal heat recovery in a single unit, directly heated by the incident solar radiation without the presence of any intermediate heat transfer fluid. The reactor includes catalytic structures to decompose H2SO4 and foamed structures in the internal heat recovery channels to increase the overall thermal conductivity. A reactor unit, made of SiC material, and directly heated from solar radiation has preliminarily been demonstrated from a modeling point of view. A detailed transport phenomena model was developed using Finite-Volume approach and run in STAR-CCM+. The kinetics parameters reported in Reference [3] were adopted for the current simulations. Results showed excellent conversion rates and reduced temperature gradients and pressure drops inside the device, achieving better performance than the bayonet configuration. The cavity receiver-reactor also allows: (1) efficient internal heat recovery from the decomposition products and (2) connections with the metallic interfaced HyS equipment (e.g. tubing, valves, etc.) at low temperatures. Results obtained from a first conceptual scale up analysis of the reactor concept also demonstrated the actual feasibility for integration in both centralized and distributed solar tower based plants.

Postirradiation examination on metallic fuel in the AFC-2 irradiation test series

Irradiation experiments on metallic fuel alloys containing rare earth additions to simulate potential carry-over from pyro-metallurgical reprocessing and minor actinide additions were performed in the Idaho National Laboratory Advanced Test Reactor. These tests were designated AFC-2A and AFC-2B. Additionally irradiation experiments were also performed on metallic fuel alloys containing minor actinide additions that were fabricated by different techniques. These tests were designated AFC-2E. These pins were removed from the reactor after achieving burnups between 6 and 11% fission per initial heave metal atom. Examination of these pins revealed that they were unintentionally operated beyond their desired power and temperature conditions. Highlights from the postirradiation examination of these experiments is discussed. A variety of issue contributed to the poor performance of these irradiation tests including capsule design, reactor power variations and fabrication issues. While all the issues contributed, the major issue was determined to be unanticipated variations in reactor power. Through comparisons between these experiments and historical off-normal experiments, it was possible to conclude bound the operating temperatures of these experiments. There likely was some in-pile melting in the rare earth addition experiments (AFC-2A, AFC-2B), but it is not likely that there was in-pile melting in the fabrication technique experiments (AFC-2E).

The strategic value of human factors engineering in control room modernization

Control room modernization is one of the most challenging and complex upgrade projects that a nuclear power plant can undertake. It can have almost as big an impact on operations as, for example, turbine replacement. The challenges of migrating an analog control system to a distributed control system are already well known and a number of nuclear utilities have embarked upon various levels of effort to upgrade some of the systems in the control room. When planning for control room upgrades, plants have to deal with a multitude of engineering, operational, and regulatory impacts. This will inevitably include several human factors considerations, such as workstation ergonomics, viewing angles, lighting, seating, new interaction modalities, new communication requirements, and new concepts of operation. In helping nuclear power utilities to deal with these challenges, the United States Department of Energy researchers located at Idaho National Laboratory (INL) have developed research-based human factors design and evaluation methods to be used in the development of end-state concepts for modernized control rooms and to manage the various phases of the upgrade life cycle. The methodology includes interactive sessions with operators in INL's Human System Simulation Laboratory, three-dimensional modeling to visualize control board changes and operator-system interaction, and development of human-system prototypes to evaluate various aspects of proposed modifications. This methodology has been applied at a number of U.S. nuclear power plants where modernization projects are underway, including Exelon's Braidwood and Byron plants, and Arizona Public Service's Palo Verde plant. It was demonstrated that including this methodology in the plant's engineering process helps to ensure an integrated and cohesive outcome that is consistent with human factors engineering principles and provide substantial improvement in operator performance.

Digital neutron and gamma-ray radiography in high radiation environments with an MCP/Timepix detector

The radiation levels from used nuclear fuel make their post irradiation examination very difficult. Handling material removed from assemblies is almost exclusively performed inside shielded hot cells. Neutron radiography provides a non-destructive option for examining the internal structure, location and integrity of irradiated fuel. This paper addresses the performance and stability of a neutron imaging detector comprised of microchannel plates (MCPs) and a Timepix readout in gamma and neutron radiation environments. The detector is being considered as an alternative to the currently used activation foils and photographic film measurement techniques. The MCP/Timepix detector was used in neutron radiography experiment on a beam line at the Neutron Radiography Reactor (NRAD) at Idaho National Laboratory (INL). The detector was installed ∼17.5 m from the reactor core with a line of sight view to the core, which results in a fast and epithermal neutron spectrum as well as high intensity gamma radiation. The gamma-ray dose rate in the beam line was 812 mSv/hr. Complementary measurements were made using 137Cs and 60Co isotopic sources at gamma dose rates in excess of 6 Sv/hr. The detector functioned normally over several hours in this environment. These conditions are comparable to planned radiography of irradiated nuclear fuel assemblies. Experimental results using ASTM sensitivity and beam purity indicators suggest that a spatial resolution of <100 μm may be possible in neutron radiographic measurements using this detector for studies of irradiated nuclear fuel. Potential radiographic and tomographic applications are discussed. The results of present experiments are not limited to the studies of irradiated fuel assemblies, but rather are relevant to any experiment where imaging needs to be performed in a high radiation environment.

Fast-neutron multiplicity counter for active measurements of uranium oxide certified material

We developed a fast-neutron multiplicity counter (FNMC) based on stilbene and EJ-309 organic scintillators. The system can detect and discriminate correlated photon and neutron multiplets emitted by fission reactions. We used the system to estimate the fissile mass of uranium oxide samples in active interrogation mode at the Zero Power Physics Reactor of Idaho National Laboratory (INL). Two sets of certified reference material (CRM) samples were characterized. The U-235 enrichment of the first set is constant at 93.2 wt%, and the UO2 mass ranges between 0.5 and 4 kg. The second set includes samples of increasing enrichment (from 20 wt% to 97 wt%) and constant UO2 mass of 230 g. We used two AmLi sources to induce fission reactions in the samples. Despite the intense gamma-ray background of the UO2 and interrogating sources, the system could measure induced fission neutrons emerging from the interrogated samples without additional shielding surrounding the sample and only relying on pulse shape discrimination to classify neutron and gamma-ray pulses. The overall neutron count rate and time-correlated counts are well correlated with the sample fissile mass. We also proved that CRM samples can be used to build a calibration curve to assay the U-235 mass of unknown samples of different mass, geometry and enrichment, with an average bias error of 8%, for U-235 mass higher than 390 g.

Investigation and Implementation of Commercially Available Optically Stimulated Luminescence Dosimeters for Use in Fixed Nuclear Accident Dosimeter Systems.

Idaho National Laboratory transitioned from an external dosimetry system reliant on thermoluminescent dosimeters to one that uses optically stimulated luminescence dosimeters in 2010. This change not only affected the dosimeters worn by personnel, but those found in the nuclear-accident dosimeters used across Idaho National Laboratory. The elimination of on-site use and processing of thermoluminescent dosimeters impacted Idaho National Laboratory's ability to process nuclear-accident dosimeters in a timely manner. This change in processes drove Idaho National Laboratory to develop an alternative method for fixed nuclear-accident dosimeter gamma-dose analyses. This new method was driven by the need to establish a simple, cost-effective, and rapid-turnaround alternative to the thermoluminescent-dosimeter-based fixed nuclear-accident dosimeter system. An adaptation of existing technologies proved to be the most efficient path to this end. The purpose of this article is to delineate the technical basis for replacing the thermoluminescent dosimeter contained within the Idaho National Laboratory fixed nuclear-accident dosimeter system with optically stimulated luminescence-based Landauer, Inc., nanoDot dosimeters.