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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Slip Tendency Analysis of Fracture Networks to Determine Suitability of Candidate Testbeds for the EGS Collab Hydroshear Experiment

Experiment 2 of the EGS Collab project is aimed at testing stimulation by hydro-shearing of existing natural fractures, versus Experiment 1, which is focused on hydraulic fracturing a rock mass to enhance permeability. The main criterion for the testbed selection in Experiment 2 is the presence of an interconnected network of fractures, at least 10 meters in extent, and with orientations such that shear slip can be induced at injection pressures less than the minimum horizontal stress (Shmin). The feasibility analysis for this experiment requires a well-constrained stress state along with well-characterized fracture networks. The fracture systems at two candidate locations in the Sanford Underground Research Facility (SURF), the 4850 Level (number refers to depth below ground surface in feet) and the 4100 Level, have been characterized to different extents. The 4850 Level has a well-characterized Discrete Fracture Network (DFN) from borehole (drilled for an experimental test bed) and drift observations. In contrast, the 4100 level has a fracture network characterized only by observations from the drift wall, which provides little constraint on fracture extents. This paper will present assessments for the interpreted slip potential for the natural fractures at the two locations. Data uncertainties are addressed by performing a probabilistic analysis that takes into account the uncertainty in the stress state, uncertainty in the fracture properties, and preliminary borehole locations. There is strong evidence of shear stimulation of some natural fractures in Experiment 1 on the 4850 level. We use examples of natural fractures strongly linked to shear stimulation in Experiment 1 from multiple monitoring indicators to test the consistency of the stress model and guide the uncertainty interpretation. This analysis is intended to guide the site selection process for Experiment 2 by highlighting the fracture orientations that are likely to be shear-stimulated in a majority of modeled realizations.

40 EE - Geothermal Technologies Office (EE-4G)↗
DOE OSTI2021

MULTIPHYSICS-MODELING OF FIRE-INDUCED URANIUM AEROSOL FORMATION – A-POSTERIORI BENCHMARKING OF EXPERIMENTS

The formulation and solution of a model that properly described the temperature profiles of uranium in fire conditions was prepared, verified and validated, to support a-posteriori benchmarking of historical experiments. Data from the multi-physics model combined with data regression from the experiments provides a useful tool for exploring sources of potential bias in historical experiments. The approaches used are augmented by visual observations and photographic evidence allowing for correlations between various phenomena, independent of whether or not that specific phenomena are part of the validated model. In this manner, the validated model is a tool for a-posteriori benchmarking, and not a model designed to replicate the experimental output exactly for any specific experiment. For experiments like the Clark (2015) experiments, this integrated model (including chemistry) demonstrated that the largest bias was associated in the metallurgy of the coupons. The model developed does serve as a basis for defending which of the historical alloys tested are the most appropriate for any other alloy being considered (e.g., one that has not yet been tested). The model identifies key reactions that impact uranium chemistry and can significantly bias experimental results. Through use of the model and energy balances, data from historical experiments by Elder and Tinkle (1980) were regressed and benchmarked. This study identifies that all uranium alloys are not equal and that there are distinct differences between the dominant alloys representing distinct metallurgical phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermo-Hydraulic-Mechanical Modelling of the Full-Scale Emplacement (FE) Experiment at the Mont Terri Rock Laboratory - 16490

The Full-Scale Emplacement Experiment at the Mont Terri underground research laboratory is a heater test carried out in a gallery constructed in Opalinus Clay. The aim of the experiment is both to develop emplacement procedures for radioactive waste canisters and their surrounding backfill as well as understanding the thermo-hydro-mechanical effects of disposing of radioactive waste on the engineering materials and the host rock. Coupled thermo-hydro-mechanical models have been used to aid planning of the locations of sensors, to inform decisions on the schedule for turning on the heaters as well as to make predictions of the temperatures and water pressures in the bentonite and the Opalinus Clay. The FE Experiment was preceded by a number of other experiments at the Mont Terri laboratory at smaller scales, but with similar materials, which provided crucial input to the models and gave confidence in the general process understanding. Blind predictions were made prior to turning on the heaters, of the temperature and relative humidity in the bentonite surrounding the heaters. The comparison between the blind predictions and early data from the experiment show that knowledge gained from smaller scale experiments on similar materials has enabled good blind predictions to be made and gives confidence that temperature in the experiment is unlikely to rise to a level that could damage the sensors. It also demonstrates that the thermal properties of the disposal system are understood well enough to produce reliable estimates of temperature in the backfill. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Artificial Neural Network Representation of Dynamic Criticality Excursion Experiment Data

This paper outlines time dependent considerations in applying the artificial neural network (ANN) paradigm to dynamic nuclear criticality excursion experiment data. The ANN amongst a variety of uses, can apply to function fitting of multiple input-multiple output data sets, pattern recognition, cluster analysis, and time series prediction. The ANN is a component of Machine Learning (ML) data analysis. The practical application of ML to criticality safety has been funded through the Y-12 Plant Directed Research and Demonstration (PDRD) program to support criticality safety and emergency planning efforts. The initial effort considered function fitting of criticality excursion data. Recent efforts successfully extended the ANN to criticality safety code validation. The initial ANN work applied to criticality excursions centered around multivariate function fitting of static (tabular) data taken from the French Consequences Radiologiques d'un Accident de Criticite (CRAC) and SILENE reactor campaigns. The former report (LeCorche) summarized important tabular data for the first 44 CRAC experiments. The initial ANN was trained to specific time-independent inputs, then applied to predict peak fission rate and total fissions for a wide variety of fissile solutions process conditions. This work extends that effort to apply the ANN paradigm to dynamic prediction of time-dependent criticality excursion experiment data. A dynamic criticality excursion in fissile solution represents a complex multiphysics phenomenon. Factors include among other things, coupled neutronic, multiphase fluid flow, temperature, pressure, acoustic shock, boiling heat transfer, reactivity feedback, and radiolytic gas considerations. There have been many efforts to model (over time) as coupled ordinary and (space) as partial differential equations (ODE, PDE). The ANN representation noted here does not rely on solving explicit ODE or PDEs. Rather the intent is to demonstrate that as a time series, the criticality excursion can be predicted by training specific ANN architectures to experimental input and output time series exclusively. The rationale for extending this work to the time-dependent regime is important in developing a 'criticality excursion emulator' solely on the basis of learning characteristics from experiment dynamic data. An emulator is different than a computer code simulator driven by the solution of ODEs and PDEs given select 'tuning parameters' that match outputs to experiments. A criticality excursion emulator, derived exclusively from an ANN or ML, could predict most likely excursion characteristics, trained from historic dynamic experiment data, presented with input values within the range of those experiments. The work noted shows that generic time-series ANN architectures can be satisfactorily applied to the prediction of dynamic, extremely chaotic criticality excursion power given an initial time dependent input series. In reality, an excursion emulator could be driven by user-supplied initial (t=0), time independent or constant values to generate time dependent outputs. Thus ongoing work includes developing a custom ANN such that initial dynamic conditions would be predicted. Those results then serve as inputs to a customized ANN architecture currently in development. (author)

42 ENGINEERING↗
DOE OSTI2016

Development of Critical Experiments to Benchmark Moderator Temperature Reactivity Worth

The simulation of nuclear reactors requires being able to accurately model several different interrelated physical processes including neutronics, heat production, fluid/thermal physics, structural mechanics, fuel behavior, chemistry, and balance-of-plant. In order to design and use new types of nuclear reactors, modern codes have been developed which couple together the behavior of these different physical models to accurately consider the feedback effects between different physical effects. The SHARP toolset, part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project, is an example of an effort to combine models of neutronics, thermal-hydraulics, and structural mechanics in a coupled simulation to model any type of nuclear reactor as accurately as possible. The validation of multi-physics coupling techniques requires new experimental benchmark experiments specifically designed for that purpose. As a preliminary step towards design of these experiments, a series of measurements have been performed to demonstrate the temperature-dependent neutronic behavior of a well-characterized reactor assembly for comparison with the available codes. The facility adopted for these experiments is Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI). The RCF is unique in several aspects and provides several advantages to this type of experiment. Built in 1954 by the American Locomotive Company (ALCO), the RCF has been used to design small compact reactors. RPI acquired the facility in 1963, and the highly-enriched uranium fuel was later swapped for low-enriched fuel pins that were excess from the SPERT facility experiments. Nowadays, The RCF is regarded as a teaching and research facility, permitting the training of licensed Senior Reactor Operators, and offering a critical experiments laboratory course to students. The reactor core standard configuration consists of 332 or 333 UO{sub 2} fuel pins, containing a total of 35.2 grams of U- 235 each (4.81 % enriched). The pins are arranged in a regular lattice with a pitch of 1.6256 cm. The number and configuration of these pins are flexible, and the reactor's overall thermal power is limited to no more than 15 W. As the low thermal output of the reactor does not induce any measurable temperature changes, the effect of temperature in the reactor is observed by applying two 18- kW electric heaters to the water moderator, with mechanical agitators to maintain uniformity in moderator temperature. Before beginning experiments, each fuel pin inside the core was surveyed for mechanical defects (bending or warping) that may contribute to uncertainty in core behavior. Every fuel pin has been individually removed from the core and checked for any irregularities on its surface (caused by shocks or usage). About 12 % of the 333 pins were found to have a small degree of warping and were replaced with excess pins available in the facility. In doing this, we made sure all the pins are identical, thereby increasing the accuracy of future models. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A New Look at Historic CRAC 'Divergence' Experiments - Implications for a CAAS Minimum Accident of Concern

This paper examines power history parameters and results based on historic fissile solution criticality 'divergence' experiments (referenced here as Divergence Experiments) from the 1960's-70's' Consequences Radiologiques d'un Accident de Criticite' (CRAC) campaign. The results from specific Divergence Experiments have implications for an ANSI/ANS-8.3 (1997) Criticality Accident Alarm System (CAAS) Minimum Accident of Concern (MAC). Specifically, this paper examines select CRAC Divergence Experiments in conjunction with 'qualifying reactivity additions' for a MAC defined by an absolute time constraint. A previous paper presented at the 2011 International Conference on Nuclear Criticality (ICNC 2011) illustrated the differences in 'qualifying reactivity additions' above delayed critical (and below prompt) for a variety of MAC constructs. That paper was the first to examine delayed critical kinetics in greater detail over a range of process and MAC considerations. The ICNC2011 work did not consider an absolute MAC time constraint from the excursion onset and assumed an initial excursion power. That study evaluated MAC constructs based on 'relative time differences' between reaching a specific fission rate and total fissions. The work presented here thus picks up where the ICNC2011 study left off. The primary focus of this work provides simulated power histories for specific Divergence Experiments that served as a basis for justifying a MAC noted in ANS-8.3 Appendix A. This paper documents computational modeling and dynamic multiphysics simulation with known Divergence Experiment parameters for reactivity additions at, below, and above the reported terminal reactivity. The initial step considers replicating what could be a reasonable power history for the terminal reactivity addition to the extent the same peak power and total fissions are obtained over the total time considered. Power histories for all other reactivity additions are simulated by varying the input reactivity and solution heights for the same model. A relationship is obtained between reactivity addition and the time to reach specific MAC power and fission levels within a given time constraint (e.g. 60 s). The 'qualifying reactivity addition' is determined for the total MAC fissions within that time. This value is then compared to the historic terminal value (e.g. $0.33) or any other values (e.g. 'excess reactivity of only a few cents') noted in ANS-8.3 Appendix A. In general, calculated MAC power levels have ranged from 1.8 kW to 173 kW depending on material type and form, H/U, enrichment, critical volume, and dose response function. A minimum accident consideration noted a SHEBA experiment with an excess reactivity 7 cents above delayed critical (TD ∼110 s). This excursion provided an additional 10{sup 15} fissions after an undetermined buildup time resulting in a 'few times 10{sup 15} fissions'. ICNC 2011 LEU solutions calculations show 0.2 Gy in air corresponds to 1.4 x10{sup 16} fissions. Thus, while historically, 0.2 Gy has been loosely associated with 'several times 10{sup 15} fissions', in reality, the actual total fissions could be an order of magnitude larger, and the power associated with 0.2 Gy/min (7.5 kW) larger as well. The reactivity addition required to obtain the actual 0.2 Gy total fissions over 60 s could be much larger than considered. Another reason the qualifying reactivity level could be larger than previously noted, involves temperature reactivity feedback. Compensated reactivity results in a departure from a pure exponential power rise, increasing the doubling time (TD) as the excursion approaches peak power. Uncompensated reactivity additions and constant TD have been traditionally assumed for a pure exponential power rise through the kW range to determine either total MAC fissions, power, or arrival time to these values.

42 ENGINEERING↗
DOE OSTI2015

The experiment filtering tool in the MACSENS workbench - 14341

The MACSENS (outils Moret d'Aide a la Caracterisation et aux etudes de SENSibilite) workbench is dedicated to the advanced validation work for criticality studies. The aim of this kind of tool is to determine the experiments from a validation database being the 'closest' to an application case and, from there, to evaluate a calculation bias with a related uncertainty. To achieve this goal, the MACSENS workbench relies on different kind of information stored in various databases depending on the origin of the data. Two main kinds of data can be identified. The first one concerns the data used to describe the experiments of the validation database (e.g. the inputs needed to perform a calculation). These values characterize the chemical and geometrical models. They can be considered as constant, as they do not vary once evaluated, and some of them can be from a non-numerical nature (reflector material for example...). The second kind of data gathers the results obtained with the use of a simulation code, such as reaction rates, fluxes and leakages, sensitivities, k{sub eff}, etc. These values depend on the version of the code and are purely numerical. In the early 2000's, regarding the increasing number of experiments evaluated in the ICSBEP handbook, the OECD/NEA (Organization for Economic Co-operation and Development/Nuclear Energy Agency), under the ICSBEP (International Criticality Safety Benchmark Evaluation Project) working group direction, decided to store such information in a database and to provide a software, called DICE (Database for the International Criticality safety benchmark Evaluation project), allowing users to search easily criticality experiments of interest. Up to now, more than 4400 critical or subcritical configurations have been compiled by ICSBEP, described and stored in the DICE database. Over time, a validation work has been done on the data, with the participation of IRSN, and a quality assurance policy has been defined. For these reasons, the DICE database is highly reliable. These features were fundamental for the MACSENS developers, who were looking for a database containing descriptive data of the experiments included in the MORET code validation database. The paper describes the solution chosen by IRSN to take advantage of the DICE database content and to overcome the DICE limits. It also explains the quality assurance policy applied to guaranty the reliability of the stored data. It will also discuss the use of the experiment-filtering module of the MACSENS workbench to select a set of experiments and to examine numerical parameter trends for an application case. (authors)

42 ENGINEERING↗
DOE OSTI2015

2013 Caliban and Prospero experiments for the study of the behavior of radiation protection instrumentations in case of a criticality accident - 14346

Experiments using the CALIBAN and PROSPERO reactors were performed in 2013 in the framework of a collaboration between IRSN and CEA. The objective was to provide data about the possibility to use information given by radiation protection instrumentations (RPI) in the case of a criticality accident. In that purpose, the behavior of RPI was studied for various kinetics of the reactors which simulated different starts of a criticality accident. The article presents the safety issues associated with these experiments, i.e. the possible use of information given by the RPI for the emergency response of a criticality accident (detection of the accident, triage of the personnel, follow-up of the accident and reentry). Then, one paragraph is dedicated to the presentation of the tested material. In addition, the CALIBAN and PROSPERO reactors are briefly described, followed by the performed experiments. The RPI are neither designed nor dedicated to be used in case of a criticality accident. However, the preliminary conclusion of the experiments performed in 2013 is that the tested RPI have a correct behavior for slow kinetic experiments performed with the PROSPERO reactor, but might give less reliable and accurate information for pulse experiments performed with the CALIBAN reactor. In the last case, they can nevertheless provide information that can be helpful for the emergency response and the limitation of the consequences of the criticality accident. However, if the RPI are included in the emergency response of a criticality accident (detection, triage, follow-up, reentry), emergency procedures should be adapted. Other experiments might be performed in order to confirm results and to extend the number of tested materials, in particular portable survey meters or contamination area monitors. (authors)

42 ENGINEERING↗
DOE OSTI2015

A new critical experiment in support of the nuclear criticality safety class - 14488

A new critical experiment has been performed on the Planet vertical assembly machine at the National Criticality Experiments Research Center (NCERC). The purpose of the experiment was to compare the moderator properties of polyethylene versus Lucite and to use the experiment as an instructional tool for criticality safety engineers. The Lucite version of this experiment has been performed many times since the mid-1970's. Similar to the Lucite critical experiment, the polyethylene experiment was designed to emphasize the factors that affect criticality; namely, enrichment, reflection, interaction, moderation, mass, shape, volume, density, concentration, and poisons. The critical mass obtained with the moderating high density polyethylene plates was 997.0 ± 0.65 grams of highly enriched uranium, which is approximately 34% lower than the critical mass obtained using the Lucite moderating plates. This 34% decrease is mostly attributable to the higher hydrogen and carbon atom densities of polyethylene versus Lucite. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Determination of critical experiment correlations using the sampler sequence within scale 6.2 - 14602

The validation of neutron transport methods used in nuclear criticality safety analyses is required by consensus American National Standards Institute/American Nuclear Society (ANSI/ANS) standards. In the last decade, there has been an increased interest in correlations among critical experiments used in validation that have shared physical attributes and which impact the independence of each measurement. The statistical methods included in many of the frequently cited guidance documents on performing validation calculations incorporate the assumption that all individual measurements are independent, so little guidance is available to practitioners on the topic. Typical guidance includes recommendations to select experiments from multiple facilities and experiment series in an attempt to minimize the impact of correlations or common-cause errors in experiments. Recent efforts have been made both to determine the magnitude of such correlations between experiments and to develop and apply methods for adjusting the bias and bias uncertainty to account for the correlations. This paper describes recent work performed at Oak Ridge National Laboratory using the Sampler sequence from the SCALE code system to develop experimental correlations using a Monte Carlo sampling technique. Sampler will be available for the first time with the release of SCALE 6.2, and a brief introduction to the methods used to calculate experiment correlations within this new sequence is presented in this paper. Techniques to utilize these correlations in the establishment of upper subcritical limits are the subject of a companion paper and will not be discussed here. Example experimental uncertainties and correlation coefficients are presented for a variety of low-enriched uranium water-moderated lattice experiments selected for use in a benchmark exercise by the Working Party on Nuclear Criticality Safety Subgroup on Uncertainty Analysis in Criticality Safety Analyses. The results include studies on the effect of fuel rod pitch on the correlations, and some observations are also made regarding difficulties in determining experimental correlations using the Monte Carlo sampling technique. (authors)

42 ENGINEERING↗
DOE OSTI2022

Development of the TREAT M2 experiment as a transient benchmark

A transient benchmark based on the Transient Reactor Test Facility (TREAT) M2 Calibration experiment (M2-CAL) is under development. TREAT, at Idaho National Laboratory, is a graphite moderated air-cooled research reactor which has been used extensively for fuel material testing under extreme and accident conditions. Accurate benchmark models are a beneficial component in the operations, experimental planning, and development of TREAT. In this work, we present a transient benchmark model for the M2-CAL experiment core loading. The benchmark model incorporates the coupling between the Monte Carlo Code SERPENT and the computational fluid dynamics code OpenFOAM to capture the temperature feedback mechanism. The M2-CAL transient 2580 was simulated in this work. A pre-transient analysis was performed to determine the optimum core composition and conditions before the beginning of the transient. The analysis was validated against historic TREAT kinetic measurements and the worth of the transient rod T-2. The axial power distribution in the flux wire for the M2-CAL experiment was determined and contrasted with the experiment. The model was then used to simulate the M2-CAL transient 2580 experiment based on the reported pre-transient and transient conditions. Several transient observables were calculated and compared to the experiment. The model shows good agreement with the experimental power traces, period, and average increase of power at the power ramp (time > 7.8 s). Inverse point kinetics analysis was performed during the period of the power ramp for the model and the experiment based on that, the temperature feedback component was isolated and contrasted with the experimental feedback. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Searching for Sterile Neutrinos In the NOvA Experiment and Measurements of Hadronic Energy Response with NOvA Test Beam Data

The NOvA (NuMI Off-Axis electron neutrino Appearance) experiment is a long-baseline neutrino oscillation experiment composed of two functionally identical detectors: a 300-ton Near Detector and a 14-kton Far Detector, separated by 809 km and placed 14 mrad off the axis of the NuMI neutrino beam created at Fermilab. This configuration enables NOvA’s rich neutrino physics program, which includes measuring neutrino mixing parameters, determining the neutrino mass hierarchy, probing CP violation in the leptonic sector, studying neutrino cross sections, searching for sterile neutrinos, and more.\\ In this thesis I will present two independent analyses developed within the NOvA experiment. The first is a search for light sterile neutrino oscillations in the 3+1 framework using $\nu_\mu$-CC and NC samples from the NOvA Near and Far Detectors. Light sterile neutrinos are hypothetical neutral leptons that do not participate in the weak interactions but can mix with the three known active neutrinos: $\nu_e$, $\nu_\mu$, and $\nu_\tau$. Anomalous results, such as $\nu_e$/$\bar{\nu}e$ appearance in a $\nu_\mu$ ($\bar{\nu}_\mu$) beam observed by the MiniBooNE and LSND experiments, can be explained by the existence of sterile neutrinos. Thus, proving the existence of this type of neutrino is essential. The latest results of these searches in the NOvA experiment, along with contributions to this effort, will be presented.\\ The second is the first measurement of the pion energy response using data from the NOvA Test Beam experiment. The NOvA Test Beam experiment, deployed at Fermilab, uses a scaled-down 30-ton NOvA detector to analyze tagged beamline particles. The beamline can select and identify electrons, muons, pions, kaons, and protons with momenta ranging from 0.4 to 1.4 GeV/c. Pions are an important component of the hadronic system in neutrino interactions, and understanding how the detector responds to these particles is crucial for the validation of the simulation and the reduction of detector calibration uncertainties, which remain one of the largest systematic uncertainties in NOvA analyses.

Dueñas Tonguino, David Francisco [Cincinnati U.] (↗
DOE OSTI2025

Genetic algorithm optimization of nuclear criticality experiment for reduction of intermediate-energy 239 Pu nuclear data uncertainties

Nuclear criticality experiments are conducted to investigate specific nuclear data important for safe handling and storage of fissile materials, reactor design and operation, and the validation of radiation transport codes. Incorrect or uncertain nuclear data can prohibitively impact operational safety limits, reactor licensing, and predictive simulation capability; therefore, integral measurements from criticality experiments are necessary and should be performed frequently. To maximize the impact of the integral measurements, it is important to consider experiment geometry, material selection, and component dimensions. When taking these considerations into account, the experiment design process becomes iterative and very time intensive. This work utilizes a genetic algorithm to efficiently explore potential nuclear criticality experiment designs for the Laboratory Directed Research & Development project PARADIGM (PARallel Approach of Differential and InteGral Measurements) at Los Alamos National Laboratory. In this paper, the building blocks of the genetic algorithm are discussed in detail, the genetic algorithm methodology is verified, and the genetic algorithm is used to produce three candidate experiment models for the final PARADIGM design. The three candidate models produced by the genetic algorithm consist of copper-reflected assemblies containing 14 repeating units of alumina, graphite, boron, and plutonium plates. Furthermore, in addition to the optimization results, final design considerations are also discussed for designs with a height and/or weight very close to or slightly above assembly machine operational limits.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Effect of Nuclear Data Covariances on Integral Experiment Design with Sensitivity and Uncertainty Analysis

Washington River Protection Solutions (WRPS) uses MCNP6.2 and the Whisper code for criticality safety analyses of the Hanford Tank Farm. Together the codes derive baseline upper subcritical limits (USLs) for the waste models using experimental benchmarks. Whisper returns higher USLs, i.e. , has less of a conservative penalty, when the neutronic similarity of the experimental benchmarks to the application is high. Unfortunately, few critical benchmarks have high similarity to the Hanford tanks. The waste in the tanks is highly dilute in plutonium and contains large masses of weakly neutron-absorbing elements like iron and manganese. Experimental benchmarks typically have low sensitivity to these absorbers because they are present as structural materials. Lacking similar benchmarks, new Thermal Epithermal eXperiment (TEX) configurations with high Pu content and interstitial iron absorbers have been designed for the criticality safety validation. The features of the design have been iterated upon to maximize the similarity between the experiment and different Hanford waste models. The similarity is quantified with sensitivity analysis and uncertainty quantification using the representativity coefficient, or c k . The representativity calculation requires nuclear data covariances, which may differ between nuclear data libraries and between library versions. Because of these variations, the optimal design may depend on the nuclear data covariances library. A scenario can be envisioned where an experiment is designed, and c k is maximized, with one set of covariance data. However, when the covariance data is changed, say from ENDF/B-VII.1 to ENDF/B-VIII.0, and the benchmark is used in a criticality safety evaluation, the experiment becomes suboptimal with respect to c k . In this paper, we present how the optimal design of the new TEX experiments varied depending on the nuclear data covariances used to calculate c k . We compare ENDF/B-VII.1 and ENDF/B-VIII.0, as if the library had been updated since the design of the experiment. Additionally, we use JEFF3.3 to simulate if the covariance data of a different library had been used. The results show that the covariances do have an important effect on the designs, less so for thermal systems (where the data are more consistent between evaluations) and more so for epithermal systems where more differences exist.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252 Cf production targets, 75 Se production capsules, 63 Ni production capsules, a 188 W production capsule, and various materials irradiation targets. Fully loaded 238 Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

TEX-HEU: Integral Experiments with Highly Enriched Uranium and Polyethylene [Slides]

This presentation discusses Thermal/Epithermal eXperiments (TEX) in relation to integral experiments spanning a wide range of fission energy spectra utilizing existing NCSP fissile material assets, TEX-Pu for plutonium, and TEX-HEU highly enriched uranium (HEU) for highly enriched uranium. It also discusses the TEX-HEU design, the use of HEU fuel, and polyethylene parts. The experiment, executed in the Nuclear Criticality Experiment Research Center of the Nevada National Security Site is discussed, specifically how it was assembled on the Comet Universal Critical Assembly Machine. Measured quantities and various critical configurations are listed. In conclusion, the presentation states that they successfully assembled and documented the baseline TEX-HEU experiments, that six critical configurations and one reproducibility measurement were produced, and that they are compiling results to submit to the International Criticality Safety Benchmark Evaluation Project. A future experiment campaign is planned in 2021 with Hafnium as a diluent.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Investigation of the Effect of Injected CO 2 on the Morrow B Sandstone through Laboratory Batch Reaction Experiments: Implications for CO 2 Sequestration in the Farnsworth Unit, Northern Texas, USA

About one million tons of CO 2 have been injected into the Farnsworth unit to date. The target reservoir for CO 2 injection is the Morrow B Sandstone, which is primarily made of quartz with lesser amounts of albite, calcite, chlorite, and clay minerals. The impact of CO 2 injection on the mineralogy, porosity, and pore water composition of the Morrow B Sandstone is a major concern. Although numerical modeling studies suggest that porosity changes will be minimal, significant alterations to mineralogy and pore water composition are expected. Given the implications for CO 2 storage effectiveness and risk assessment, it is crucial to verify the accuracy of theoretical model predictions through laboratory experiments. To this end, batch reaction experiments were conducted to model conditions near an injection well in the Morrow B Sandstone and at locations further away, where the CO 2 has been diluted by formation water. The laboratory experiments involved submerging thin sections of both coarse- and fine-grained facies of the Morrow B Sandstone in formation water samples with varying levels of CO 2 . The experiments were conducted at the reservoir temperature of 75 °C. Two experimental runs were conducted, one lasting for 61 days and the other for 72 days. The initial fluid composition used in the second run was the same as in the first. The mineralogy changes in the thin sections were analyzed using SEM and the Tescan Integrated Mineral Analyzer (TIMA), while changes in the composition of the formation water were determined using ICP-AES. During each experiment, a thin layer of white fine-grained particles consisting mainly of dolomite and silica formed on the surface of the thin sections, leading to significant reductions in Ca, Mg, and Sr in the formation water. This outcome is consistent with numerical model predictions that dolomite would be the primary mineral that would react with injected CO 2 and that silica would be oversaturated in the formation water. Changes in mineral abundance in the thin sections themselves were much less systematic than in the theoretical modeling experiments, perhaps reflecting heterogeneities in the mineral grain size surface area to volume ratios and mineral distributions in the thin sections not considered in the numerical models.

58 GEOSCIENCES↗
NASA NTRS2024

Space Environmental Exposure of the MISSE 9-15 Polymers and Composites Experiment 1-4 (PCE 1-4)

Spacecraft in low Earth orbit (LEO) are subjected to harsh environmental conditions, including radiation (cosmic rays, ultraviolet (UV), x-ray and charged particle radiation), micrometeoroids and orbital debris, temperature extremes, thermal cycling, and atomic oxygen (AO). These environmental exposures can result in erosion, embrittlement and optical property degradation of susceptible materials threatening spacecraft performance and durability. To increase our understanding of effects such as AO erosion and radiation induced embrittlement of spacecraft materials, NASA Glenn Research Center has developed a series of experiments that were flown as part of the Materials International Space Station Experiment (MISSE) missions on the exterior of the International Space Station (ISS). These experiments have provided critical space environmental durability data such as AO erosion data of polymers and composites, and radiation induced mechanical property degradation of spacecraft insulation materials, after long term space exposure. Recently, four Glenn experiments with 365 flight samples were flown on ISS’s MISSE-Flight Facility (MISSE-FF). These experiments are the Polymers and Composites Experiment-1 (PCE-1) flown as part of the MISSE-9 mission, the PCE-2 flown as part of the MISSE-10 mission, the PCE-3 flown as part of the MISSE-12 and MISSE-15 missions, and the PCE-4 flown as part of the MISSE-13 mission. This paper provides details on the space environment exposure of Glenn’s PCE 1-4 flight samples. Passive samples were flown in each flight orientation for post-flight AO fluence determination and on-orbit contamination analyses. Solar exposure estimates were computed by Aegis Aerospace, Inc. The environment exposure details include the total space vacuum duration, time on the MISSE-FF, the direct space exposure duration, the AO fluence and the computed mission Equivalent Sun Hours (ESH) solar exposure. Typical on-orbit temperatures are also provided along with a summary of analyses of PCE 1-4 contamination witness samples. Knowledge of the space environmental exposure of the PCE 1-4 flight samples is crucial for flight data interpretation.

Atomic oxygen↗