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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Quantification of LEU Holdup using gamma ray imaging and inverse transport solver

Holdup is the residual amount of special nuclear material (SNM) remaining in a processing facility after the bulk materials have been cleaned out. In commercial uranium processing facilities, quantification of holdup is a major challenge because of the highly variable shapes and sizes of the deposits. Any method that attempts to generalize and calibrate deposit shapes in order to quantify holdup will be prone to high uncertainties. Uncertainties on the order of ±50% are typical in holdup results. In international safeguards applications, a ±50% uncertainty can result in a large amount of material unaccounted for (MUF) thereby increasing the difficulty of detecting material diversion and facility misuse. An imaging-based methodology has been developed with the objective of significantly reducing this uncertainty by using the true deposit shape, instead of relying on oversimplified geometric assumptions. The project is a collaboration between ORNL, Y-12, and the University of Tennessee, Knoxville, TN. Uranium sources of known masses were measured using the Germanium Gamma-ray Imager (GeGI), a high-resolution imaging spectrometer, creating a pixelated map for each spectral bin. Two different gamma imaging methods are employed in this work: coded aperture imaging and Compton imaging. A validated MonteCarlo model of the detector has been developed using the GEANT4 code for determining the intrinsic response of the detector, its enclosure, and the coded aperture mask. An inverse transport solver based on the Markov Chain Monte-Carlo approach known as Differential Evolution Adaptive Metropolis (DREAM) is employed to use the measurement data from the image pixels (coded aperture or Compton) to solve for the mass of 235 U in the deposit. A reliable method based on the DREAM solver has been developed to flag the infinite thickness condition of a uranium deposit. The project team is working towards improving the image reconstruction for Compton imaging so that a better localization of the source can be achieved. Besides treating the coded aperture and Compton imaging methods independently, the project is also evaluating a combined method that uses the Compton scatter data from a coded aperture measurement. GEANT4 simulations are being performed to evaluate the combined approach. The impact on the DREAM optimization as the source thickness progressively approaches infinite thickness is being evaluated. A number of uranium sources available at ORNL have been measured, and the DREAM results have been tested and validated for the coded aperture imaging. A similar effort will be carried out to validate the Compton based method once the development of algorithms for better localization are complete. The imaging based quantification is very amenable to unattended monitoring of holdup accumulation at key measurement points. A proof of concept measurement has been completed to demonstrate this capability The current work used the high energy resolution imager GeGI. However, the approach and methodologies are applicable to other imagers such as the cadmium zin telluride (CZT) based imager manufactured by H3D, Inc.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Baseline Hypothetical Facility for the Production of 131 I and 99 Mo using Activation Targets

This report describes a hypothetical facility for production of medical radioisotopes via activation under the Proliferation Resistance and Optimization (PRO-X) program. The facility uses neutron activation of non-special nuclear material (SNM) to produce the medical isotopes 131 I and 99 Mo at a throughput of 60 Ci/week of 131 I and 5 Ci/week of 99 Mo. The hypothetical design was carried out using a 10 MWt research reactor. The precursors used for the activation process were TeO2 for 131 I and MoO 3 for 99 Mo. The processes are performed in 3 hot cells used for target receipt, extraction, purification low specific activity (LSA) generator introduction, and packaging. A fourth hotcell is used for waste processing. The hot cell processing area takes up a footprint of 15.4 m 2 with the total footprint of the facility, including space for administrative offices, non-rad labs, quality assurance, and radiation buffer areas set at 763 m 2 . Waste is produced at a weekly rate of 257.8 g low activity solid waste and 8032.7 mL of low activity liquid waste, 8032 mL of which is water. This baseline hypothetical facility for production of medical isotopes via activation was then compared and contrasted to the hypothetical facility for production of medical isotopes via fission products to show the differences in approach for the two production modes. The two production modes had several highlighted differences including the overall facility and hot cell layout, the type and amount of waste produced by the respective facilities, and economic factors impacting production mode. Finally, a decision tree for which production mode might be more beneficial for an entrant into medical isotope production was developed based on the differences examined and the desired output of medical isotopes desired by the entrant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

A probabilistic inverse prediction method for predicting plutonium processing conditions

In the past decade, nuclear chemists and physicists have been conducting studies to investigate the signatures associated with the production of special nuclear material (SNM). In particular, these studies aim to determine how various processing parameters impact the physical, chemical, and morphological properties of the resulting special nuclear material. By better understanding how these properties relate to the processing parameters, scientists can better contribute to nuclear forensics investigations by quantifying their results and ultimately shortening the forensic timeline. This paper aims to statistically analyze and quantify the relationships that exist between the processing conditions used in these experiments and the various properties of the nuclear end-product by invoking inverse methods. In particular, these methods make use of Bayesian Adaptive Spline Surface models in conjunction with Bayesian model calibration techniques to probabilistically determine processing conditions as an inverse function of morphological characteristics. Not only does the model presented in this paper allow for providing point estimates of a sample of special nuclear material, but it also incorporates uncertainty into these predictions. This model proves sufficient for predicting processing conditions within a standard deviation of the observed processing conditions, on average, provides a solid foundation for future work in predicting processing conditions of particles of special nuclear material using only their observed morphological characteristics, and is generalizable to the field of chemometrics for applicability across different materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Investigation of the thermal decomposition of Pu(IV) oxalate: a transmission electron microscopy study

The degradation of the internal structure of plutonium (IV) oxalate during calcination was investigated with Transmission Electron Microscopy (TEM), electron diffraction, Electron Energy-Loss Spectroscopy (EELS), and 4D Scanning TEM (STEM). TEM lift-outs were prepared from samples that had been calcined at 300°C, 450°C, 650°C and 950°C. The resulting phase at all calcination temperatures was identified as PuO 2 with electron diffraction. The grain size range was obtained with high-resolution TEM. In addition, 4D STEM images were analyzed to provide grain size distributions. In the 300°C calcined sample, the grains were <10 nm in diameter, at 650°C, the grains ranged from 10 to 20 nm, and by 950°C, the grains were 95–175 nm across. Using the Kolmogorov-Smirnov (K-S) two sample test, it was shown that morphological measurements obtained from 4D-STEM provided statistically significant distributions to distinguish samples at the different calcination conditions. Using STEM-EELS, carbon was shown to be present in the low temperature calcined samples associated with oxalate but had formed carbon (possibly graphite) deposits in the 950°C calcined sample. This work highlights the new methods of STEM-EELS and 4D-STEM for studying the internal structure of special nuclear materials (SNM).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1985

Samarkand complex setup for investigation of cosmic ray variation in the energy range of 7 10 (9) - 10 (15) eV

The Samarkand complex setup is aimed at the study of cosmic ray variations in a wide energy range from 7 billion eV (which corresponds to the geomagnetic threshold in the region of Samarkand) up to approx 10 to the 15th power to 10 to the 16th power eV. The setup consists of four 6-counter sections of neutron supermonitor with counters SNM-15 and 48 scintillator detectors (1 sq m each) placed under and above the supermonitor. The effective area of the setup for recording neutrons and muons is 24 sq m. The setup can register time variations of the following cosmic ray components: (1) the total neutron counting rate, (2) counting rates for neutrons of different multiplicity, (3) soft-muon fluxes, (4) hard-muon fluxes at various zenith and azimuth angles, (5) electron-photon component, (6) extensive air showers (EAS) induced by primary particles in a wide energy range and accompanied or not accompanied by muons and neutrons.

Dorman, L. I.↗
DOE OSTI2018

Application of a Triple Bubbler Sensor for Determining the Density, Surface Tension, and Depth in Molten Salts

Electrochemical (i.e. pyroprocessing) technology is being widely studied throughout the world as a potential alternative to aqueous reprocessing of spent nuclear fuel (SNF). Central to this process is the electrorefiner (ER), in which the useful uranium in the SNF is electrochemically transported through a molten salt electrolyte to a cathode for later recovery. As part of this process, uranium, plutonium, and other actinides accumulate in the ER salt over time. Material accountancy and safeguards of these special nuclear material buildups in near real-time is a significant challenge due to the elevated processing temperatures, remote operation within a hotcell, and high radiation fields. Analytical techniques outside of the hotcell can be used to provide the SNM concentrations of salt samples from the ER. However, determining the actual mass of these materials in the ER presents a significant challenge because the density and volume within the vessel is largely unknown. The Idaho National Laboratory (INL) has developed a triple bubbler sensor capable of determining the density, surface tension, and depth of a fluid in near real-time. Previous experiments in aqueous media has shown that the sensor is capable of determining the density, surface tension and depth with accuracies on the order of 0.2%. The goal of this current work is to test the sensor in several molten salt media to calibrate and validate the triple bubbler. Experiments have been performed in LiCl-KCl eutectic salt. Calibration factors were 1.0015 and 1.26 for density and surface tension, respectively. With the above calibration factors applied, the depth of the salt in the test vessel was determined to within 0.4%. These results indicate that the triple bubbler sensor has the potential to accurately (less than 1% uncertainty) determine the density, surface, tension, and depth of the ER salt in near real-time. The triple bubbler sensor designed at INL can significantly enhance the material accountancy, safeguards, and process monitoring of the electrochemical processing of SNF.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

An automated approach to the alignment of compound refractive lenses

Compound refractive lenses (CRLs) are established X-ray focusing optics, and are thus a key element of many beamlines at X-ray facilities, used to either to focus a beam or image a sample. While CRLs are quite established, the stack of single lens elements affords a very small numerical aperture because of the thick lens profile, making them far more difficult to align than classical optical lenses that obey the thin-lens approximation. This means that the alignment must be very precise and is highly sensitive to changes to the incident beam; aligning CRLs is thus typically treated as a time-consuming manual process. Many groups circumvent day-to-day realignment by using engineering controls (e.g. mounting optics) that sacrifice some of the beam’s focusing characteristics, i.e. spot size, or resolution. While these choices minimize setup time, there are clear disadvantages. This work presents a new automated approach to align CRLs using a simple alignment apparatus that is easy to adapt and install at different types of X-ray experiments or facilities. Our approach builds on recent CRL modeling efforts, using an approach based on the Stochastic Nelder-Mead (SNM) simplex method. We demonstrate the efficacy of this method by numerical simulation and then extend this test to real experiments conducted at the Advanced Photon Source to confirm its performance with a synchrotron X-ray beam. This work provides an opportunity to automate key instrumentation at X-ray facilities.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Cumulative short-lived photofission product yields for nuclear forensic application

Experiments have long been carried out with linear accelerator (linac) produced bremsstrahlung X-rays to study high-energy photon-induced fission, known as “photofission”. One benefit of photofission, when compared with neutron-induced fission, is the ability to investigate fundamental physics of fission at excitation energies lower than what is possible with neutrons [1]. Absorption of an incident neutron results in a compound nucleus with an excitation energy at least equal to the neutron binding energy; thus, the lowest energy region of the fission barrier cannot be studied. Fission product yield distributions are dependent on the incident particle energy and the target nucleus mass. They are generally represented by a double-humped curve, with low and high mass “peaks” and an intermediate mass “valley”. As incident particle energy increases, the excitation energy of the nucleus increases, thereby increasing the overall yield of isotopes in the valley with respect to the two maxima or peaks of the fission product curve [3]. Photofission is most probable for incident photons around 14 MeV due to the giant dipole resonance. Induced photofission around this energy results in a fission product yield distribution that resembles 14 MeV DT neutron fission more than thermal or fast (500 keV) neutron fission. In post-detonation nuclear forensics, the peak-to-valley ratio of the fission product yield curve could be used for characterizing device type by providing an indication of the neutron energy spectrum that had induced fission (e.g., 500 keV vs. 14 MeV). The goal of this work is to improve nuclear data for isotopes of interest in nuclear forensics and special nuclear material (SNM) detection, specifically the fission product yield data via photofission experiments for the isotopes of 238U and 232Th. Photofission has been proposed as a less costly and more flexible production method for the valley isotopes that are representative of DT neutron fission product yield distributions, which would benefit nuclear forensics exercises in testing for analysis techniques. Natural uranium and thorium targets were irradiated with bremsstrahlung photons (nominal endpoint energies of 8, 14, and 20 MeV), generated via impingement of high-energy electrons on a tungsten electron-to-photon radiator. Gamma-ray spectroscopy was used to observe and identify the decay of the resulting fission products. The photofission yields of 102 fission products were measured for the set of the 238U and 232Th targets at the three bremsstrahlung X-ray endpoint energies 8, 14, and 20 MeV. These fragments included masses ranging from A=84 to A=144, with half-lives as short as 1.07 seconds.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Integration of Nuclear Material Accounting Data and Process Monitoring Data for Improvement on Detection Probability in Safeguarding Electrochemical Processing Facilities (Final Technical Report)

The KAERI advanced spent fuel conditioning process (ACP) process is a critical component of the US- South Korean nuclear cooperation and the following “123 Agreement.” Its development has received considerable attention in both countries. The ACP is an electrochemical processing (pyroprocessing) that recycles over 96% of the used nuclear fuel (UNF). It is also intrinsically proliferation-resistant in theory. In normal operation, the U/TRU product is very hot radiologically. In addition, the Cm provides a high level of spontaneous neutrons, making the product unsuitable for weapon use. However, as pointed in some study, “the need for safeguards to protect against the diversion and misuse of separated plutonium applies essentially equally to all grades of plutonium.” As pointed by many studies, the well-established traditional Nuclear Material Accounting (NMA) approach cannot be directly applied to electrochemical processing because of the lack of an input accountability tank, the non-continuous material flow, and the unsatisfactory level of confidence in sampling methods. Therefore, nuclear safeguards remain a grand challenge in the developing of commercial electrochemical separations facilities, especially around the heart of such facilities, the electrorefiner (ER) systems. In contrast to NMA data, process monitoring (PM) data is normally an indirect measurement of the SNM and is acquired much more frequently. In a broad sense, PM includes monitoring by various types of equipment, e.g. radiation detectors, cameras, voltage, current sensors. Because it is already being collected by the operator, the additional cost to safeguards is low. It has long been believed that PM data can supplement NMA data and help improve safeguards, although the benefits are hard to quantify. The U.S. DOE’s Material Protection, Accounting, and Control Technology (MPACT) campaign has made substantial investments into innovative PM sensor technology and predictive model development for real- or near real-time measurement and prediction of molten salt density and level, salt composition and actinide concentration especially Pu, the cell voltage, and the cell current to supplement traditional NMA. For aqueous-based reprocessing facilities, it is reported that PM, integrated with traditional NMA, have a high detection probability for specific diversions. For electrochemical reprocessing, preliminary studies have shown that PM data can support traditional NMA in various ways by providing a basis to estimate some of the in-processing nuclear material inventories. Despite early success, further studies on fusion of PM data and NMA data are still needed, which is the goal of this proposed work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Performance of Large Neutron Detectors Containing Lithium-Gadolinium-Borate Scintillator

This paper describes the development and testing of a neutron counter, spectrometer, and dosimeter that is compact, efficient, and accurate. A self-contained neutron detection instrument has wide applications in health physics, scientific research, and programs to detect, monitor, and control strategic nuclear materials (SNM). The 1.3 liter detector head for this instrument is a composite detector with an organic scintillator containing uniformly distributed {sup 6}Li{sub 6}{sup nat}Gd{sup 10}B{sub 3}O{sub 9}:Ce (LGB:Ce) microcrystals. The plastic scintillator acts to slow impinging neutrons and emits light proportional to the energy lost by the neutrons as they moderate in the detector body. Moderating neutrons that have slowed sufficiently capture in one of the Lithium-6, Boron-10, or Gadolinium-157 atoms in the LGB:Ce scintillator, which then releases the capture energy in a characteristic cerium emission pulse. The measured captured pulses indicate the presence of neutrons. When a scintillating fluor is present in the plastic, the light pulse resulting from the neutron moderating in the plastic is paired with the LGB:Ce capture pulse to identify the energy of the neutron. About 2% of the impinging neutrons lose all of their energy in a single collision with the detector. There is a linear relationship between the pulse areas of this group of neutrons and energy. The other 98% of neutrons have a wide range of collision histories within the detector body. When these neutrons are 'binned' into energy groups, each group contains a distribution of pulse areas. This data was used to assist in the unfolding of the neutron spectra. The unfolded spectra were then validated with known spectra, at both neutron emitting isotopes and fission/accelerator facilities. Having validated spectra, the dose equivalent and dose rate are determined by applying standard, regulatory damage coefficients to the measured neutron counts for each energy bin of the spectra. Testing at the Los Alamos Neutron Science Center (LANSCE), Edwards Accelerator Laboratory (EAL) at Ohio University and the Radiation Center at University of Massachusetts-Lowell has demonstrated that the instrument can measure neutrons and their spectra over the range between 0.8 MeV and 150 MeV with an uncertainty of only ± 8%. An independent test of the LGB:Ce neutron spectrometer was conducted by a US Defense Threat Reduction Agency (DTRA) team at the Idaho National Laboratory (INL). The results of this evaluation showed that the neutron spectrometer accurately identified bare radioactive isotopes by their spectra. Further, masking and shielding materials alter those spectra in predictable ways that permit an extrapolation from the observed spectra back to the identity of the isotopic spectrum. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Performance of an RPM based on Gd-lined plastic scintillator for neutron and gamma detection [ANIMMA--2015-IO-x19]

INFN and ANN designed and developed a Radiation Portal Monitor (RPM) to detect gamma and neutron radiations using plastic scintillator coupled to Gadolinium. The system is a portal for the screening of vehicle and cargo containers aiming at detecting the presence of radioactive elements and, in particular, Special Nuclear Materials (SNM). The system was calibrated based on static measurements at the INFN laboratory and characterized with dynamic tests at the the JRC ITRAP+10 facility. Results showed that the system is not only compliant to international standards of its category, but exceeds them in most of the cases, demonstrating the validity of the novel technology and of the implemented layout. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Operator Evaluation of Mobile Robotic Systems for Inventory, Inspection, and Contamination Survey Tasks in D and D Environments - 17314

The Nuclear and Applied Robotics Group (NRG) has developed two mobile systems with simple, intuitive Human-Machine Interfaces (HMIs). Thus, the systems can be operated by operators who are not robotics experts to perform inventory, inspection and routine contamination surveys in areas where Special Nuclear Material (SNM) is stored and handled. This effort presents feedback collected from EM operators using the two systems at the Portsmouth Gaseous Diffusion Plant in Piketon, OH who used the robotic solutions to perform tasks on site and with minimal training. This approach has the potential to address issues related to ALARA, ergonomics, and reliably completing dull, repetitive tasks while maintaining domain-specific experts in the control loop. The two systems were shipped to - and demonstrated in - a DOE designated area and demonstrated completing contamination testing using a combination of real and surrogate sensors. This paper reviews the initial hardware development, demonstration requirements, and the modifications/improvements made to deploy the systems for this effort. It then summarizes the feedback garnered from EM operators who used the systems at the DOE facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Collated experience of the use of multi-group analysis (MGA) in non-destructive waste assay systems - 15257

The quantification of plutonium is an important aspect of non-destructive assay of Special Nuclear Material (SNM). The calculation of {sup 240}Pu{sub eff} can be done through prior knowledge of the composition of the waste but more often is done through the analysis of a measured plutonium gamma-ray spectrum performed using an isotopic analysis code. In this work, the {sup 240}Pu{sub eff} results from a variety of non-destructive waste assay systems in the UK are presented and compared using MGA v9.63C and v10. The body of the results include real measurement results of a variety of waste materials containing plutonium, the results of reanalysis of these initial waste measurements with changes to certain setup parameters and the results of reanalysis with a later version of MGA (v10). Additionally, measurements of test matrices with known source isotopics are measured and the results analysed and presented here. The use of the 'waste' flag parameter introduced to allow measurement of unusual spectra in v10 is also evaluated. When using MGA as part of the waste assay characterisation process in automatic or batch mode, results have proven to be very dependent on the quality of the spectra measured and on the setup parameters that the code is reliant upon. With a modern version of the code, the likelihood of under-reporting the total plutonium content of a 200 litre/55 gallon drum using PNCC and MGA isotopics in combination is low. Additionally, the latest 'hardened' code will also reduce the number of cases where MGA fails to run and the use of default isotopics results in an over-pessimistic evaluation of the total plutonium mass. (authors)

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

Macro-encapsulation as Manufactured Components for Disposal at the Nevada National Security Site (NNSS) - 16213

Macro-encapsulation as Manufactured is a relatively new terminology that supports the disposal of classified legacy weapon components at the Area 5 Disposal Cell at the Nevada National Security Site (NNSS). These components are composed of hundreds and sometimes thousands of parts. Examples of this type of item include firing sets, trainers, radar assemblies, neutron generators and other similar components. Sometimes there are hazardous Resource Conservation and Recovery Act-regulated (RCRA) metals involved in the parts that make up the components. The typical regulated metals in the components can be lead and silver from circuit boards and wiring, cadmium or chromium from plating operations of the parts. To perform the Toxicity Characteristic Leaching Procedure (TCLP) components have to be completely disassembled and sanitized prior to being sent to laboratories for analysis. The results vary but the majority of times there are regulated metals present in these samples. The levels are relatively low and since analysis has been performed then a disposal process needs to be followed. A process needed to be developed and approved by governing boards to allow disposal of these items as long as this material meets the Waste Acceptance Criteria (WAC) (No liquids, compressed gases, explosives, SNM, etc.) before being sent to the Nevada desert for disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Automated Glovebox Work-cell Design - 18370

Los Alamos National Laboratory (LANL) is the primary plutonium science and manufacturing facility in the United States supporting multiple national security programs. It is responsible for completing a high variety of small-batch manufacturing tasks requiring the handling and precision machining of Special Nuclear Materials (SNM). Given the myriad of tasks that must be completed in a limited workspace, LANL intends to utilize flexible, multi-use robotic work-cells capable of completing a variety of mission critical tasks in a single glovebox with no re-tooling needed between batch jobs. This approach would allow LANL more flexibility in its production schedule and to meet its future production and research objectives. In this paper, we present a system that integrates multiple sensors and other manufacturing peripherals into existing glovebox infrastructure. Currently, a technician can place an object in the work area where vision sensors and supervisory software identify the object and its pose and automatically determine which manufacturing task to complete. Pose uncertainty is addressed in real-time by the task planner, eliminating the need for fixtures. Grasps are validated using pressure sensors, payload sensors, and grasp quality metrics to ensure that objects are not dropped or damaged during manipulation. The robot, sensors, and peripherals are controlled using open-source software built utilizing ROS (Robot Operating System), which minimizes redevelopment and allows for re-use and extensibility. Established capabilities support vision, collision-free path planning, and more, minimizing development time. For this effort, particular attention was given to the integration of general peripheral devices (drill, chuck, etc.), which is critical to efficiently utilizing ROS in manufacturing environments. This integration is achieved with an in-development framework that allows for quick integration of peripherals, including I/O devices and motors. Contact tasks utilize the Generalized Contact Control Framework (GCCF) previously developed at UT Austin. GCCF allows application developers to easily define multiple contact control tasks for a single robot. This paper summarizes the key software components necessary to develop the multi-use work-cell including existing ROS components. UT-developed software for grasp validation, peripheral integration, and contact control is presented in more depth. Finally the glovebox system is demonstrated completing a set of tasks with no fixturing necessary. Primary tasks include surface preparation/ablation, drilling, press-fitting, and sorting, as well as supporting material transfer and inspection tasks. The glovebox work-cell reduces cost, operator dose, ergonomic issues, and time loss due to material transfer between gloveboxes that only support one task. If multiple glovebox work-cells are in use, they have the potential to eliminate bottlenecks and scale production capabilities beyond current limits. (authors)

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

Recent Progress on the PNL Accelerator-Based Intense Fusion Neutron Source

Phoenix Nuclear Labs (PNL) has developed a full line of accelerator-based neutron generators. This paper describes the highest yield DD fusion-based system. This device utilizes a microwave ion source (MWS), 300 kV DC accelerator, magnetic solenoid focus element, differential pumping system, and gaseous deuterium target to achieve DD neutron yields of 3x10{sup 11} n/s. Lower-yield variations of the device have been built using variable voltages and a solid titanium target, and design for a DT version of the gas target system is underway that will increase neutron yield to 5x10{sup 13} n/s. PNL has delivered a number of systems to government and commercial customers and has identified a number of longer-term commercial applications for this high-yield neutron generator. These include medical isotope production, neutron radiography, active interrogation for explosives and SNM detection, and Cf-252 replacement. Most applications require the development of specialized moderator assemblies and fixtures to meet customer requirements. A photograph of the neutron generator prototype is provided in Figure 1. This state-of-the-art neutron generator technology is essential for many applications of critical global importance including neutron radiography, medical isotope production, detection of explosives and nuclear material, materials characterization, and others. The presentation will discuss these applications and recent progress made on the base accelerator technology. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Group Structure Optimization Using the PyGroup Code

Group constants are used as coefficients in the energy domain discretized Linear Boltzmann Equation (LBE), in which the Continuous-Energy (CE) fluxes are also integrated over energy bins and transformed into group fluxes. This procedure, based on the Multi-Group (MG) theory, is widely applied in radiation transport and diffusion simulations Careful preparation of MG cross section data, or group constants, is thus a prerequisite for success in acquiring high-fidelity LBE solutions. Generation of MG cross section data itself is a complicated multi-step task, since group constants are indispensably coupled with the solution of the LBE, i.e. the flux spectrum in the problem to be solved. Theoretically, a MG cross section library is only valid for the problem that it is built upon. Nevertheless, in practice MG cross section libraries are prepared for similar type of problems, such as thermal reactors, fast reactors, etc., to facilitate the neutronic simulations. Generally, CE point-wise cross section data set, generated by both theoretical models and experimental data, is first converted to Fine-Group (FG) libraries using a pre-defined or precalculated flux spectra, then, Broad-Group (BG) libraries are collapsed from the fine-group libraries using problem-dependent flux spectra. Proper treatments of energy and spatial self-shielding and resonance cross section are required in the CE-to-MG converting process Previously, we have developed a computer code, YGROUP, for fine-to-broad group cross section collapsing. To maintain high accuracy of the resulting BG libraries, YGROUP provides users several options on collapsing strategies and weighting schemes for group binning and evaluation of BG cross section data in users' specific problems.. An approach based on Particle Swarm Optimization was applied to automatically determine the optimal group structure. We also developed a hybrid weighting scheme to further improve BG library performance for Special Nuclear Materials (SNM) detection applications. In this work, a new computer code, PyGroup, has been developed to implement the PSO approach with an improved simulation-driven fitness function design that includes both response and forward/adjoint flux errors. Together with YGROUP, and deterministic transport code TITAN/PENMSHXP, the PyGroup code package is tested on a Dual Range Coincident Counter (DRCC) model. In this paper, we demonstrated the capability of the PyGroup code for group structure optimization using the DRCC detector model. In the future, an iteration scheme can be implemented to determine the number of broad groups based on a user-defined tolerance. we will also apply PyGroup to more complicated reactor models, and investigate physics-based fitness function designs to further reduce computational cost. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Estimation Technique of Cs Retention Fraction in Irradiated Nuclear Fuel with Intact/Damaged Form

Even in the case of severe accident such as in TMI-2 or Chernobyl-4 or Fukushima Dai-ichi nuclear power station, waste partitioning in the decommission process or special nuclear material (SNM) accountancy in molten core material is important. Though Cs-137 concentration information is used for burnup estimation in conventional spent nuclear fuels, it could not be applied for damaged nuclear fuel of severe accidents because of both a large release fraction of Cs and significant change of its shape and form by core melting. The ratio of {sup 134}Cs/{sup 137}Cs has often been used as an index of the burnup in the cleanup of TMI-2 and the Chernobyl Drum Assay System in Chernobyl-4 and other cleanup activities, assuming that the Cs isotope ratio inside the fuel matrix would be the same as in the fuel debris, whereas its retention fraction would vary. Cs retention fraction information would be very important for the reliability of the burnup estimation as well as for the severe accident safety analysis. In the present research, a principle of new technique to estimate the Cs retention fraction is proposed and examined by sensitivity analysis of the FP inventory in conventional BWR irradiated fuel numerically, and validated by comparing the past experimental data based on gamma spectrometry. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗