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DOE OSTI2017

Structures and Heats of Formation of Simple Alkaline Earth Metal Compounds II: Fluorides, Chlorides, Oxides, and Hydroxides for Ba, Sr, and Ra

Here, geometry parameters, vibrational frequencies, heats of formation, bond dissociation energies, cohesive energies, and selected fluoride affinities (difluorides) are predicted for the late alkaline earth (Sr, Ba and Ra) oxides, fluorides, chlorides, and hydroxides at the coupled cluster theory [CCSD(T)] level. Additional corrections (scalar relativistic and pseudopotential corrections, vibrational zero-point energies, and atomic spin-orbit effects) were included to accurately calculate the total atomization energies and heats of formation following the Feller-Peterson-Dixon methodology. The calculated values are compared to the experimental data where available. In some cases, especially for Ra compounds, there are no experimental results or the experimental energetics and geometries are not reliable or have very large error bars. All of the Sr, Ba and Ra difluorides, dichlorides and dihydroxides are bent structures with the OMO bond angles decreasing going down the group. The cohesive energy of bulk Be dihalides are predicted to be quite low while those of Ra are relatively large. The fluoride affinities show that the difluorides are moderately strong Lewis acids and that such trifluorides may form under the appropriate experimental conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2013

Quantitative comparison between PGNAA measurements and MCNP calculations in view of the characterization of radioactive wastes in Germany and France

The Forschungszentrum Juelich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Fermi-Löwdin orbital self-interaction correction using the strongly constrained and appropriately normed meta-GGA functional

Despite the success of density functional approximations (DFAs) in describing the electronic properties of many-electron systems, the most widely used approximations suffer from self-interaction errors (SIE) that limit their predictive power. Herein we describe the effects of removing SIE from the strongly constrained and appropriately normed (SCAN) meta-generalized gradient approximation (GGA) using the Fermi-Lowdin Orbital Self-Interaction Correction (FLOSIC) method. FLOSIC is a size-extensive implementation of the Perdew-Zunger self-interaction correction (PZ-SIC) formalism. We find that FLOSIC-SCAN calculations require careful treatment of numerical details and describe an integration grid that yields reliable accuracy with this approach. We investigate the performance of FLOSIC-SCAN for predicting a wide array of properties and find that it provides better results than FLOSIC-LDA and FLOSIC-PBE in nearly all cases. Furthermore, it gives better predictions than SCAN for orbital energies and dissociation energies where self-interaction effects are known to be important, but total energies and atomization energies are made worse. For these properties, we also investigate the use of the self-consistent FLOSIC-SCAN density in the SCAN functional and find that this DFA@FLOSIC-DFA approach yields improved results compared to pure, self-consistent SCAN calculations. Thus FLOSIC-SCAN provides improved results over the parent SCAN functional in cases where SIEs are dominant, and even when they are not, if the SCAN@FLOSIC-SCAN method is used.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2015

Multi-step Monte Carlo calculations applied to nuclear reactor instrumentation - source definition and renormalization to physical values

Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g. miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. These data are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations with satisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm{sup 3}). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position, the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presented in the paper, the injudicious use of special options aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from different types of SPNDs, which were recently irradiated in the Jozef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia, and provides recommendations on how they can be overcome. The paper concludes with a discussion on the renormalization of the results from the second step calculations, to obtain accurate physical values. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development at the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Long-Term Infrasound Sensor Calibration and Characterization

Previous testing has shown that infrasound sensors deployed in the field can exhibit notable deviations from their nominal, lab-based calibrations. These variations may be due to changes in environmental conditions, long-term sensor drift, or other unresolved features. In early 2018, we installed two identical infrasound elements with five infrasound sensors at each element (Chaparral M50A, Chaparral M64LN, CEA/Martec MB2005, CEA/Seismowave MB3a, and Hyperion IFS-5113A). These sensors were accepted or under consideration for use in the International Monitoring System network of the Comprehensive Nuclear-Test-Ban Treaty. Each element had all sensors connected to a single digitizer and port to the atmosphere. We also recorded instrument enclosure air temperature and humidity and external air temperature. Using the MB2005 as the reference, we examine the relative sensor response (both magnitude and phase) as a function of time and frequency and compare it with quarterly laboratory calibrations and environmental conditions. Here, we find that the magnitude response for all sensors exhibits some variability in both the lab and field, with the amplitude variations often >5%. The field-based variations are more severe and occur on both long-term (months) and short-term (hours) timescales. Short-term variability correlates with changes in environmental conditions and is considerable (up to 25%) for the Chaparral M50A and noticeable (∼5%) for the French Alternative Energies and Atomic Energy Commission (CEA) MB3a. Long-term magnitude variability for the Chaparral M50A was up to 50% and does not clearly correlate with environmental conditions. The other sensors show some long-term magnitude offsets, but they have relatively stable responses in the conditions we examined. The MB3a also displays some frequency-dependent magnitude variability and shows a minor dependence on temperature. Phase estimates are stable and near zero for all sensors tested. These results strongly suggest sensor response and variability due to environmental conditions should be considered in future infrasound data interpretation and sensor selection and development.

Fee, David↗

DOE OSTI2014

MB3a Infrasound Sensor Evaluation.

Sandia National Laboratories has tested and evaluated a new infrasound sensor, the MB3a, manufactured by Seismo Wave. These infrasound sensors measure pressure output by a methodology developed by researchers at the French Alternative Energies and Atomic Energy Commission (CEA) and the technology was recently licensed to Seismo Wave for production and sales. The purpose of the infrasound sensor evaluation was to determine a measured sensitivity, transfer function, power, self-noise, dynamic range, seismic sensitivity, and self- calibration ability. The MB3a infrasound sensors are being evaluated for potential use in the International Monitoring System (IMS) of the Comprehensive Nuclear Test-Ban-Treaty Organization (CTBTO).

Merchant, Bion J.↗

DOE OSTI2014

NEET Micro-Pocket Fission Detector. Final Project report

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

LLNL Results from CALIBAN-PROSPERO Nuclear Accident Dosimetry Experiments in September 2014

Lawrence Livermore National Laboratory (LLNL) uses thin neutron activation foils, sulfur, and threshold energy shielding to determine neutron component doses and the total dose from neutrons in the event of a nuclear criticality accident. The dosimeter also uses a DOELAP accredited Panasonic UD-810 (Panasonic Industrial Devices Sales Company of America, 2 Riverfront Plaza, Newark, NJ 07102, U.S.A.) thermoluminescent dosimetery system (TLD) for determining the gamma component of the total dose. LLNL has participated in three international intercomparisons of nuclear accident dosimeters. In October 2009, LLNL participated in an exercise at the French Commissariat à l’énergie atomique et aux énergies alternatives (Alternative Energies and Atomic Energy Commission- CEA) Research Center at Valduc utilizing the SILENE reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison at CEA Valduc, this time with exposures at the CALIBAN reactor (Hickman et al. 2011). This paper discusses LLNL’s results of a third intercomparison hosted by the French Institut de Radioprotection et de Sûreté Nucléaire (Institute for Radiation Protection and Nuclear Safety- IRSN) with exposures at two CEA Valduc reactors (CALIBAN and PROSPERO) in September 2014. Comparison results between the three participating facilities is presented elsewhere (Chevallier 2015; Duluc 2015).

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

Summary of Technical Meeting To Compare US/French Approaches for Physical Protection Test Beds

In September 2015, representatives of the US Department of Energy/National Nuclear Security Administration, including test bed professionals from Sandia National Laboratories, and representatives of the French Alternative Energies and Atomic Energy Commission participated in a one-week workshop to share best practices in design, organization, operations, utilization, improvement, and performance testing of physical protection test beds. The intended workshop outcomes were to (1) share methods of improving respective test bed methodologies and programs and (2) prepare recommendations for standards regarding creating and operating testing facilities for nations new to nuclear operations. At the workshop, the French and American subject matter experts compared best practices as developed at their respective test bed sites; discussed access delay test bed considerations; and presented the limitations/ constraints of physical protection test beds.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

CEA SMAD 2016 Digitizer Evaluation.

Sandia National Laboratories has tested and evaluated an updated SMAD digitizer, developed by the French Alternative Energies and Atomic Energy Commission (CEA). The SMAD digitizers are intended to record sensor output for seismic and infrasound monitoring applications. The purpose of this digitizer evaluation is to measure the performance characteristics in such areas as power consumption, input impedance, sensitivity, full scale, self-noise, dynamic range, system noise, response, passband, and timing. The SMAD digitizers have been updated since their last evaluation by Sandia to improve their performance when recording at a sample rate of 20 Hz for infrasound applications and 100 Hz for hydro-acoustic seismic stations. This evaluation focuses primarily on the 20 Hz and 100 Hz sample rates. The SMAD digitizers are being evaluated for potential use in the International Monitoring System (IMS) of the Comprehensive Nuclear Test- Ban-Treaty Organization (CTBTO).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

Characterizing the Performance of Fission Chambers for Local Neutron Flux and Spectrum Measurement

Three fission chambers were tested at The Ohio State University Research Reactor (OSURR) to evaluate fission chambers for neutron flux and spectrum measurements. Two of the fission chambers are commercial Photonis CFUR43 fission chambers—one utilizing highly enriched uranium-235 fissile deposits for thermal neutron detection and the other with highly depleted uranium-238 for fast neutron detection. This experiment was also performed in collaboration with French Alternative Energies and Atomic Energy Commission (CEA) with CEA suppling a uranium-235 loaded fission chamber built from the Photonis CFPR CE8/9 kit. The fission chambers were irradiated in a movable 6.5 inch dry tube at reactor powers of 50 kW, 200 kW, and 450 kW at ambient temperature and at 350°C. While all sensors demonstrated good linear sensitivity to reactor power, a leakage current of 6 nA was measured for the CFUR43 detectors. Additionally, the measured signal ratios of the two CFUR43 detectors did not match the theoretical values and are subject to further investigation on individual signal contributors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

Keeping LANL at the forefront of list mode capability for safeguards and MC&A [Slides]

Neutron based detection forms key capability to enable and support international and domestic nuclear safeguards and non-proliferation missions. Key stakeholders include International Atomic Energy Agency (IAEA), European Atomic Energy Community (EURATOM), regional safeguards inspectorates, domestic stakeholders. LANL has been historically a leader in development of Non-destructive Assay (NDA) instruments and associated data acquisition hardware and software.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2022

Modeling and Simulation of Austenitic Welds and Coarse-grained Specimens: Part II

The Pacific Northwest National Laboratory (PNNL) is conducting confirmatory research for the U.S. Nuclear Regulatory Commission (NRC) to evaluate commercially available nondestructive examination (NDE) modeling and simulation software used in the nuclear industry. Simulation results from ultrasonic testing (UT) models can inform the design and qualification of inspection techniques and help interpret inspection results. CIVA is a modeling and simulation package developed by the French Alternative Energies and Atomic Energy Commission (CEA). CIVA was selected for this study because it is readily available and has been used for NDE in the US commercial nuclear power industry. This report is focused on completing the efforts initiated in the previous PNNL report to evaluate UT modeling and simulation performance, reliability, and accuracy in relation to common inservice inspection (ISI) scenarios in nuclear power plants (NPP). This work will be used to provide guidance when establishing methods to perform and evaluate simulations for more standardized model implementation, simulation analysis, and interpretation of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Performance Benchmark of Commercial and Developmental Fission Chambers in Elevated Temperatures

This report documents the testing of two in-core fission chamber technologies for high-temperature irradiation environments. This work is in collaboration with the French Alternative Energies and Atomic Energy Commission (CEA). The fission chamber evaluated by Idaho National Laboratory is the micro-pocket fission detector (MPFD). The fission chamber evaluated by the CEA are the 3 mm miniaturized fission chamber and the 7 mm high-temperature fission chamber. Demonstrations of the MPFD were performed at the Neutron Radiography Facility and the Massachusetts Institute of Technology Reactor. Demonstrations of the CEA fission chambers were performed at the Ohio State University Research Reactor. All demonstrations were conducted with a heated experiment rig up to 850°C.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2026

2025 CEA SMAD Digitizer Type Approval Evaluation

47 OTHER INSTRUMENTATION↗

NASA NTRSDate not supplied

Ground Tests on the BOLT Geometry at Mach 7: Cross-Facility Comparison and Stability Analysis

Transition measurements have been obtained through two experimental campaigns conducted independently by the German Aerospace Center and the French Aerospace Lab with the French Alternative Energies and Atomic Energy Commission on subscale models of the BOLT-1 flight experiment geometry. This paper details a cross-facility comparison of measurements obtained at Mach 7, as well as subsequent computational analysis. Infrared (IR) thermography measurements obtained by both campaigns have facilitated a global comparison of the transition front across facilities at analagous conditions, which are found to be in good agreement. High-frequency surface pressure fluctuation data demonstrate significant amplification of instabilities with Mack-mode characteristics in the outboard regions of the acreage. These measurements are compared to stability analyses of varying fidelity. The computational methods employed to characterize the boundary-layer transition phenomena include the traditional line-marching implementation of the Parabolized Stability Equations (PSE), 2D eigenvalue analysis coupled with PSE, and Adaptive Mesh Refinement Wavepacket Tracking. While the 2D eigenvalue analysis is found to predict instabilities, which correlate in terms of frequency and acreage location to the experimental measurements, the predicted amplification for these instabilities is lower than would typically be expected for transition. Line-marching results for traveling crossflow produce the best match to the experimental transition front, with a consistent transition N factor of approximately 3-3.5.

Boundary Layer Transition↗