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At least 1,477 records · Page 82
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Coded-mask-based X-ray phase-contrast and dark-field imaging

Phase contrast and dark-field X-ray imaging enable imaging of objects that absorb or reflect very little X-ray light. Disclosed is a method and systems for performing coded-mask-based multi-contrast imaging (CMMI). The method includes providing radiation to a coded mask that has a known phase and absorption profile according to a pre-determined pattern. The radiation is then impingent upon a sample, and the radiation is detected to perform phase-reconstruction and image processing. The method and associated systems allow for the use of maximum-likelihood and machine learning methods for reconstruction images of the sample from the detected radiation.

Shi, Xianbo↗
DOE OSTI2013

Recent Improvement of Measurement Instrumentation to Supervise Nuclear Operations and to Contribute Input Data to 3D Simulation Code - 13289

The CEA has developed many compact characterization tools to follow sensitive operations in a nuclear environment. Usually, these devices are made to carry out radiological inventories, to prepare nuclear interventions or to supervise some special operations. These in situ measurement techniques mainly take place at different stages of clean-up operations and decommissioning projects, but they are also in use to supervise sensitive operations when the nuclear plant is still operating. In addition to this, such tools are often associated with robots to access very highly radioactive areas, and thus can be used in accident situations. Last but not least, the radiological data collected can be entered in 3D calculation codes used to simulate the doses absorbed by workers in real time during operations in a nuclear environment. Faced with these ever-greater needs, nuclear measurement instrumentation always has to involve on-going improvement processes. Firstly, this paper will describe the latest developments and results obtained in both gamma and alpha imaging techniques. The gamma camera has been used by the CEA since the 1990's and several changes have made this device more sensitive, more compact and more competitive for nuclear plant operations. It is used to quickly identify hot spots, locating irradiating sources from 50 keV to 1500 keV. Several examples from a wide field of applications will be presented, together with the very latest developments. The alpha camera is a new camera used to see invisible alpha contamination on several kinds of surfaces. The latest results obtained allow real time supervision of a glove box cleaning operation (for {sup 241}Am contamination). The detection principle as well as the main trials and results obtained will be presented. Secondly, this paper will focus on in situ gamma spectrometry methods developed by the CEA with compact gamma spectrometry probes (CdZnTe, LaBr{sub 3}, NaI, etc.). The radiological data collected is used to quantify the activity of hot spots and can also then be entered in 3D models of nuclear plants to simulate intervention scenarios. Recent developments and results will be presented regarding this. Finally, thanks to a large amount of feedback, the interest of using complementary measurements will be discussed. In fact, the recent use of 3D simulation codes requires very accurate knowledge of nuclear plant radiological data. The use of coupled devices such as imaging devices, (gamma and alpha cameras), gamma spectrometry, dose rate mapping, collimated / un-collimated measurements and many other physical values gives an approach to the radiological knowledge of a process or plant with the lowest possible uncertainty. In line with this, the paper will conclude with the future developments and trials that could be assessed in that field of application. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Thermomechanical simulation of the DIAMINO irradiation experiment using the LICOS fuel design code

Two separate-effect experiments in the HFR and OSIRIS Material Test Reactors (MTRs) are currently under Post- Irradiation Examinations (MARIOS) and under preparation (DIAMINO) respectively. The main goal of these experiments is to investigate gaseous release and swelling of Am-bearing UO2-x fuels as a function of temperature, fuel microstructure and gas production rate. First, a brief description of the MARIOS and DIAMINO irradiations is provided. Then, the innovative experimental in-pile device specifically developed for the DIAMINO experiment is described. Eventually, the thermo-mechanical computations performed using the LICOS code are presented. These simulations support the DIAMINO experimental design and highlight some of the capabilities of the code. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

PAREX, A numerical Code for process design and integration

The modeling approach for process design is illustrated in this paper by the work done to qualify the PAREX code to design the flow sheet of a uranium/plutonium splitting operation in a pulsed column. The different flow sheet experimentations are presented along with the simulation results. The qualified code has been successfully used to design the flow sheet of this operation in the first purification cycle of UP2-800 plant at La Hague. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Testing the Jump Finding Code for NEXUS Qubit Analysis

At Fermilab, the NEXUS dilution refrigerator is repeating the same experiment from Wilen et al. 1 underground to further investigate the effect of ionizing radiation and cosmic ray muons on qubit decoherence. This refrigerator is 100m underground in the MINOS cavern to significantly lower the incident rate of cosmic ray muons, meaning the majority of incidents is due to ionizing radiation. To gain a better understanding of how the rate of ionizing radiation affects the qubits, we use 4 radiation configurations to collect data from the chip. In this work, we examine the efficiency of our jump detection code to find jumps of various sizes. We do so using simulated data with a known number of injected jumps of known sizes. These jumps are charge offsets of the qubit, thought to be due to quasiparticle poisoning. To test this, we run the jump finding code on an array of simulated data in which two main parameters, the smoothing factor and the threshold at which a charge offset is considered a jump, are changed to test an array of these parameters. We then perform a χ2 analysis to find the best combination of these two parameters.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

Multiphysics Analysis of the MSRE Experiment Using Griffin-SAM Coupled Code System

Molten salt reactors (MSRs) with flowing fuel have the unique feature of utilizing the fuel salt for heat generation and extraction at the same time since the fuel salt is circulating through the whole primary loop of the reactor. This movement of the fuel salt results in a partial decay of the delayed neutron precursors (DNPs) outside the core and corresponding redistribution in the active core region. To capture this phenomenon accurately, neutronics and thermal hydraulics computational tools need to be modified to handle the movement of the DNPs and their decay. Idaho and Argonne National Laboratories are actively working on developing neutronics and thermal hydraulics tools to model MSRs with flowing fuel. The neutronics code Griffin and the thermal hydraulics code SAM were extended to handle flowing fuel with the drift of the DNPs under the MOOSE framework. In this work, a simplified Multiphysics model of the Molten Salt Reactor Experiment (MSRE) is developed and utilized to perform steady-state and transient analyses. The MSRE experiments were designed to show the impact of the DNPs losses on core reactivity. The following section provides a description of the MSRE experiment with the core main parameters used to develop the model. Then, the developed Multiphysics model of the MSRE is presented. Finally, steady-state verification tests and unprotected loss of flow transient tests are discussed followed by a summary and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Preliminary study of nuclear fuel element testing based on coded source neutron imaging

Neutron radiography (NR) is one of the most important nondestructive testing methods, which is sensitive to low density materials. Especially, Neutron transfer imaging method could be used to test radioactivity materials refraining from γ effect, but it is difficult to realize tomography. Coded source neutron imaging (CSNI) is a newly NR method developed fast in the last several years. The distance between object and detector is much longer than traditional NR, which could be used to test radioactivity materials. With pre-reconstruction process from fold-cover projections, CSNI could easily realize tomography. This thesis carries out preliminary study on the nuclear fuel element testing by coded source neutron imaging. We calculate different enrichment, flaws and activity in nuclear fuel elements tested by CSNI with Monte-Carlo simulation. The results show that CSNI could be a useful testing method for nuclear fuel element testing. (authors)

42 ENGINEERING↗
DOE OSTI2017

Validation of MVP code with HTC critical experiments

In the 1980's, a series of high burnups critical experiments (HTC) was conducted at the Apparatus B experimental facility in Valduc with the support of the IRSN and AREVA NC. Four series of experiments were designed to assess profit associated with actinide-only burn-up credit in the criticality safety analysis for fuel handling (Phases 1 and 2), pool storage (Phase 3) and spent fuel cask (Phase 4) conditions. The HTC rods, specifically fabricated for the experiments, were designed to simulate a typical pressurized water reactor uranium oxide spent fuel. In order to validate the MVP-2.0 code and relatively new nuclear data libraries for the criticality safety analysis, the HTC critical experiments were analyzed using the MVP-2.0 code with JENDL-4.0, ENDF/B-VII.1 or JEFF-3.2. The effective neutron multiplication factor k{sub eff} values were obtained through analyses of all phases of the HTC experiments. It was shown that averaged values of k{sub eff} using JENDL-4.0 and ENDF/B-VII.1 were in good agreement with experimental results for all the phases, while those using JEFF-3.2 showed slight underestimation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

TRACE Code Validation of BWR SVEA Fuel Assembly Spray Cooling

Emergency spray cooling injection is a safety feature implemented in many boiling water reactor designs to reflood a reactor during an accident. Significant experimental work has qualified the efficacy of spray cooling, and ongoing computational modeling efforts strive to more accurately portray the phenomena involved. This work examines the physical phenomena pertaining to emergency spray cooling injection over SVEA-type fuel assemblies. BWR emergency core cooling system (ECCS) designs commonly incorporate a Low-Pressure Core Spray (LPCS) system that introduces coolant through spray nozzles directed to the top of the core, and are located in the upper plenum region. This design is effective for LOCA response, particularly during the initial refill and reflood stages of accident response where the goal is limiting the peak cladding temperature rise in the core. The U.S. NRC thermal-hydraulics code TRACE version 5.0 Patch 4 has been chosen to simulate the separate-effect tests performed by ASEA-ATOM. The computational model was evaluated by performing forward uncertainty quantification using Dakota as the analysis tool and code driver on 31 identified parameters from the input model and TRACE physical constitutive models. This paper discusses the sensitivity of the investigated parameters with respect to spray cooling simulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

On-the-Fly Treatment of Temperature Dependent Cross Sections in the Unresolved Resonance Regions in RMC Code

Monte Carlo method can provide high fidelity neutronics analysis of different nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. It can also be coupled with thermal-hydraulics codes to consider the feedback. For thermal reactors such as PWR and HTGR, two main temperature effects should be considered. The first is the effect of thermal motion of target nuclei in resolved resonance energy regions, which is known as Doppler effect. The second is the thermal scattering and bound effect in thermal energy regions. The traditional approach of pre-generated cross sections has difficulty in memory footprint for detailed temperature modelling in multi-physics calculations. Recently, on-the-fly (OTF) technique has been proposed in order to reduce the memory consumption for both resolved resonance energy and thermal energy. For fast reactors calculations, the temperature dependence of cross sections in the unresolved resonance energy regions is also important. However, less attention was paid to the unresolved resonance regions compared with resolved resonance energy and thermal energy. In this paper, the on-the-fly temperature treatment of cross sections in unresolved resonance regions was proposed and developed in RMC codes. The results show that the on-the-fly treatment has high efficiency and satisfactory fidelity. The on-the-fly treatment of unresolved resonance region was also combined with the target motion sampling (TMS) method for resolved resonance energy. The formats of libraries were changed, in which the probability tables were separated from the cross sections. Then the interpolation of probability table was performed on-the-fly during the transport routine. The results show that the on-the-fly treatment has high efficiency and satisfactory fidelity. The on-the-fly treatment of unresolved resonance region was also combined with the target motion sampling (TMS) method for resolved resonance energy, showing that TMS method is effective and accurate for cross sections treatment in resolved resonance energy regions. The memory saving of the on-the-fly treatment is also considerable. For the high-fidelity 'transport-burnup-thermal-hydraulics' coupling calculation, the memory requirement is 60 Mb. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Investigating Accident Tolerant Fuel Concepts using the BISON Fuel Performance Code

There has been an increased interest in research of accident tolerant fuels (ATF) and claddings for use in commercial reactors ever since the events that occurred at the Fukushima Daiichi nuclear power plant in 2011. The economic impact of these events on the nuclear energy sector is significant. As a result, the United States Department of Energy's Office of Nuclear Energy accelerated research in nuclear fuel concepts with enhanced accident tolerance as part of its Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). A report produced by the AFC entitled Light Water Reactor Accident Tolerant Fuel Performance Metrics outlines a set of metrics that can be used to guide selection of promising ATF concepts. Moreover, the report specifies that a down-selection is to occur in the 2016-2017 time frame, at which time the program will move from a proof-of-concept to a proof-of-principle stage with continued research and development into a small set of concepts. Given the time necessary for comprehensive irradiation experiments and post-irradiation examination, it is impossible to perform a complete set of experiments for material characterization. Therefore, the AFC is utilizing computational analysis tools in an effort to understand the proposed materials under a variety of normal operating and accident conditions. For some time, the DOE's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has been developing computational analysis tools for nuclear fuel simulation. These codes include BISON and Marmot which investigate fuel behavior at the continuum and microstructural scales, respectively. In 2014, NEAMS introduced what it calls High Impact Problems (HIPs) into its program plan. These HIPs are intended to make significant advancements in a particular area of nuclear power research in a short period of time (3 years or less). NEAMS has chosen an ATF project, which emphasizes utilizing BISON and Marmot to model proposed materials, as its first HIP. This summary focuses on the methodology being utilized for investigation of ATF concepts using the BISON fuel performance code. First, we give an overview of the model development for two candidate ATF materials, FeCrAl for cladding and uranium silicide (U{sub 3}Si{sub 2}) for fuel. Then, we present the approach for deciding which simulations are necessary given the importance of such simulations in providing a detailed report into the fuel performance behavior of these materials by the conclusion of the HIP. In absence of experimental data and given the aggressive schedule outlined by the AFC for ATF research and development, computational analysis tools are required. These tools are being used to help guide experiment design and to perform bounding and sensitivity studies to gain valuable information for which experimental data is limited. In support of this effort, models have been incorporated into BISON for simulation of U{sub 3}Si{sub 2} and FeCrAl. Continued material model development and engineering analysis will help address questions related to the performance of these ATF materials under normal operating and accident conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Frequency spectrum analysis code for nuclear noise based on improved wavelet algorithm - 337

The core of the narrow flow channel generally faced the problem of local bubble boiling with the temperature rise of the nuclear fuel element, it is necessary and meaningful to answer and deal with the problem on the technical level. Reactor noise analysis begins with the study of the physical characteristics of the core, after nearly 30 years of research and development it has become a powerful tool in the auxiliary monitoring of the power reactor operating. By using this noise frequency spectrum analysis technology, the reactor dynamic parameter such as prompt neutron decay constant can be measured. Also, through such system, the frequency characteristics of reactor can be real-time monitored, which is helpful for safety operation and fault diagnosis. Spectrum analysis technology in the Fourier analysis era is now barely available, the reason is that the local spectral characteristics of this technology are not obvious, but this defect can be greatly improved by wavelet analysis technique, however, the traditional wavelet analysis will introduce the problem of spectral confusion. To solve the problem, based on the improved wavelet algorithm, frequency spectrum analysis code for nuclear noise is developed,from the perspective of fine spectral analysis, the local boiling accident of the coolant can be prejudged and avoid. Based on the analysis and test of experimental data,The results show that the calculation ability of the code is in accordance with the expected, it can be applied in the operation of research reactor. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Comparison of the MCNP6 and FLUKA Codes in Shielding Calculation at ProNova Proton Therapy Facility

The Monte Carlo method is a powerful tool to simulate particle transport in materials, and can be effectively used to design radiation shielding. The purpose of this study is to compare the results from two Monte Carlo codes (MCNP6 and FLUKA ) in shielding design. The calculation was also compared with the measured results from the ProNova Superconducting 360 degrees (SC360) installation at the Knoxville Provision Center for Proton Therapy (PCPT) facility. The SC360 230 MeV continue wave (CW) proton beam is produced by an isochronous cyclotron, the Sumitomo P235. The maximum beam that is extracted into the beamline is 300 nA, which is of 60% of the full current generated in cyclotron. The SC360 facility has an energy selection system for adjusting the proton range, one fixed beam and one gantry room for rotating the beam around the stationary patient. At the entrance to each treatment room beamline, a thickness adjustable Beryllium degrader inserted in the beam path is used to achieve the desirable beam energy. When protons interact with beamline components, the secondary neutrons become the dominant source to determine the radiation shielding. In this study, the dose rate outside the lateral wall of the beamline vault was discussed. The calculations from two Monte Carlo codes were compared and they agreed very well. The dose rates measured by a Wendi-II Neutron Detector agreed within a factor of less than seven compared with the calculation results. The difference may be explained by a variety of factors such as the uncertainty in knowing the exact composition of shielding materials, geometry difference, and the measurement uncertainty due to instrumentation. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Addressing Over-Solving in Multiphysics Codes with Nested Solvers

Examples of multiphysics problems in the field of nuclear engineering are numerous. For light water reactors the coupling of neutronics with thermal-hydraulics is the most predominant. Often this coupling is modeled by combining existing codes in a Picard Iteration scheme, that is, by solving one physics then using that answer to solve the other and iterating until both answers stabilize. However, simply solving to a high precision on every Picard (multiphysics) iteration is much less efficient than progressively solving to higher precisions. This unneeded work is known as over-solving. Over-solving in single-physics solvers has been studied previously for the Newton method and Chebyshev iteration. Both of these methods solve an inner linear problem that determines the next update step for the outer loop. If the inner problem is solved to a high precision in every outer iteration, the result will be substantially more computational expense (over-solving) and a higher probability of reaching the 'wrong' solution. Methods for reducing over-solving in partitioned multiphysics codes have only been developed recently. This work introduces further improvements, including an investigation of over-solving methods on both the multiphysics and single-physics levels.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Verification of High-Fidelity Neutronics Code PROTEUS for C5G7 Benchmark Problems

Under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the 3D whole core transport code PROTEUS has been developed at Argonne National Laboratory to perform the detailed 3D transport calculation for various reactor types including irregular or complex geometric reactors. PROTEUS includes two distinct transport solvers: SN and MOC. Among them, the MOC solver based on the extruded geometry, thus named MOCEX, can produce the faithful 3-D transport solutions by employing the rigorous and robust 2D/3D MOC formulation. This formulation overcomes the drawbacks of a full 3D MOC method requiring enormous memory and computational resources and resolves the instability issues of the 2D/1D method as well. The solution of MOCEX converges to that of a Monte Carlo code as the discretization of angle, space and energy is refined, unlike the 2D/1D method. As a verification and validation (V and V) effort of PROTEUS-MOCEX, this paper presents verification test results using the well-known 3D C5G7 benchmark problems by solving and comparing eigenvalues and pin powers with MCNP solutions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Verification of the Two-Fluid Model in the Spectral Element Code Nek-2P for Adiabatic Flow Conditions

The simulation of multiphase flow is a difficult physical, computational, and numerical problem. Sophisticated tools are necessary to provide results encompassing a wide range of physical phenomena. The development of these tools must be supported by rigorous analysis and ongoing efforts of verification and validation to ensure the accuracy of simulations. The Nek-2P multiphase code is being developed at Argonne with the general goal of simulating a wide range of multiphase flows and the primary goal of simulating boiling flows in LWR fuel assemblies. As part of this, there is an ongoing effort to verify the code and validate the underlying physical models. The objective of this work is to present recent results in the verification of Nek-2P. As a starting point, this work is focused on adiabatic flows for the sake of simplicity with the immediate next planned step to expand the effort to include heat and mass transfer. Details of the verification of the implemented mass fraction transport, momentum conservation, and pressure equations in Nek-2P are presented. Results of simulations are compared to analytic solutions when possible, however due to the complexity of multiphase flows, these are seldom available. To overcome this difficulty, an independent model is used to provide solutions for comparison.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Numerical Simulation of Twin Jets using LES with the Nek5000 Code

The twin jet activity, a joint effort between Texas A and M University, University of Tennessee - Knoxville, and Argonne National Laboratory, is being used to analyze the performance of various computational fluid dynamics (CFD) code packages. The twin jet water facility (TJWF) is based off a semi-classical problem of combining planar turbulent jets in a large free-shear environment or enclosure. The problem is similar to physics seen in advanced reactor designs such as in the large upper plena of a sodium fast reactor. Utilizing the experimental data sets created using laser Doppler velocimetry (LDV) and particle image velocimetry (PIV), the Nek5000 code, among others, is qualitatively compared to the experimental work. This effort is a starting point for determining the appropriate simulation techniques needed for resolving the twin jet physics.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗