Analysis of Uranium Enrichment with New CZT Detectors
Abstract not provided
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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Beryllium Laboratory Analysis--The Regulations May Drive the Science
Beryllium has many industry-specific applications, such as medical X-ray windows for diagnostic equipment, nuclear reactors, aerospace applications, precision instrumentation, and other consumer products for which lightness and rigidity is essential. According to the National Toxicology Program, beryllium oxide (BeO) is one of the most significant beryllium compounds in production. Although beryllium and its compounds have a wide array of beneficial uses, due to its unique properties it is not an ideal metal to be used in all situations. Exposure to beryllium is linked to beryllium sensitization and Chronic Beryllium Disease (CBD), which is incurable, debilitating, and potentially fatal. The International Agency for Research on Cancer classifies beryllium and beryllium compounds as “carcinogenic to humans” (Group I), and EPA classifies beryllium as a likely human carcinogen, the lung being the primary target organ. Laboratory analysis for beryllium samples has always presented a challenge to the analytical community. While most metals of interest to industrial hygienists have occupational exposure limits (OELs) in milligrams per cubic meter (mg/m 3 ), the beryllium OELs are in micrograms per cubic meter (μg/m3). Some regulatory agencies have recently published beryllium OELs so low that in some cases a laboratory limit of detection (LOD) in nanograms (ng) is required. For most substances, science drives the regulations, but for beryllium, regulations appear to be driving science to develop laboratory analytical methods that can adequately support the proposed OELs. (EPA has issued guidelines regarding ambient and community airborne beryllium exposure, but this article focuses on beryllium from an occupational exposure perspective.)
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Nuclear Science User Facility: Neutron Irradiation Challenges and Opportunities
The Nuclear Science User Facilities (NSUF), established in 2007 as the Advanced Test Reactor (ATR) National Scientific User Facility is chartered to provide cost-free access to unique and important research and development capabilities and infrastructure to nuclear energy-focused researchers. Users can apply for access to any of the NSUF INL or partner facilities using the same process, either through the annual Office of Nuclear Energy Consolidated Innovative Nuclear Research funding announcement or the NSUF Rapid Turnaround Experiment (RTE) opportunities, offered three times each year. The NSUF has several research and test reactors available as well as ion beamlines and x-ray light sources for experimental use in support of the R&D objectives of the DOE Office of Nuclear Energy. This presentation reviews the current capabilities, needs, awarded irradiation experiments for FY2018.
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Control rod position indicator
A rod position indication system includes a drive rod operably coupled to a control rod that is configured to be both withdrawn from and inserted into a reactor core. A number of sensing devices are arranged along a path of the drive rod, and an end of the drive rod passes by or through one or more of the sensing devices in response to movement of the control rod relative to the reactor core. The sensing devices are arranged into a plurality of groups, and each group includes two or more of the sensing devices electrically coupled together. The rod position indication system further includes a control rod monitoring device electrically coupled to each group of sensing devices by a routing wire.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
FY 2019 Nuclear Energy Infrastructure Database Accomplishments
The database, known as the Nuclear Energy Infrastructure Database (NEID), is accessed through a web-based application (at nsuf-infrastructure.inl.gov) and built specifically to track research and development (R&D) facilities and associated equipment throughout the DOE-NE complex. The NEID serves internal users (such as DOE personnel and national laboratory staff and researchers) and potential infrastructure users who will be able to view the inventory of DOE-NE-compatible R&D capabilities potentially available to them. NSUF utilizes the NEID to support the annual nuclear energy gap analysis process. Infrastructure needs are gathered from DOE-NE R&D programs and researchers at U.S. national laboratories, universities, and the U.S. nuclear industry. The fiscal year (FY) 2019 Nuclear Energy Infrastructure Database (NEID) Accomplishments report captures all the additions and improvements made to the database in FY 2019.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
An Expert System to Assist Researchers with Nuclear Materials Experiment Design
The Nuclear Science User Facilities (NSUF) was created in 2007 to open National Laboratory capabilities to university, laboratory, and industry researchers. Originally founded at the Idaho National Laboratory (INL), NSUF leveraged excess capacity in the Advanced Test Reactor (ATR) and the Hot Fuel Examination Facility (HFEF) hot cells. In the eleven years since its inception, the NSUF has expanded to include most of the capabilities at the Idaho National Laboratory as well as twenty external ‘partner’ facilities. The inclusion of the partner facilities makes the NSUF the first distributed user facility, a ‘laboratory without walls.’ This paper will present recent developments in CoMET, including how the databases interconnect to create a powerful tool for NSUF users and how the Wizard is expected to change the workflow for NSUF users, staff, and partner facilities. The paper will conclude with some remarks of how CoMET demonstrates the feasibility of a broader integration of nuclear information and tools, paving the way for further collaboration and innovation following the NuKIT paradigm.
07 ISOTOPE AND RADIATION SOURCES↗
Non-invasive in-situ imaging of interior of nuclear reactors
Techniques, systems, and devices are disclosed for non-invasive monitoring and imaging of nuclear fuel inside a nuclear reactor using muon detector arrays. In one aspect, these detector arrays are placed outside the reactor vessel or building for investigating the reactors without access to the cores, therefore the imaging process is non-invasive. In some implementation, these detector arrays measure both muon scattering and absorption to enable imaging and characterizing not only the very high-Z fuel materials, but also other materials in the reactor, thereby obtaining a more complete picture of reactor status. When applied to damaged reactors, the disclosed proposed techniques, systems, and devices, through the process of providing an image, can reveal the presence (or absence) of damage to fuel rod assemblies or puddles of molten fuel at the bottom of the containment vessel, thus providing crucial information to guide decisions about remedial actions.
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VERA Software Validation using In-Plant Data
Explore the source record for details and available documents.
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FY 2022 Scientific Infrastructure Support Funding Opportunity Announcement
Presentation to support the FY2022 Scientific Infrastructure Support FOA at the DOE-HQ webinar.
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Quantitative Risk Analysis of High Safety-significant Safety-related DI&C Systems using IRADIC Technology
Two slides prepared for RISA Stakeholder Meeting held on October 14th.
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Quantifying Burst Repeatability for Godiva IV
The minimum resolvable relative change in the neutron source emission of Godiva IV between bursts, i.e., burst repeatability, was quantified using a combination of activation monitors and a resistance temperature detector (RTD). Measurements support changes in the neutron source emission between bursts can be determined with a precision of 2.7%. The measurements were performed during a series of twelve Godiva IV bursts and were a collaborative effort between Los Alamos National Laboratory, Sandia National Laboratories, and Pacific Northwest National Laboratory. Gold, iron, nickel, and sulfur activation monitors were positioned in a sample holder in the central cavity of Godiva IV for each burst. The RTD is part of the Godiva burst diagnostic system and is in a fixed location in the safety block of Godiva IV. It is used to determine the temperature rise of the burst, i.e., the difference between the maximum temperature recorded during the burst and the initial temperature prior to the burst. Differences in either the specific activity of the activation monitor reactions, or the temperature rise can be used to determine the relative change in the neutron source emission for separate bursts, but historic results have proven that each method is susceptible to various sources of biases which generally cannot be isolated using a single method. We address this issue by applying both measurement methods and analyzing the correlation between them. The results presented in this work will be used to support future experiments with Godiva IV at the National Criticality Experiments Research Center (NCERC).
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Optimization of a neutron detector design using adjoint transport simulation
A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Organic scintillation detector response simulation using non-analog MCNPX-PoliMi
Organic liquid scintillation detectors are valuable for the detection of special nuclear material since they are capable of detecting both neutrons and gamma rays. Scintillators can also provide energy information which is helpful in identification and characterization of the source. In order to design scintillation based measurement systems appropriate simulation tools are needed. MCNPX-PoliMi is capable of simulating scintillation detector response; however, simulations have traditionally been run in analog mode which leads to long computation times. In this paper, non-analog MCNPX-PoliMi mode which uses variance reduction techniques is applied and tested. The non-analog MCNPX-PoliMi simulation test cases use source biasing, geometry splitting and a combination of both variance reduction techniques to efficiently simulate pulse height distribution and then time-of-flight for a heavily shielded case with a {sup 252}Cf source. An improvement factor (I), is calculated for distributions in each of the three cases above to analyze the effectiveness of the non-analog MCNPX-PoliMi simulations in reducing computation time. It is found that of the three cases, the last case which uses a combination of source biasing and geometry splitting shows the most improvement in simulation run time for the same desired variance. For pulse height distributions speedup ranging from a factor 5 to 25 is observed, while for time-of-flights the speedup factors range from 3 to 10. (authors)
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Inverse transport problem solvers based on regularized and compressive sensing techniques
According to the direct exposure measurements from flash radiographic image, regularized-based method and compressive sensing (CS)-based method for inverse transport equation are presented. The linear absorption coefficients and interface locations of objects are reconstructed directly at the same time. With a large number of measurements, least-square method is utilized to complete the reconstruction. Owing to the ill-posedness of the inverse problems, regularized algorithm is employed. Tikhonov method is applied with an appropriate posterior regularization parameter to get a meaningful solution. However, it's always very costly to obtain enough measurements. With limited measurements, CS sparse reconstruction technique Orthogonal Matching Pursuit (OMP) is applied to obtain the sparse coefficients by solving an optimization problem. This paper constructs and takes the forward projection matrix rather than Gauss matrix as measurement matrix. In the CS-based algorithm, Fourier expansion and wavelet expansion are adopted to convert an underdetermined system to a well-posed system. Simulations and numerical results of regularized method with appropriate regularization parameter and that of CS-based agree well with the reference value, furthermore, both methods avoid amplifying the noise. (authors)
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Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels
The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)
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Neutron noise measurements at the Delphi subcritical assembly
The paper presents the results and evaluations of a comprehensive set of neutron noise measurements on the Delphi subcritical assembly of the Delft Univ. of Technology. The measurements investigated the effect of different source distributions (inherent spontaneous fission and {sup 252}Cf) and the position of the detectors applied (both radially and vertically). The evaluation of the measured data has been performed by the variance-to-mean ratio (VTMR, Feynman-{alpha}), the autocorrelation (ACF, Rossi-{alpha}) and the cross-correlation (CCF) methods. The values obtained for the prompt decay constant show a strong bias, which depends both on the detector position and on the source distribution. This is due to the presence of higher modes in the system. It has been observed that the {alpha} value fitted is higher when the detector is close to the boundary of the core or to the {sup 252}Cf point-source. The higher alpha-modes have also been observed by fitting functions describing two alpha-modes. The successful set of measurement also provides a good basis for further theoretical investigations including the Monte Carlo simulation of the noise measurements and the calculation of the alpha-modes in the Delphi subcritical assembly. (authors)
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In-Containment Signal Acquisition and Data Transmission via Power Lines within High Dose Areas of Nuclear Power Plants
Signal acquisition and data transmission for innovative sensor systems and networks inside the containment of nuclear power plants (NPPs) is still a challenge with respect to safety, performance, reliability, availability, and costs. This especially applies to equipment upgrades for existing plants, special measurements, but also for new builds. This paper presents a novel method for efficient and cost-effective sensor signal acquisition and data transmission via power lines, in order to cope with the disadvantages of common system architectures that often suffer from poor signal integrity due to raw data transmissions via long cables, huge efforts and costs for installation, and low flexibility with respect to maintenance and upgrades. A transmitter-receiver architecture is proposed that allows multiplexing of multiple sensor inputs for unidirectional point-to-point transmission by superimposing information signals on existing AC or DC supply lines, but also on active and inactive sensor wires, or spare cables, using power line communication (PLC) technology. Based on commercial off-the-shelf (COTS) electronic parts, a radiation hard transmitter hardware is designed to operate in harsh environment within the containment during full plant operation. The system's modular approach allows application specific trade-offs between redundancy and throughput regarding data transmission, as well as various sensor input front-ends which are compatible with state of the art systems. PLC technology eliminates the need for costly installation of additional cables and wall penetrations, while providing a complementary and diverse communication technology for upgrades of existing systems. At the receiver side in low dose areas, signals are extracted from the power line, demodulated, and de-multiplexed, in order to regain the original sensor signal information and provide it either in analog or digital output format. Successful laboratory qualification tests, field trails, and a demonstrator setup installation within a NPP verify the overall performance of the proposed system. (authors)
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An Industrial Steerable Needle for Inspecting Extremely Confined Spaces in Nuclear Facilities - 18316
We present a novel, compliant mechanism inspired by steerable needles for endoscopic surgery that provides the capability to navigate extremely confined spaces for the purpose of infrastructure inspection. Extremely confined spaces are commonly encountered during infrastructure inspection of nuclear facilities. These spaces can include pipes and ventilation ducts. Often these infrastructure features go uninspected simply because there is no viable way to access their interior. In addition, it is not uncommon for extremely confined spaces to be laid-out in a spider-web fashion that must be selectively navigated in order to execute an inspection. Recently, millimeter-scale imagers have become commercially available. These imagers are small enough to fit inside an extremely confined space, but there is currently no means to bring the imager to the portion of the structure that must be inspected. In this work this problem is addressed by taking inspiration from the field of steerable needles used in endoscopic surgery. Steerable needles are long flexible shafts that are capable of being directed through the human body using control forces applied external to the body. Unfortunately, steerable needles cannot be directly applied to the problem of navigating maze-like arrangements of extremely confined spaces. Traditional steerable needles must operate within a solid medium that can be penetrated in order to be steered so they are not directly applicable to navigating extremely confined spaces. However surgical steerable needles do suggest an approach to the problem. Steerable needles are typically made from super-elastic nitinol. This material has the property that it can experience high strain without yielding. Furthermore, superelastic nitinol is stiff enough that it is possible to push the tube down a long, maze-like arrangement of a confined space without buckling. The stiffness properties of nitinol combined with its high yield strain makes it an attractive material for navigating a tube through a maze-like arrangement of extremely confined spaces. The ability to selectively navigate extremely confined spaces opens up new possibilities to use emerging miniature imaging technology for infrastructure inspection. (authors)
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