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DOE OSTI2013

Design and implementation of a 150 GHz single-channel millimeter wave interferometer on Joint TEXT tokamak

A simple, single-channel millimeter-wave interferometer system has been designed, fabricated, and installed on the J-TEXT tokamak. For the plasma density anticipated on J-TEXT, a solid-state source operating at 150 GHz has been chosen to minimize errors due to both vibration along the beam path and refraction in the plasma. The new aspect of the interferometer design is to use a subharmonic mixer for detection with a frequency doubled 150 GHz source. It employs a single source which is bias-tuned and modulated with a sawtooth wave form up to 100 kHz in order to generate the intermediate frequency. The 12.5 GHz voltage-controlled oscillator is multiplied to 75 GHz before a final doubler raises it to 150 GHz. A portion of the 75 GHZ power is used for the local oscillator (LO) and is directly connected to the LO input of the subharmonic mixer. The phase is evaluated by a digital phase comparator using a software-based algorithm. Detection noise limits the minimum resolvable phase change with the interferometer to {+-}0.05 fringe, which corresponds to an averaged electron density change along the chord of {+-}1.1 Multiplication-Sign 10{sup 17} m{sup -2}. The maximum measurable electron density is expected to be {approx}9 Multiplication-Sign 10{sup 19} m{sup -3}. A comparison of preliminary results from the millimeter wave interferometer with that from the far-infrared hydrogen cyanide laser (wavelength of 337 {mu}m) interferometer shows good agreement during the pulse flat-top period. The millimeter wave interferometer system will be used as a part of the density feedback control system in the future.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

TH-CD-BRA-07: MRI-Linac Dosimetry: Parameters That Change in a Magnetic Field

Purpose: In MRI-linac integrated systems, the presence of the magnetic (B-)field has a large impact of the dose-distribution and the dose-responses of detectors; yet established protocols and previous experience may lead to assumptions about the commissioning process that are no longer valid. This study quantifies parameters that change when performing dosimetry with an MRI-linac including beam quality specifiers and the effective-point-of-measurement (EPOM) of ionization chambers. Methods: We used the Geant4 Monte Carlo code for this work with physics parameters that pass the Fano cavity test to within 0.1% for the simulated conditions with and without a 1.5 T B-field. A point source model with the energy distribution of an MRI-linac beam was used with and without the B-field to calculate the beam quality specifiers %dd(10)× and TPR{sup 20}{sub 10}, the variation of chamber response with orientation and the how the B-field affects the EPOM of ionization chambers by comparing depth-dose curves calculated in water to those generated by a model PTW30013 Farmer chamber. Results: The %dd(10)× changes by over 2% in the presence of the B-field while the TPR{sup 20}{sub 10} is unaffected. Ionization chamber dose-response is known to depend on the orientation w.r.t. the B-field, but two alternative perpendicular orientations (anti-parallel to each other) also differ in dose-response by over 1%. The B-field shifts the EPOM downstream (closer to the chamber center) but it is also shifted laterally by 0.27 times the chamber’s cavity radius. Conclusion: The EPOM is affected by the B-field and it even shifts laterally. The relationship between %dd(10)× and the Spencer-Attix stopping powers is also changed. Care must be taken when using chambers perpendicular to the field as the dose-response changes depending on which perpendicular orientation is used. All of these effects must be considered when performing dosimetry in B-fields and should be accounted for in future dosimetry protocols. This project was partially funded by Elekta Ltd.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

Predicting deviations in ex-core detector response in Krsko NPP during an earthquake with Monte-Carlo neutron transport methods

To study the response of ex-core detectors during an earthquake, an approach using Monte Carlo code MCNP was developed. As a first step to predict the response of the ex-core detectors a detailed MCNP model of the Krsko reactor core was created. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krsko nuclear power plant. Due to the outside the core position of the neutron detectors, the hybrid code ADVANTG is used to generate weight windows to speed up neutron transport outside the reactor core. To be able to use ADVANTG, fixed neutron source had to be reconstructed from the criticality core calculation. It was found that the power range ex-core detector movement has a negligible effect on the value of the thermal neutron flux at the active part of the detector. However, the 5 cm radial movement of the intermediate-range detector leads to 7 % - 8 % change in thermal neutron flux within the active intermediate-range detector region. The analysis continued with the evaluation of the effect of core barrel movement on the ex-core detector response. It was determined that the 2 mm core barrel radial oscillation can lead to ∼ 4% change in thermal neutron flux within the active detector region. It was evaluated that reactor core movement inside the baffle can lead up to ∼ 1.9% and ∼ 1.3% deviation in power-range and intermediate-range detector signal respectively. Analysis showed that the mechanical movement of ex-core neutron detectors could not explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Radiation shielding analysis of the concrete hutch of a long beamline of the Advanced Photon Source Upgrade project

The radiation shielding analysis of the external concrete enclosure of a feature long beamline of the Advanced Photon Source Upgrade project is described in this work. The beamline has 2 undulators that can operate in two different configurations and the source terms are estimated for the worst configuration using the SRW code in the OASYS package as well as the STAC8 code. The shielding analysis is carried out using the STAC8 code, and for the case of pink beam, with the FLUKA code also. Effective dose rates on contact with the outside surface of the concrete walls for pink, reflected full spectrum and monochromatic beams are calculated. For the monochromatic beams, 2 sets of discrete bandwidths (BW) from the Double Crystal Monochromator (DCM) and the full reflectance of the Double Multilayer Monochromator (DMM) are used. Further, due to the size of the direct beam and the energy domain of interest in the direct and scattered fields, the importance of the equivalent dose to the lens of the eye is discussed with respect to the shielding calculations of such facilities. The dose rates outside the concrete hutch in the lateral directions calculated using STAC8 is about 50% higher than the FLUKA results for the case of pink beam but may be considered as a reasonable agreement due to the differences in the calculational methodologies. Those dose rates in the lateral and forward directions are higher for the pink beam to allow full occupancy from the radiation protection point of view. For the full spectrum after 3 reflections, the effective dose rates are less than 0.5 Sv/h in the lateral direction but is higher in the forward region below about 22° with respect to the beam direction. For the DCM beams, except for the scenario with one mirror and 0.1%BW, all dose rates are less than 0.5 Sv/h outside the concrete hutch in the lateral directions. Here, the higher harmonics are what contributes to the dose outside the shield and the bandwidths assumed to derive the photon flux thus becomes important. The dose rates with the DMM are generally lower compared to the 0.1% BW DCM beam but are higher than the results obtained with XOP calculated bandwidths.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Sensitive Resources Assessment and Forest Analysis for the Proposed Versatile Test Reactor, Oak Ridge, Tennessee

The US Department of Energy’s (DOE’s) Oak Ridge National Laboratory (ORNL) is a leading institution in advanced materials, supercomputing, neutrons, and nuclear science. As a research laboratory managed by UT-Battelle, LLC for DOE, ORNL has national priorities in energy, security, and scientific discovery that necessitate facility improvements and expansions. DOE is also committed to environmental stewardship. The laboratory is located on the ~32,000-acre (~13,000-ha) Oak Ridge Reservation (ORR), much of which is categorized as a National Environmental Research Park (NERP) and a state Wildlife Management Area. DOE works with the Tennessee Wildlife Resources Agency (TWRA), Tennessee Department of Environment and Conservation (TDEC), US Fish and Wildlife Service (USFWS), US Department of Agriculture, and other agencies to serve as an effective steward of the ORR. Accordingly, project managers must conform to environmental regulations, agreements, and policies at the federal, state, and institutional levels. Per 40 CFR (Code of Federal Regulations) 1508.14, potential effects on research and science education also represent potential effects of federal actions on the NERP, and impacts on, e.g., deer harvest, must be considered on the Oak Ridge Wildlife Management Area when other aspects of the human environment are affected. The United States currently has no fast neutron testing capability to support advanced nuclear research and development. The proposed Versatile Test Reactor (VTR) will take advantage of current investments by the US government and private industry in nuclear reactors to expedite the design and construction process, using proven technology to create a world-class scientific infrastructure. The VTR will take advantage of fast neutrons provided by this proven technology, along with a capability to rapidly insert, conduct, and remove state-of-the-art experiments. An advantage of the VTR is that it can support future innovations in experimental capabilities without modifying the facility. The VTR will support progress in a variety of science and technology areas, including testing and qualification of advanced reactor fuels; testing and qualification of innovative structural materials; testing of innovative components and instruments; validation of advanced modeling and simulation tools; and versatility for future technical missions. Through proven technology, the VTR can take advantage of existing reactor designs and operating experience to reduce the risk, cost, and time for design and construction. The top available resources of DOE laboratories, industry, and universities will be used to expedite reactor design and construction toward developing the scientific infrastructure that affords a strong testing capability that can be sustained over many years. This report summarizes current knowledge of natural and cultural resources primarily within the VTR construction area. At the time of this report, the proposed VTR site design includes a construction area of ~150.4 acres (~69.9 ha), which contains an ~51.3-acre (~20.8-ha) operations area, located within forested natural areas of the ORR. The primary goal of the work presented here was to evaluate potential effects on sensitive resources that might result from development and construction activities associated with VTR. In addition to on-the-ground surveys during spring and summer 2020 by the ORNL Natural Resources Management Program and Aquatic Ecology Group staff, this report makes use of historical (pre-1995) and contemporary (1995 to present) data from additional confirmed sources (e.g., TDEC). Likewise, forest conditions were compiled from a 2011 forest inventory and supplemented with limited ground observations in 2020. The individuals who obtained and compiled the data presented here are familiar with and routinely assess sensitive resources on the ORR. Anyone who references this report must consider that the timing of surveys did not permit a complete delineation of the resources that will be affected. If the VTR project proceeds, additional surveys will be required to account for the seasonal patterns of various threatened and endangered species. Data deficiencies and potential resources that likely went undetected are indicated where possible. Accordingly, this report should facilitate more environmentally sound decisions during planning and 2 development of the VTR site, provide a foundation for further assessment of sensitive and cultural resources, and help project managers better address regulatory guidance and DOE policies on sustainable development in compliance with, for example, the US Endangered Species Act (ESA), Migratory Bird Treaty Act (MBTA), Tennessee Rare Plant Protection and Conservation Act of 1985, Tennessee Nongame and Endangered or Threatened Wildlife Species Conservation Act of 1974, several federal and state regulations regarding aquatic resource protection, and site-specific policies as outlined in various ORR management plans developed by ORNL and TWRA for DOE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2014

Self characterization of a coded aperture array for neutron source imaging

The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning deuterium-tritium plasma during the stagnation stage of inertial confinement fusion implosions. Since the neutron source is small (∼100 μm) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-μm resolution are 20-cm long, triangular tapers machined in gold foils. These gold foils are stacked to form an array of 20 apertures for pinhole imaging and three apertures for penumbral imaging. These apertures must be precisely aligned to accurately place the field of view of each aperture at the design location, or the location of the field of view for each aperture must be measured. In this paper we present a new technique that has been developed for the measurement and characterization of the precise location of each aperture in the array. We present the detailed algorithms used for this characterization and the results of reconstructed sources from inertial confinement fusion implosion experiments at NIF.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Development of the Intelligent, Preventive Infrared (IR) Inspection System Housed in Hybrid Robotic Platforms

Robots and robotic systems that are designed for inspection, environmental study, and health and safety aid are becoming an increasing necessity. However, there is a number of challenges that accompany robots that are designed for these specific applications. These challenges include: navigating compact, enclosed spaces, travelling over multiple terrains and large obstacles, using the proper sensing and detection methods to assess an environment, and the use of lightweight and durable materials. The robotic platforms currently in development look at all of these challenges and attempt to overcome them. These designs specific use of hybrid robotic platforms, or platform that utilizes soft and rigid materials, allows for a more flexible platform and makes environments more navigable. To further improve the navigation of the platforms and environmental assessment, a novel infrared detection system is housed in the platforms to create a robot that can be used for the applications listed above and more. Objectives: Further develop two types of robotic platforms that utilize additive manufacturing, soft materials, and rigid materials. Continue the development of an intelligent and inhibitory infrared detection system based on an artificial intelligence (AI) algorithm. Continued study and fabrication of active soft materials designed for both sensing and actuation in hybrid robotic systems. Improve additive manufacturing fabrication to design rigid and semi-rigid components for hybrid robotic platforms. Transformable Wheel Robotic Platform: The chassis, wheels, tires and inspection system housing use different additive manufacturing techniques for fabrication. Continued work with additive manufacturing has lead to studies in metal-based printing and modular design and manufacturing. The new platform design with integrated electrical component printed. This will allow integration of the sensor housing onto the platform. Electrical components are being tested for battery life and performance. To improve this performance, such as integration of Lithium Polymer (LiPo) batteries. Snake Robotic Platform: The main focus of the development has centered around liquid-based soft actuators. That act on the principles of electrostatic and hydraulic actuation. A liquid dielectric sits between two compliant electrodes, contained by a flexible polymer shell. The electrodes and film gradually collapse toward each other from one corner of the electrode to the other. When the electrodes and film close together, a majority of the fluid is pushed into the area not covered by an electrode. A thin layer of the liquid dielectric remains between the electrode. The actuators will be stacked to cause large displacement, and move the linkages. The chassis of this platform uses purely additively manufactured linkages. Intelligent, Preventive IR Inspection System: Development of the AI for the system has lead to using a Scikit-Learn which assists in creating predictive models based on Regression, clustering, classification etc. To improve the infrared thermometry for low emissivity sources, work on the fabrication of a tandem photoconductive infrared thermometer was a main focus. Distance-Voltage-Temperature response data has been collected in the range of 7 cm - 100 cm and 200-400 deg. C. Modifications were made to the existing test bench to have a better control over the data. Automated data collection was realized using a Python code and an Arduino controlled stepper motor to increase the sample rate. A protective enclosure has been built around the setup to minimize the effect of the environment on the measurements. To better predict temperature, different AI models are being optimized. The regression model, LARS showed a high accuracy but had convergence issues and only works for the current test set-up. Results: The Transformable Wheel Robot has developed into a more flexible and modular platform. With the improvements to the current work, effort on the tire or soft gripper design been a large focus. The soft grippers will be interchange able to allow for increased performance in identified terrain types. The development of the liquid-based actuators allows the snake robotic platform to achieve the goals of being flexible and able to navigate confined spaces. However, there is room for improvement. Optimization work is currently being done in COMSOL Multiphysics. With the current IR system set-up the LARS model perfectly predicts the data; however, considering the mobility aspect of the project other models will allow for an optimized system. Testing of other model types is currently being done.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Radiation Detection Computational Benchmark Scenarios

Modeling forms an important component of radiation detection development, allowing for testing of new detector designs, evaluation of existing equipment against a wide variety of potential threat sources, and assessing operation performance of radiation detection systems. This can, however, result in large and complex scenarios which are time consuming to model. A variety of approaches to radiation transport modeling exist with complementary strengths and weaknesses for different problems. This variety of approaches, and the development of promising new tools (such as ORNL’s ADVANTG) which combine benefits of multiple approaches, illustrates the need for a means of evaluating or comparing different techniques for radiation detection problems. This report presents a set of 9 benchmark problems for comparing different types of radiation transport calculations, identifying appropriate tools for classes of problems, and testing and guiding the development of new methods. The benchmarks were drawn primarily from existing or previous calculations with a preference for scenarios which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22. From a technical perspective, the benchmarks were chosen to span a range of difficulty and to include gamma transport, neutron transport, or both and represent different important physical processes and a range of sensitivity to angular or energy fidelity. Following benchmark identification, existing information about geometry, measurements, and previous calculations were assembled. Monte Carlo results (MCNP decks) were reviewed or created and re-run in order to attain accurate computational times and to verify agreement with experimental data, when present. Benchmark information was then conveyed to ORNL in order to guide testing and development of hybrid calculations. The results of those ADVANTG calculations were then sent to PNNL for compilation. This is a report describing the details of the selected Benchmarks and results from various transport codes.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

Active Interrogation using Photofission Technique for Nuclear Materials Control and Accountability

Innovative systems with increased sensitivity and resolution are in great demand to detect diversion and to prevent misuse in support of nuclear materials management for the U.S. fuel cycle. Nuclear fission is the most important multiplicative process involved in non-destructive active interrogation. This process produces the most easily recognizable signature for nuclear materials. In addition to thermal or high-energy neutrons, high-energy gamma rays can also excite a nucleus and cause fission through a process known as photofission. Electron linear accelerators (linacs) are widely used as the interrogating photon sources for inspection methods involving photofission technique. After photofission reactions, prompt signals are much stronger than the delayed signals, but it is difficult to quantify them in practical measurements. Delayed signals are easily distinguishable from the interrogating radiation. Linac-based, advanced inspection techniques utilizing the delayed signals after photofission have been extensively studied for homeland security applications. Previous research also showed that a unique delayed gamma ray energy spectrum exists for each fissionable isotope. In this work, high-energy delayed γ-rays were demonstrated to be signatures for detection, identification, and quantification of special nuclear materials. Such γ-rays were measured in between linac pulses using independent data acquisition systems. A list-mode system was developed to measure low-energy delayed γ-rays after irradiation. Photofission product yields of 238 U and 239 Pu were determined based on the measured delayed γ-ray spectra. The differential yields of delayed γ-rays were also proven to be able to discriminate nuclear from non-nuclear materials. The measurement outcomes were compared with Monte Carlo simulation results. It was demonstrated that the current available codes have capabilities and limitations in the simulation of photofission process. A two-fold approach was used to address the high-rate challenge in used nuclear fuel assay based on photofission technique. First, a standard HPGe preamplifier was modified to improve its capabilities in high-rate pulsed photofission environment. Second, advanced pulse processing algorithms were shown to greatly improve throughput rate without large sacrifice in energy resolution at ultra-high input count rate. Two customized gamma spectroscopy systems were also developed in real-time on FPGAs. They were shown to have promising performance matching available commercial units.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2017

Validation and uncertainty quantification of detector response functions for a $1"×2"$ NaI collimated detector intended for inverse radioisotope source mapping applications

Detector response functions (DRFs) are often used for inverse analysis. We compute the DRF of a sodium iodide (NaI) nuclear material holdup field detector using the code named g03 developed by the Center for Engineering Applications of Radioisotopes (CEAR) at NC State University. Three measurement campaigns were performed in order to validate the DRF’s constructed by g03: on-axis detection of calibration sources, off-axis measurements of a highly enriched uranium (HEU) disk, and on-axis measurements of the HEU disk with steel plates inserted between the source and the detector to provide attenuation. Furthermore, this work quantifies the uncertainty of the Monte Carlo simulations used in and with g03, as well as the uncertainties associated with each semi-empirical model employed in the full DRF representation. Overall, for the calibration source measurements, the response computed by the DRF for the prediction of the full-energy peak region of responses was good, i.e. within two standard deviations of the experimental response. In contrast, the DRF tended to overestimate the Compton continuum by about 45–65% due to inadequate tuning of the electron range multiplier fit variable that empirically represents physics associated with electron transport that is not modeled explicitly in g03. For the HEU disk measurements, computed DRF responses tended to significantly underestimate (more than 20%) the secondary full-energy peaks (any peak of lower energy than the highest-energy peak computed) due to scattering in the detector collimator and aluminum can, which is not included in the g03 model. We ran a sufficiently large number of histories to ensure for all of the Monte Carlo simulations that the statistical uncertainties were lower than their experimental counterpart’s Poisson uncertainties. The uncertainties associated with least-squares fits to the experimental data tended to have parameter relative standard deviations lower than the peak channel relative standard deviation in most cases and good reduced chi-square values. The highest sources of uncertainty were identified as the energy calibration polynomial factor (due to limited source availability and NaI resolution) and the Ba-133 peak fit (only a very weak source was available), which were 20% and 10%, respectively.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2018

sUAS Support for Mars 2020

The release of airborne radioactive material presents a health risk hazard to many individuals, including emergency responders and the public. Characterizing the unknown radioactive dangers produced during such incidents is essential. An advantage to utilizing unmanned aerial systems in this effort is the reduction of risk to personnel otherwise tasked with characterization, as well as more quickly obtaining data points during cloud immersion. By pairing a continuous air monitor (CAM) and detector with a drone, quantification of unknown material in the plume can be determined, and thus the extent of dose to people in the area can be estimated. A small, low-flow air sampler with a Geiger-Mueller counter may be characterized using experimental measurements of point sources representing a cloud of material and modelled using the Monte Carlo N-Particle Transport Code (MCNP), a software package for simulating nuclear processes, to find the usefulness and advantages of the system for detection of nuclides of concern in various nuclear plume scenarios. The flying time of small-unmanned aircraft systems (sUASs) is limited; thus, the minimum detectable activity and concentration rates possible must be determined over the system’s window of operation. While the activity concentrations corresponding to external and internal doses of concern can be detected with certainty for 90 Sr and 137 Cs in a plume, those corresponding to 238 Pu are more difficult to detect. The central principle of employing the sUAS is that the cockpit does not have an operator onboard the aircraft; therefore, control of the aircraft must take place by other means. With a sUAS, command and control can be separated into three distinct methods: (1) ground control or remote piloting, (2) semi-autonomous control, and (3) autonomous control. The isotope 238 Pu is used in the space industry, providing long-lived heat generation for fuel in outer space (NASA 2014). Due to the inherent risks with rocket launches to get objects into space, monitoring for release of these hazardous substances is a major undertaking. The current means is through employment of environmental CAMs. It is conceivable, however, that drone monitoring for release of 238 Pu in the event of a space launch mishap would give a greater chance of detection of plutonium particulate, because the particulate’s diffusion in air upon settling to the ground would make the probability of collecting a particle on a filter in a CAM very low.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Image processing analysis of nuclear track parameters for CR-39 detector irradiated by thermal neutron

CR-39 detector which covered with boric acid (H{sub 3}Bo{sub 3}) pellet was irradiated by thermal neutrons from ({sup 241}Am - {sup 9}Be) source with activity 12Ci and neutron flux 10{sup 5} n. cm{sup −2}. s{sup −1}. The irradiation times -T{sub D} for detector were 4h, 8h, 16h and 24h. Chemical etching solution for detector was sodium hydroxide NaOH, 6.25N with 45 min etching time and 60 C° temperature. Images of CR-39 detector after chemical etching were taken from digital camera which connected from optical microscope. MATLAB software version 7.0 was used to image processing. The outputs of image processing of MATLAB software were analyzed and found the following relationships: (a) The irradiation time -T{sub D} has behavior linear relationships with following nuclear track parameters: i) total track number - N{sub T} ii) maximum track number - MRD (relative to track diameter - D{sub T}) at response region range 2.5 µm to 4 µm iii) maximum track number - M{sub D} (without depending on track diameter - D{sub T}). (b) The irradiation time -T{sub D} has behavior logarithmic relationship with maximum track number - M{sub A} (without depending on track area - A{sub T}). The image processing technique principally track diameter - D{sub T} can be take into account to classification of α-particle emitters, In addition to the contribution of these technique in preparation of nano- filters and nano-membrane in nanotechnology fields.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2014

Novel Deployment of Elpasolites as a Dual Neutron/Gamma-Ray Directional Detector

At a time when upholding national security has never been more important, there exists a need for the advancement of radiation detection technologies. Neutron and photon detectors are essential to fulfilling mission areas including detection and localization of missing, stolen or smuggled radiological or nuclear materials, quantification of the effects of a radiological or nuclear event, and supporting nonproliferation efforts. The aim of this study was to evaluate a new radiation detector based on the scintillation elpasolite compound Cs 2 LiYCl 6 :Ce (CLYC) for simultaneous measurements of neutron and photon flux and the localization of radiation sources. Previous studies performed on the CLYC scintillator indicate its potential for thermal neutron and gamma-ray measurements. This study is dedicated to the novel application of the CLYC as a dual neutron / photon detector and as part of a directional detection system. Both computational modeling and an experimental study were carried out within this research project. As part of the computational study, the response of a CLYC scintillator detector to gamma rays induced by thermal neutron interaction with Cl and 7 Li nuclei was investigated using the MCNP6 code. In addition, arrays of three and four CLYC detectors were modeled in order to evaluate the directional detection of both a thermal neutron source and a gamma-ray source. It was shown that little or no quality of source direction determination would be lost when three detectors were used in the array compared to four detectors. In the experimental study, the photon spectroscopy capabilities of the CLYC detectors were evaluated. A gamma-ray energy resolution of 4.9% was measured for the 662-keV peak of 137 Cs and 3.6% for the 1.33-MeV peak of 60 Co. Using a thermal neutron source, the pulse shape discrimination analysis was successfully performed for the CLYC detector signal waveforms. Thermal neutrons and gamma rays were separated with an exceptional figure of merit (FOM) of 2.3. An array of three CLYC detectors was assembled for the purpose of directional neutron / gamma-ray detection. The intrinsic peak efficiency of CLYC detectors was evaluated. The three-CLYC detector array was deployed for directional measurements with a single gamma-ray 137 Cs source, two gamma-ray sources of 137 Cs and 60 Co isotopes and a thermal neutron source designed using a 239PuBe neutron source supplied with a polyethylene moderator. Measurements were carried out using sources located in the longitude and latitude planes over the angles from 0° to 360°. The measured data were processed through a maximum likelihood estimation algorithm providing a possible direction for which the radioactive source in each case was positioned. The estimated directions were close if not exact matches for the actual directions to the radioactive source. The largest discrepancy in direction produced by the algorithm was approximately 11%. However, it was hypothesized that this percent error can be decreased by homogenizing the directional detection system to consist of scintillators of the same size and quality, identical photomultiplier tubes and identical aluminum housings. The feasibility of this hypothesis to decrease the percent error was confirmed by the zero percent error achieved in the directional measurements produced in the computational study utilizing a homogenous directional detection system. The results of computational and experimental studies completed within this research project provide means to propose the array of three CLYC scintillators as an efficient dual neutron / gamma-ray directional detector.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

SU-E-T-107: An Investigation Into Attenuation Effects On the Energy Spectrum for {sup 137}Cs Using Monte Carlo Techniques

Purpose: The calibration of radiation protection instrumentation including ionization chambers, scintillators, and Geiger Mueller (GM) counters used as survey meters are often done using {sup 137}Cs irradiators. During calibration, irradiators use a combination of attenuators with various thicknesses to modulate the beam to a known air-kerma rate. The variations in energy spectra as a result of these attenuators are not accounted for and may play a role in the energy-dependent response of survey meters. This study uses an experimentally validated irradiator geometry modeled in the MCNP5 transport code to characterize the effects of attenuation on the energy spectrum. Methods: A Hopewell Designs G-10 {sup 137}Cs irradiator with lead attenuators of thicknesses of 0.635, 1.22, 2.22, and 4.32 cm, was used in this study. The irradiator geometry was modeled in MCNP5 and validated by comparing measured and simulated percent depth dose (PDD) and cross-field profiles. Variations in MCNP5 simulated spectra with increasing amounts of attenuation were characterized using the relative intensity of the 662 keV peak and the average energy. Results: Simulated and measured PDDs and profiles agreed within the associated uncertainty. The relative intensity of the 662 keV peak for simulated spectra normalized to the intensity of the unattenuated spectra ranged from 0.16% to 100% based on attenuation thickness. The average energy for simulated spectra for attenuators ranged from 582 keV with no attenuation to 653 keV with 5.54 cm of attenuation. Statistical uncertainty for MCNP5 simulations ranged from 0.11% to 3.69%. Conclusion: This study successfully used MCNP5 to validate a {sup 137}Cs irradiator geometry and characterize variations in energy spectra between different amounts of attenuation. Variations in the average energy of up to 12% were determined through simulations, and future work will aim to determine the effects of these differences on survey meter response and calibration.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI1995

Nuclear Safety [Vol. 36, No. 2, July-December 1995]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. THE CHORNOBYL ACCIDENT: 195 The Chornobyl Accident Revisited, Part III: Chernobyl Source Term Release Dynamics and Reconstruction of Events During the Active Phase, A. R. Sich; GENERAL SAFETY CONSIDERATIONS: 218 Second ANS Workshop on the Safety of Soviet-Designed Nuclear Power Plants, R. A. Bari; 234 Elements of a Nuclear Criticality Safety Program, C. M. Hopper; 243 Rickover, Excellence, and Criticality Safety Programs, R. E. Wilson; ACCIDENT ANALYSIS: 249 Transient Analysis of the PIUS Advanced Reactor Design with the TRAC-PF1/MOD2 Code, B. E. Boyack, J. L Steiner, S. C. Harmony, H. J. Stumpf, and J. F. Lime; 278 The Hierarchy-By-Interval Approach to Identifying Important Models that Need Improvement in Severe-Accident Simulation Codes, T. J. Heames, M. Khatib-Rahbar, J. E. Kelly, R. P. Jenks-Johnson, and Y.-S. Chen; 290 RELAP5/MOD3 Code Coupling Model, R. P. Martin; 299 Missiles Caused by Severe Pressurized-Water Reactor Accidents, R. Krieg; DESIGN FEATURES: 310 Validation of COMMIX with Westinghouse AP-600 PCCS Test Data, J. G. Sun, T. H. Chien, J. Ding, and W T. Sha; ENVIRONMENTAL EFFECTS: 321 Spent Nuclear Fuel Characterization for a Bounding Reference Assembly for the Receiving Basin for Off-Site Fuel, S. D. Kahook, R. L. Garrett, L. R. Canas, and M J. Beckum; OPERATING EXPERIENCES: 332 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 335 Reactor Coolant System Blowdown at Wolf Creek on September 17, 1994, J. V. Kauffman and S. L. Israel; RECENT DEVELOPMENTS: 344 Reports, Standards, and Safety Guides, D. S. Oueener; 349 Proposed Rule Changes as of June 30, 1995; ANNOUNCEMENTS: 320 Symposium on Acceptability of Risk From Radiation—Application to Manned Space Flight; 320 24th DOE/NRC Nuclear Air Cleaning and Treatment Conference; 361 Radiation Biology and Radiation Protection— Modern Developments and Tendencies in Radiation Biology; 362 1997 IEEE Sixth Conference on Human Factors and Power Plants; 354 The Authors; 360 Reviewers of Nuclear Safety, Vol. 36.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Transport Test Problems for Radiation Detection Scenarios

Modeling forms an important component of radiation detection development, allowing for testing of new detector designs, evaluation of existing equipment against a wide variety of potential threat sources, and assessing operation performance of radiation detection systems. This can, however, result in large and complex scenarios which are time consuming to model. A variety of approaches to radiation transport modeling exist with complementary strengths and weaknesses for different problems. This variety of approaches, and the development of promising new tools (such as ORNL’s ADVANTG) which combine benefits of multiple approaches, illustrates the need for a means of evaluating or comparing different techniques for radiation detection problems. This report presents a set of 9 test problems for comparing different types of radiation transport calculations, identifying appropriate tools for classes of problems, and testing and guiding the development of new methods. Cases are drawn primarily from existing or previous calculations with a preference for cases which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22. From a technical perspective, cases are chosen to span a range of difficulty and to include gamma transport, neutron transport, or both and represent different important physical processes and a range of sensitivity to angular or energy fidelity. Following test case identification, existing information about geometry, measurements, and previous calculations were assembled. Monte Carlo results (MCNP decks) were reviewed or created and re-run in order to attain accurate computational times and to verify agreement with experimental data, when present. Test case information was then conveyed to ORNL in order to guide testing and development of hybrid calculations. The results of those ADVANTG calculations were then sent to PNNL for compilation. This is a report describing the details of the selected test cases and results from available transport codes. Additional results from other transport codes will be added during the next year, along with additional cases if needed for software testing. Future work will extend the test problems to include more extensive cross-code comparisons as well as more refined metrics for evaluating the tradeoffs between operator time, computational time, and solution accuracy, to assemble a formalized benchmark set.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI1995

Nuclear Safety [Vol. 36, No. 2, July-December 1995]

05 NUCLEAR FUELS↗

DOE OSTI2015

Collaborative development of the EPICS Qt framework Phase I Final Report

At Lyncean, a private company spun-off from technology developed at the SLAC National Lab, we have been using EPICS for over a decade. EPICS is ubiquitous on our flagship product – the Compact Light Source. EPICS is not only used to control our laser and accelerator systems, but also to control our x-ray beamlines. The goal of this SBIR is for Lyncean Technologies to spearhead a worldwide collaborative effort for the development of control system tools for EPICS using the Qt framework, a C++-based coding environment that could serve as a competitive alternative to the Java-based Control System Studio (CSS). This grant's Phase I, not unlike a feasibility study, is designed for planning and scoping the preparatory work needed for Phase II or other funding opportunities. The three main objectives of this Phase I are (1) to become better acquainted with the existing EPICS Qt software and Qt framework in order to evaluate the best options for ongoing development, (2) to demonstrate that our engineers can lead the EPICS community and jump-start the Qt collaboration, and (3) to identify a scope for our future work with solicited feedback from the EPICS community. This Phase I report includes key technical findings. It clarifies the differences between the two apparently-competing EPICS Qt implementations, caQtDM and the QE Framework; it explains how to create python-bindings, and compares Qt graphical libraries. But this report is also a personal story that narrates the birth of a collaboration. Starting a collaboration is not the work of a single individual, but the work of many. Therefore this report is also an attempt to publicly give credit to many who supported the effort. The main take-away from this grant is the successful birth of an EPICS Qt collaboration, seeded with existing software from the PSI and the Australian Synchrotron. But a lot more needs to be done for the collaboration founders' vision to be realized, and for the collaboration to reach its full potential. To help define the scope of future work, a useful approach we have identified is user experience design (UXD) and is discussed herein.

43 PARTICLE ACCELERATORS↗