Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Isotope Program”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Charge breeding results and future prospects with electron cyclotron resonance ion source and electron beam ion source (invited)

The Californium Rare Ion Breeder Upgrade (CARIBU) of the Argonne National Laboratory ATLAS facility will provide low-energy and reaccelerated neutron-rich radioactive beams for the nuclear physics program. A 70 mCi {sup 252}Cf source produces fission fragments which are thermalized and collected by a helium gas catcher into a low-energy particle beam with a charge of 1+ or 2+. An electron cyclotron resonance (ECR) ion source functions as a charge breeder in order to raise the ion charge sufficiently for acceleration in the ATLAS linac. The final CARIBU configuration will utilize a 1 Ci {sup 252}Cf source to produce radioactive beams with intensities up to 10{sup 6} ions/s for use in the ATLAS facility. The ECR charge breeder has been tested with stable beam injection and has achieved charge breeding efficiencies of 3.6% for {sup 23}Na{sup 8+}, 15.6% for {sup 84}Kr{sup 17+}, and 13.7% for {sup 85}Rb{sup 19+} with typical breeding times of 10 ms/charge state. For the first radioactive beams, a charge breeding efficiency of 11.7% has been achieved for {sup 143}Cs{sup 27+} and 14.7% for {sup 143}Ba{sup 27+}. The project has been commissioned with a radioactive beam of {sup 143}Ba{sup 27+} accelerated to 6.1 MeV/u. In order to take advantage of its lower residual contamination, an EBIS charge breeder will replace the ECR charge breeder in the next two years. The advantages and disadvantages of the two techniques are compared taking into account the requirements of the next generation radioactive beam facilities.

43 PARTICLE ACCELERATORS↗
DOE OSTI2013

Parallel computation of multigroup reactivity coefficient using iterative method

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

LANL Activities: FY12-Q2 National Risk Assessment Partnership (NRAP): Industrial Carbon Capture Program (Quarterly Report)

This report summarizes progress of LANL activities related to the tasks performed under the LANL FWP FE102-002-FY10, National Risk Assessment Partnership (NRAP): Industrial Carbon Capture Program. This FWP is funded through the American Recovery and Reinvestment Act (ARRA). Overall, the NRAP activities are focused on understanding and evaluating risks associated with large-scale injection and long-term storage of CO 2 in deep geological formations. One of the primary risks during large-scale injection is due to changes in geomechanical stresses to the storage reservoir, to the caprock/seals and to the wellbores. These changes may have the potential to cause CO 2 and brine leakage and geochemical impacts to the groundwater systems. While the importance of these stresses is well recognized, there have been relatively few quantitative studies (laboratory, field or theoretical) of geomechanical processes in sequestration systems. In addition, there are no integrated studies that allow evaluation of risks to groundwater quality in the context of CO 2 injection-induced stresses. The work performed under this project is focused on better understanding these effects. LANL approach will develop laboratory and computational tools to understand the impact of CO 2 -induced mechanical stress by creating a geomechanical test bed using inputs from laboratory experiments, field data, and conceptual approaches. The Geomechanical Test Bed will be used for conducting sensitivity and scenario analyses of the impacts of CO 2 injection. The specific types of questions will relate to fault stimulation and fracture inducing stress on caprock, changes in wellbore leakage due to evolution of stress in the reservoir and caprock, and the potential for induced seismicity. In addition, the Geomechanical Test Bed will be used to investigate the coupling of stress-induced leakage pathways with impacts on groundwater quality. LANL activities are performed under two tasks: (1) develop laboratory and computational tools to understand CO 2 -induced mechanical impacts and (2) use natural analog sites to determine potential groundwater impacts. We are using the Springerville-St. John Dome as a field site for collecting field data on CO 2 migration through faults and groundwater impacts as well as developing and validating computational models. During the FY12 second quarter we have been working with New England Research Company to construct a tri-axial core-holder. We have built fluid control system for the coreflood system that can be ported to perform in-situ imaging of core. We have performed numerical simulations for groundwater impacts of CO 2 and brine leakage using the reservoir model for Springerville-St John's Dome site. We have analyzed groundwater samples collected from Springerville site for major ion chemistry and isotopic composition. We are currently analyzing subsurface core and chip samples acquired for mineralogical composition.

25 ENERGY STORAGE↗
DOE OSTI2012

SCWR Once-Through Calculations for Transmutation and Cross Sections

It is the purpose of this report to document the calculation of (1) the isotopic evolution and of (2) the 1-group cross sections as a function of burnup of the reference Super Critical Water Reactor (SCWR), in a format suitable for the Fuel Cycle Option Campaign Transmutation Data Library. The reference SCWR design was chosen to be that described in [McDonald, 2005]. Super Critical Water Reactors (SCWR) are intended to operate with super-critical water (i.e. H2O at a pressure above 22 MPa and a temperature above 373oC) as a cooling – and possibly also moderating – fluid. The main mission of the SCWR is to generate lower cost electricity, as compared to current standard Light Water Reactors (LWR). Because of the high operating pressure and temperature, SCWR feature a substantially higher thermal conversion efficiency than standard LWR – i.e. about 45% versus 33%, mostly due to an increase in the exit water temperature from ~300oC to ~500oC – potentially resulting in a lower cost of generated electricity. The coolant remains single phase throughout the reactor and the energy conversion system, thus eliminating the need for pressurizers, steam generators, steam separators and dryers, further potentially reducing the reactor construction capital cost. The SCWR concept presented here is based on existing LWR technology and on a large number of existing fossil-fired supercritical boilers. However, it was concluded in [McDonald, 2005], that: “Based on the results of this study, it appears that the reference SCWR design is not feasible.” This conclusion appears based on the strong sensitivity of the design to small deviations in nominal conditions leading to small effects having a potentially large impact on the peak cladding temperature of some fuel rods. “This was considered a major feasibility issue for the SCWR” [McDonald, 2005]. After a description of the reference SCWR design, the Keno V 3-D single assembly model used for this analysis, as well as the calculated results, are presented. Additionally, the follwing information, presented in the appendixes, is intended to provide enough guidance that a researcher repeating the same task in the future should be able to obtain a vector of nuclei and cross sections ready for insertion into the transmutation library without any need for further instructions: (1) Complete TRITON/KENO-V input used for the analysis; (2) Inputs and detailed description of the usage of the OPUS utility, used to postproces and to extract the nuclei concentrations for the transmutation library; (3) Inputs and detailed description of the usage of the XSECLIST utility, used to postproces and to extract the 1-group cross sections for the transmutation library; (4) Details of an ad-hoc utility program developed to sort the nuclei and cross sections for the transmutation library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Which nuclear data can be validated with LLNL pulsed-sphere experiments?

It is shown here that 14-MeV D+T LLNL pulsed-sphere experiments bring complementary information into the process of validating nuclear data compared to experiments that are traditionally used for this purpose—such as critical assemblies. To be more specific, the 14-MeV D+T LLNL pulsed-sphere neutron-leakage spectra enable to validate scattering and fission nuclear data up to 15 MeV (compared to approximately up to 5 MeV when using criticality experiments) and employ to this end simple compound targets containing only few isotopes. In this work, sensitivity profiles of the spectra to nuclear data are calculated in order to understand in detail which isotopes, observables, and energy ranges of nuclear data contribute significantly to their simulation. These profiles are presented for a few selected spheres containing 16 O, 12 C, 56 Fe, and 239 Pu. It is shown that the neutron-leakage spectra of spheres containing light elements are mostly sensitive to elastic- and inelastic-scattering cross sections on discrete levels and corresponding angular distributions. Spheres of structural materials are sensitive to elastic- and inelastic-scattering cross sections, including scattering on discrete levels and the continuum, and double-differential cross sections. Actinide spheres are also strongly sensitive to the fission observables, in particular to the total-fission neutron spectrum. Thin spheres (in which neutrons experience on average less than one scatter) are mostly sensitive to data near the elastic peak, in the energy range from 12–15 MeV, while thicker ones can be sensitive to data at lower incident-neutron energies due to multiple-scattering effects. This information is brought together with simulations of 71 pulsed-sphere neutron-leakage spectra using the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear-data libraries. This analysis highlights ENDF/B-VIII.0 data that could be further investigated for potential shortcomings ( 6 Li, 12 C, 16 O, 24-26 Mg, 27 Al, 48 Ti, 56 Fe, and 208 Pb) or are likely reliable ( 1,2 H, 7 Li, 9 Be, 14 N, 235,238 U, and 239 Pu) as indicated by validating with LLNL pulsed-sphere experiments.

14-MeV D+T LLNL pulsed-sphere neutron-leakage spec↗
DOE OSTI2025

Machine Learning to Select Experiments Driven by Fundamental Science and Applications for Targeted Nuclear Data Improvement

This work describes a blueprint for a process that accelerates progress in science by quantitatively answering the following question: What is the optimal combination of fundamental-science and application-driven experiments to maximally reduce pertinent data uncertainties? Answering this question entails solving a high-dimensional and complex optimization problem that is best solved with advanced statistic techniques often classified as machine learning. We apply this process within the framework of nuclear data with the aim to select an experiment combination that will reduce uncertainties in 239 Pu nuclear data for neutron energies between 1 and 600 keV. In this field, fundamental-physics driven data, called differential, look at one nuclear physics observable at a time. They are contrasted to application-driven, integral, data where one or few resulting values inform a broad set of nuclear data across several nuclides and energies. The candidates for integral experiments are criticality measurements that were refined by a genetic algorithm to be maximally sensitive to 239 Pu fission cross sections in the desired energy range. Twenty-three candidate differential experiments were investigated and span multiple nuclear physics observables (e.g., total, capture cross sections) for isotopes appearing in the integral experiments. The optimal combination among these candidate experiments was investigated via generalized least squares fitting, augmented with Gaussian processes to ameliorate statistical irregularities in data, and the D-optimality criterion. The latter evaluates for each pair of candidates the joint reduction in uncertainties of all 12200 nuclear data appearing in the integral experiments compared to the knowledge we have from 168 past experiments, theory, and nuclear data. We chose as differential measurements those that investigate 63 Cu and 239 Pu total cross sections, based on D-optimality rank and feasibility constraints. Two integral (criticality) experiments were selected: An experiment with Al 2 ⁢O 3 and graphite interleaved with Pu and a thick Cu reflector explores 1–30 keV, while we target the 30–600 keV range with an experiment that swaps boron in place of graphite with a different geometry.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

SU-E-T-136: Assessment of Seasonal Linear Accelerator Output Variations and Associated Impacts

Purpose: Application of process control theory in quality assurance programs promises to allow earlier identification of problems and potentially better quality in delivery than traditional paradigms based primarily on tolerances and action levels. The purpose of this project was to characterize underlying seasonal variations in linear accelerator output that can be used to improve performance or trigger preemptive maintenance. Methods: Review of runtime plots of daily (6 MV) output data acquired using in house ion chamber based devices over three years and for fifteen linear accelerators of varying make and model were evaluated. Shifts in output due to known interventions with the machines were subtracted from the data to model an uncorrected scenario for each linear accelerator. Observable linear trends were also removed from the data prior to evaluation of periodic variations. Results: Runtime plots of output revealed sinusoidal, seasonal variations that were consistent across all units, irrespective of manufacturer, model or age of machine. The average amplitude of the variation was on the order of 1%. Peak and minimum variations were found to correspond to early April and September, respectively. Approximately 48% of output adjustments made over the period examined were potentially avoidable if baseline levels had corresponded to the mean output, rather than to points near a peak or valley. Linear trends were observed for three of the fifteen units, with annual increases in output ranging from 2–3%. Conclusion: Characterization of cyclical seasonal trends allows for better separation of potentially innate accelerator behaviour from other behaviours (e.g. linear trends) that may be better described as true out of control states (i.e. non-stochastic deviations from otherwise expected behavior) and could indicate service requirements. Results also pointed to an optimal setpoint for accelerators such that output of machines is maintained within set tolerances and interventions are required less frequently.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

2015 Advanced Site Investigation and Monitoring Report Riverton, Wyoming, Processing Site September 2016

The U.S. Department of Energy conducted initial groundwater characterization of the Riverton, Wyoming, Processing Site in the 1990s. The characterization culminated in a Site Observational Work Plan in 1998 that recommended a natural flushing compliance strategy. Results of verification monitoring indicated that natural flushing was generally progressing as expected until June 2010, when significant increases in contaminant concentrations were measured in several monitoring wells downgradient of the site after the area flooded. In response to the unexpected results following the flood, an enhanced characterization of the surficial aquifer was conducted in 2012, which included installation of 103 boreholes along nine transects with a Geoprobe, collection of 103 water samples and 65 soil samples, laboratory tests on the soil samples, and additional groundwater modeling. This advanced site investigation report summarizes additional investigation in 2015 through the use of backhoe trenching, sonic drilling, multilevel monitoring wells, direct-push drilling, and temporary well points to collect soil and groundwater samples. Additional surface water measurements were made included the installation of a stilling well and the measurement of stream elevation along the Wind River to approximate upgradient groundwater heads. Groundwater sampling included the addition of geochemical constituents and isotopes that have not been sampled in the past to better understand post-flood conditions and the possibility of additional or ongoing contaminant sources. This sampling was performed to (1) better define the contaminant plumes, (2) verify the occurrence of persistent secondary contaminant sources, (3) better understand the reason for the contaminant spikes after a 2010 flood, and (4) assess contaminant plume stagnation near the Little Wind River. This report provides data analyses and interpretations for the 2015 site investigation that addresses these issues and provides recommendations for future efforts. Observations from trenches and sonic drilling indicate the general lithology of the shallow, unconsolidated sediments consists of a silt zone at the surface that ranges from 2.5 to 4.8 feet below ground surface, underlain by sand and gravel, underlain by the top of the weathered bedrock (Wind River Formation). Soil data from trenches and sonic drilling indicate (1) elevated concentrations of several constituents in the silt zone, likely due to the formation of evaporites, (2) uranium is the only measured element that appears to be concentrated in the silt over the groundwater contaminant plume, (3) in the former tailings impoundment area, there may be a thin unsaturated zone with elevated uranium in the native material just below the fill, (4) in the former tailings impoundment area, slightly higher uranium concentrations occur in the underlying saturated sand and gravel, and (5) several bedrock samples have a unique geochemical signature, generally related to a higher silt content. Assessment of groundwater flow included measuring river elevations along the Wind River and installing the temporary well points adjacent to the Little Wind River that provided additional data points to refine contours for water table elevations. These data confirm past interpretations of groundwater flow to the southeast across the site toward the Little Wind River. Hydraulic head elevations between paired surficial and semiconfined aquifer wells indicate variable vertical gradients across the site with the potential for upward and downward flow. Additional direct-push drilling and groundwater sampling confirmed the contaminant plume configuration, but it also revealed a low-sulfate-concentration zone at the edge of the former tailings impoundment. Temporary well points provided better definition of plume concentrations at the bank of the Little Wind River, and data from these wells indicate plume discharge to the river. Additional sampling in an area southwest of the plume that had elevated uranium groundwater concentrations in the past did not have any uranium concentrations above the U.S. Environmental Protection Agency maximum concentration limit for uranium. Results from multilevel monitoring wells indicate some geochemical differences with depth, but overall concentrations are similar to those in nearby conventional monitoring wells in the long-term monitoring program. Geochemistry data from these multilevel monitoring wells confirm the general increase in contaminant concentrations toward the river and toward the plume centerline for chloride, sulfate, and uranium but highlight geochemical controls on calcium. Iron data indicate slightly reducing conditions, especially near water table and bedrock surfaces, with more oxidizing conditions in the middle of the sand and gravel. Uranium activity ratios (234U/238U) confirm the uranium plume in the surficial aquifer as being mill related, and the area to the southwest outside the plume as natural, non-mill related. In the semiconfined aquifer, evidence of aquifer connection and impacts from the mill is inconclusive. Values of δD and δ 18 O suggest water is derived from different sources and tritium data confirm that the semiconfined aquifer water is generally older than the surficial aquifer water. However, these data do indicate some groundwater communication from the surficial aquifer into the semiconfined aquifer, which resulted in δD and δ 18 O and tritium values in the semiconfined aquifer that are more similar to those in the surficial aquifer. Values of δ 34 S sulfate in the semiconfined aquifer combined with sulfate concentrations indicate the potential for some mill- related sulfate in the semiconfined aquifer, albeit limited to an area near and beneath the former tailings impoundment. Uranium and molybdenum concentrations in the semiconfined aquifer are below groundwater standards in all wells. However, the elevated molybdenum concentrations at one semiconfined aquifer well underneath the former tailings impoundment suggests a mill- derived source for the uranium and molybdenum in that well. It is possible that aquifer cross- communication occurred when the tailings impoundment was active and created a higher head. Current cross-communication appears unlikely given the large differences in tritium values and an upward hydraulic head at this location. Nine domestic wells are located within the institutional control boundary (eight in the confined aquifer and one in the semiconfined aquifer). Uranium and molybdenum concentrations in all samples collected from these wells were one or two orders of magnitude below the groundwater standards. Surface water flow in the Little Wind River in September 2015 was low compared to historical averages for that time of year. As a result, the uranium concentration measured in the Little Wind River was at a historical maximum at the sampling location upstream of the site. However, the impact of uranium discharge from the groundwater plume into the Little Wind River was not measureable. Elevated sulfate concentrations were observed in an outfall ditch related to an active sulfuric acid plant. Uranium concentrations in the oxbow lake remain at concentrations above the groundwater standard. Plume contaminant concentrations had returned to levels found prior to the 2010 flood by the end of 2015. However, these concentrations still exceed model predictions for natural flushing, and the current data indicate that natural flushing to achieve remediation goals within the 100-year time period is not likely, especially with the high potential for additional floods in the update to the conceptual site model (CSM), soil data indicate additional contaminant sources, specifically uranium, in evaporites within the silt layer over the uranium plume and in naturally reduced zones (NRZs). Additional zones of slightly elevated uranium concentration are in the native sediments just above the water table but below the fill layer in the former tailings impoundment area. This area also has slightly elevated uranium in the sand and gravel below the water table. Mass balance calculations indicate that small amounts of dissolution in any of these zones with increased uranium in the solid phase can produce groundwater uranium concentrations above the groundwater standard and could account for the post-flood uranium spike. The additional uranium near the former tailings impoundment provides a mechanism for a continuing source for the uranium plume that was not considered in earlier natural flushing models. In addition, uranium released from the silt layer or the NRZs seasonally and during flooding may add uranium to the groundwater plume near the Little Wind River. These mechanisms provide a possible explanation for plume persistence, along with spikes in concentrations during floods, that creates the current plume configuration. Additional updates to the CSM include (1) chloride flushes more rapidly than uranium beneath the former mill site, (2) chloride in the silt layer provides a scenario in which chloride cannot be used as a conservative tracer (especially in areas prone to flooding), (3) uranium concentrations with depth can be variable (especially below NRZs), and (4) calcite and gypsum solubility limits appear to provide important geochemical controls on groundwater geochemistry. The conclusion of this study provides several recommendations for additional work to refine the CSM and continue assessment of the natural flushing compliance strategy. Recommendations for additional work include targeted soil and groundwater sampling to assess geochemical conditions, distribution of contaminants, and groundwater/surface water interaction; additional column tests to provide data for geochemical modeling; and development of an updated groundwater flow model, which will be used in conjunction with a geochemical model to assess the viability of the natural flushing compliance strategy.

2015 Monitoring Riverton, WY↗
DOE OSTI2015

Evaluation on the Effect of Composition on Radiation Hardening and Embrittlement in Model FeCrAl Alloys

This report details the findings of post-radiation mechanical testing and microstructural characterization performed on a series of model and commercial FeCrAl alloys to assist with the development of a cladding technology with enhanced accident tolerance. The samples investigated include model alloys with simple ferritic grain structure and two commercial alloys with minor solute additions. These samples were irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to nominal doses of 7.0 dpa near or at Light Water Reactor (LWR) relevant temperatures (300-400 C). Characterization included a suite of techniques including small angle neutron scattering (SANS), atom probe tomography (APT), and transmission based electron microscopy techniques. Mechanical testing included tensile tests at room temperature on sub-sized tensile specimens. The goal of this work was to conduct detailed characterization and mechanical testing to begin establishing empirical and/or theoretical structure-property relationships for radiation-induced hardening and embrittlement in the FeCrAl alloy class. Development of such relationships will provide insight on the performance of FeCrAl alloys in an irradiation environment and will enable further development of the alloy class for applications within a LWR environment. A particular focus was made on establishing trends, including composition and radiation dose. The report highlights in detail the pertinent findings based on this work. This report shows that radiation hardening in the alloys is primarily composition dependent due to the phase separation in the high-Cr FeCrAl alloys. Other radiation induced/enhanced microstructural features were less dependent on composition and when observed at low number densities, were not a significant contributor to the observed mechanical responses. Pre-existing microstructure in the alloys was found to be important, with grain boundaries and pre-existing dislocation networks acting as defect sinks, resulting in variations in the observed microstructures after irradiation. Dose trends were also observed, with increasing radiation dose promoting changes in the size and number density of the Cr-rich α' precipitates. Based on the microstructural analysis, performed tensile testing, and prior knowledge from FeCr literature it was hypothesized that the formation of the Cr-rich α' precipitates could lead to significant radiation-induced embrittlement in the alloys, and this could be composition dependent, a result which would mirror the trends observed for radiation-induced hardening. Due to the limited database on embrittlement in the FeCrAl alloy class after irradiation, a series of radiation experiments have been implemented. The overarching point of view within this report is the radiation tolerance of FeCrAl is complex, with many mechanisms and factors to be considered at once. Further development of the FeCrAl alloy class for enhanced accident tolerant applications requires detailed, single (or at least limited) variable experiments to fully comprehend and predict the performance of this alloy in LWRs. This report has been submitted as fulfillment of milestone M2FT-15OR0202321 titled, Summary report on the effect of composition on the irradiation embrittlement of Gen 1 ATF FeCrAl for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

36 MATERIALS SCIENCE↗
DOE OSTI2016

SU-F-T-574: MLC Based SRS Beam Commissioning - Minimum Target Size Investigation

Purpose: To implement a MLC accelerator based SRS program using small fields down to 1 cm × 1 cm and to determine the smallest target size safe for clinical treatment. Methods: Computerized beam scanning was performed in water using a diode detector and a linac-head attached transmission ion chamber to characterize the small field dosimetric aspects of a 6 MV photon beam (Trilogy-Varian Medical Systems, Inc.). The output factors, PDD and profiles of field sizes 1, 2, 3, 4, and 10 cm{sup 2} were measured and utilized to create a new treatment planning system (TPS) model (AAA ver 11021). Static MLC SRS treatment plans were created and delivered to a homogeneous phantom (Cube 20, CIRS, Inc.) for a 1.0 cm and 1.5 cm “PTV” target. A 12 field DMLC plan was created for a 2.1 cm target. Radiochromic film (EBT3, Ashland Inc.) was used to measure the planar dose in the axial, coronal and sagittal planes. A micro ion chamber (0.007 cc) was used to measure the dose at isocenter for each treatment delivery. Results: The new TPS model was validated by using a tolerance criteria of 2% dose and 2 mm distance to agreement. For fields ≤ 3 cm{sup 2}, the max PDD, Profile and OF difference was 0.9%, 2%/2mm and 1.4% respectively. The measured radiochromic film planar dose distributions had gamma scores of 95.3% or higher using a 3%/2mm criteria. Ion chamber measurements for all 3 test plans effectively met our goal of delivering the dose accurately to within 5% when compared to the expected dose reported by the TPS (1 cm plan Δ= −5.2%, 1.5 cm plan Δ= −2.0%, 2 cm plan Δ= 1.5%). Conclusion: End to end testing confirmed that MLC defined SRS for target sizes ≥ 1.0 cm can be safely planned and delivered.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Novel principal component analysis tool based on python for analysis of complex spectra of time-of-flight secondary ion mass spectrometry

Time-of-flight secondary ion mass spectrometry (ToF-SIMS) is a powerful surface analysis tool, which can simultaneously provide elemental, isotopic, and molecular information with part per million (ppm) sensitivity. However, each spectrum may be composed of hundreds of ion signals, which makes the spectra data complex. Principal component analysis (PCA) is a multivariate analysis technique that has been widely used to figure out the variances among samples in ToF-SIMS spectra data analysis and is showing great success in the explanation of complex ToF-SIMS spectra. So far, several software tools have been developed for PCA of ToF-SIMS spectra; however, none of them are freely available. Such a situation leads to some difficulties in extending applications of PCA to various research fields. More importantly, it has long been challenging for common researchers to understand PCA plots and extract chemical differences among samples. In this work, we developed a new and flexible software tool (named “advanced spectra pca toolbox”) based on python for PCA of complex ToF-SIMS spectra along with an easy-to-read manual. It can generate data analysis reports automatically to explain chemical differences among samples, allowing less experienced researchers to easily understand tricky PCA results. Moreover, it is expandable and compatible with artificial intelligence/machine learning functions. Pure goethite and different lignin adsorbed goethite samples were used as a model system to demonstrate our new software tool, proving that our software tool can be readily used in complex spectra data processing. Our new software tool is open-source, convenient, flexible, and expandable. We expect this open-source tool will benefit the ToF-SIMS community.

47 OTHER INSTRUMENTATION↗
DOE OSTI2021

Nuclear Data and Cross Section Testing Using ENDF/B-VIII.0

With the release of the Evaluated Nuclear Data File (ENDF)/B-VIII.0 library, nuclear criticality safety practitioners and engineers have access to the latest cross section sets available for their analyses. However, these cross sections must be rigorously tested and validated to ensure that the nuclear data are responsive to the needs of the individuals responsible for developing, implementing, and maintaining computational tools for criticality safety applications. Thus, the ENDF/B-VIII.0 library is tested and validated with a large collection of experiments that were vetted by the International Criticality Safety Benchmark Evaluation Project and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. A selection of benchmark experiments for use within the criticality safety community were prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID), which is maintained by the Nuclear Energy and Fuel Cycle Division at Oak Ridge National Laboratory. The performance of the ENDF/B-VIII.0 library is assessed by using VALID models of benchmark experiments with the beta 12 version of SCALE 6.3 KENO V.a and KENO-VI Monte Carlo codes. The performance is compared with the results obtained from with the ENDF/B-VII.1 library. This report considers multigroup (MG) and continuous energy (CE) formats of the ENDF/B-VIII.0 and -VII.1 libraries. The benchmark experiments within VALID that validate the ENDF/B-VIII.0 library cover 15 broad system categories by using a range of fissile materials, uranium enrichments, plutonium isotopic vectors, and mixed uranium/plutonium systems. These forms are represented as metals, solutions, or various arrays of rods or plates that cover a variety of neutron energy spectra: thermal, fast, mixed, and intermediate. Over 600 cases were considered for use with the KENO V.a and KENO-VI codes with the ENDF/B-VIII.0 library. The results of the Monte Carlo comparison of ENDF/B-VIII.0 to ENDF/B-VII.1 with both KENO V.a and KENO-VI indicate that there is a less than 0.53% Δk difference between the bias of calculated k eff from the expected values. The CE ENDF/B-VIII.0 library results in smaller magnitude biases than the ENDF/B-VII.1 data for HEU-MET-FAST, HEU-SOL-THERM, IEU-MET-FAST, LEU-SOL-THERM, PU-SOL-THERM, and U233-MET-FAST systems, while the MG results yielded smaller magnitude biases for HEU-MET-FAST, HEU-SOL-THERM, IEU-MET-FAST, LEU-COMP-THERM, LEU-SOL THERM, MIX-COMP-FAST, and U233-MET-FAST systems. Most notable are the adjustments to the plutonium and 233 U cross section data, which has resulted in noticeably lower biases in the ENDF/B VIII.0 results for the mixed, plutonium, and 233 U systems. Results of the sensitivity data file comparison generated from TSUNAMI-3D for selected VALID cases for the ENDF/B-VIII.0 library indicate a very high level of agreement with correlation coefficients of the effect of nuclear data uncertainty on k eff (the c k integral parameter) all above 0.99. This indicates that cases with the ENDF/B-VIII.0 library would see very similar responses to any nuclear data errors or change as those with the ENDF/B-VII.1 library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Materials Irradiation Facility (MIF) at the High Flux Isotope Reactor (HFIR)

The Material Irradiation Facility (MIF) at the Oak Ridge National Laboratory (ORNIL) High Flux Isotope Reactor (HIFR) controls and monitors instrumented irradiation experiments for materials and fuel research. In 2013 ORNIL founded a Laboratory Directed Research and Development project to reinvigorate the use of instrumented experiments by redesigning MIF to enhance its capabilities. MIF improvements include an updated and more sophisticated experiment monitoring system, a modular design allowing easier addition of simultaneous experiments, and features that allow rapid incorporation of additional instrumentation Interest in materials and fuel irradiation has increased significantly over the last decade. Much of this interest is related to investigating the performance of advanced materials for fusion environments, designing and developing improved fuel and cladding materials, and examining material characteristics and behavior under irradiation conditions. At ORNL, the irradiation engineering team has more than 60 years of experience designing and fabricating various types of irradiation experiments. Rabbit capsule experiments are small scale and relatively inexpensive. These capsules are typically designed to be irradiated in the flux-trap region of the core. Full-length target capsules are larger scale and could be either instrumented or non-instrumented. These more complicated capsules can be irradiated in any of the experiment facilities at HFIR. Instrumented experiment irradiation provides significant advantages for researchers. The unique advantage is that the researcher can monitor and acquire in situ data and change capsule operating conditions in real time during the irradiation. This capability enables tight control of desired experimental conditions so that they match the design specifications throughout the irradiation. Also, the experiment conditions (e.g., temperature and pressure) can be altered during the test, enabling significant flexibility and the ability to collect considerably more data from a single experimental capsule. This capability allows measurement of a variety of material properties (conductivity, fission product composition, etc.) that otherwise would only be available upon post-irradiation examination, where non-equilibrium conditions brought on by irradiation do not exist. The data generated by these experiments contribute significantly to the understanding of material behavior under irradiation. MIF is an experimental platform that can host fully instrumented and monitored experiments in a high neutron flux environment. Its primary functions are to provide clean inert gas (helium, neon, argon) to experiments at controlled pressures and flows, monitor irradiation conditions (typically temperatures and pressures) during irradiation, and monitor effluent gases from experiments. This facility allows monitoring of chemical processes during a fueled irradiation, where this becomes important for detecting fuel failures. This paper reports the efforts towards reestablishing the value of HFIR for instrumented irradiation experiments and the improvements to MIF that ultimately reduce the costs of instrumented irradiation. This project advances ORNL materials capabilities to new levels by providing a state-of-the art, robust and flexible facility that can be expanded as customer requirements change, while supporting the goals of the fusion, advanced fuels, and structural materials programs. (authors)

42 ENGINEERING↗
DOE OSTI2016

Viability Study for an Unattended UF 6 Cylinder Verification Station: Phase I Final Report

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS) that could provide automated, independent verification of the declared relative enrichment, 235 U mass, total uranium mass and identification for all declared UF 6 cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs to the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The US Support Program team consisted of Pacific Northwest National Laboratory (PNNL, lead), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL) and Savanah River National Laboratory (SRNL). At the core of the viability study is a long-term field trial of a prototype UCVS system at a Westinghouse fuel fabrication facility. A key outcome of the study is a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This report provides context for the UCVS concept and the field trial: potential UCVS implementation concepts at an enrichment facility; an overview of UCVS prototype design; field trial objectives and activities. Field trial results and interpretation are presented, with a focus on the performance of PNEM and HEVA for the assay of over 200 “typical” Type 30B cylinders, and the viability of an “NDA Fingerprint” concept as a high-fidelity means to periodically verify that the contents of a given cylinder are consistent with previous scans. A modeling study, combined with field-measured instrument uncertainties, provides an assessment of the partial-defect sensitivity of HEVA and PNEM for both one-time assay and (repeated) NDA Fingerprint verification scenarios. The findings presented in this report represent a significant step forward in the community’s understanding of the strengths and limitations of the PNEM and HEVA NDA methods, and the viability of the UCVS concept in front-end fuel cycle facilities. This experience will inform Phase II of the UCVS viability study, should the IAEA pursue it.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Latent uncertainties of the precalculated track Monte Carlo method

Purpose: While significant progress has been made in speeding up Monte Carlo (MC) dose calculation methods, they remain too time-consuming for the purpose of inverse planning. To achieve clinically usable calculation speeds, a precalculated Monte Carlo (PMC) algorithm for proton and electron transport was developed to run on graphics processing units (GPUs). The algorithm utilizes pregenerated particle track data from conventional MC codes for different materials such as water, bone, and lung to produce dose distributions in voxelized phantoms. While PMC methods have been described in the past, an explicit quantification of the latent uncertainty arising from the limited number of unique tracks in the pregenerated track bank is missing from the paper. With a proper uncertainty analysis, an optimal number of tracks in the pregenerated track bank can be selected for a desired dose calculation uncertainty. Methods: Particle tracks were pregenerated for electrons and protons using EGSnrc and GEANT4 and saved in a database. The PMC algorithm for track selection, rotation, and transport was implemented on the Compute Unified Device Architecture (CUDA) 4.0 programming framework. PMC dose distributions were calculated in a variety of media and compared to benchmark dose distributions simulated from the corresponding general-purpose MC codes in the same conditions. A latent uncertainty metric was defined and analysis was performed by varying the pregenerated track bank size and the number of simulated primary particle histories and comparing dose values to a “ground truth” benchmark dose distribution calculated to 0.04% average uncertainty in voxels with dose greater than 20% of D{sub max}. Efficiency metrics were calculated against benchmark MC codes on a single CPU core with no variance reduction. Results: Dose distributions generated using PMC and benchmark MC codes were compared and found to be within 2% of each other in voxels with dose values greater than 20% of the maximum dose. In proton calculations, a small (≤1 mm) distance-to-agreement error was observed at the Bragg peak. Latent uncertainty was characterized for electrons and found to follow a Poisson distribution with the number of unique tracks per energy. A track bank of 12 energies and 60000 unique tracks per pregenerated energy in water had a size of 2.4 GB and achieved a latent uncertainty of approximately 1% at an optimal efficiency gain over DOSXYZnrc. Larger track banks produced a lower latent uncertainty at the cost of increased memory consumption. Using an NVIDIA GTX 590, efficiency analysis showed a 807 × efficiency increase over DOSXYZnrc for 16 MeV electrons in water and 508 × for 16 MeV electrons in bone. Conclusions: The PMC method can calculate dose distributions for electrons and protons to a statistical uncertainty of 1% with a large efficiency gain over conventional MC codes. Before performing clinical dose calculations, models to calculate dose contributions from uncharged particles must be implemented. Following the successful implementation of these models, the PMC method will be evaluated as a candidate for inverse planning of modulated electron radiation therapy and scanned proton beams.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI1988

Nuclear Safety [Vol. 29, No. 1, January-March 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. THE CHERNOBYL ACCIDENT: 1 Core History and Nuclide Inventory of the Chernobyl Core at the Time of Accident by Gerald Kirchner and Cornelius C. Noack; GENERAL SAFETY CONSIDERATIONS: 6 Nonprescriptive Nuclear Safety Regulation: The Example of Loss of Offsite Power by M. W. Golay, V. P. Manno, and C. Vlahoplus, Jr., 20 Nuclear Power Safety Goals in Light of the Chernobyl Accident by C. Whipple and C. Starr; ACCIDENT ANALYSIS: 29 Containment Loads from Severe Accidents—U.S. Program by W. Kerr and M. K. Dey; PLANT SAFETY FEATURES: 36 Safety Characteristics of Modern High-Temperature Reactors: Focus on German Designs by W. Kröger, H. Nickel, and R. Schulten; ENVIRONMENTAL EFFECTS: 49 Radiation Hormesis and Nuclear Safety by C. C. Congdon; OPERATING EXPERIENCES: 58 Evaluation of Nonradiological Water Chemistry at Power Reactors by Harvey Zibulsky, James J. Kottan, Walter J. Pasciak, Mary Ann Castrogivanni, and Sujit Banerjee, 64 Reactor Shutdown Experience Compiled by J. W. Cletcher, 67 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 85 General Administrative Activities Compiled by E. G. Silver, 94 Reports, Standards, and Safety Guides by D. S. Queener, 100 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 103 Proposed Rule Changes as of Sept. 30, 1987; ANNOUNCEMENTS: 5 Northwestern University Short Course on Radiation Safety, 28 International Approach to Nuclear Safety (After Three-Mile Island and Chernobyl), 57 RPI Short Course on Modern Developments in Boiling Heat Transfer and Two-Phase Flow, 107 20th DOE/NRC Nuclear Air Cleaning Conference, 110 Third International Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations, 108 The Authors, 111 Indexes to Nuclear Safety, Volume 28.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Numerical modeling of the SNS H{sup −} ion source

Ion source rf antennas that produce H- ions can fail when plasma heating causes ablation of the insulating coating due to small structural defects such as cracks. Reducing antenna failures that reduce the operating capabilities of the Spallation Neutron Source (SNS) accelerator is one of the top priorities of the SNS H- Source Program at ORNL. Numerical modeling of ion sources can provide techniques for optimizing design in order to reduce antenna failures. There are a number of difficulties in developing accurate models of rf inductive plasmas. First, a large range of spatial and temporal scales must be resolved in order to accurately capture the physics of plasma motion, including the Debye length, rf frequencies on the order of tens of MHz, simulation time scales of many hundreds of rf periods, large device sizes on tens of cm, and ion motions that are thousands of times slower than electrons. This results in large simulation domains with many computational cells for solving plasma and electromagnetic equations, short time steps, and long-duration simulations. In order to reduce the computational requirements, one can develop implicit models for both fields and particle motions (e.g. divergence-preserving ADI methods), various electrostatic models, or magnetohydrodynamic models. We have performed simulations using all three of these methods and have found that fluid models have the greatest potential for giving accurate solutions while still being fast enough to perform long timescale simulations in a reasonable amount of time. We have implemented a number of fluid models with electromagnetics using the simulation tool USim and applied them to modeling the SNS H- ion source. We found that a reduced, single-fluid MHD model with an imposed magnetic field due to the rf antenna current and the confining multi-cusp field generated increased bulk plasma velocities of > 200 m/s in the region of the antenna where ablation is often observed in the SNS source. We report here on comparisons of simulated plasma parameters and code performance using more accurate physical models, such as two-temperature extended MHD models, for both a related benchmark system describing a inductively coupled plasma reactor, and for the SNS ion source. We also present results from scaling studies for mesh generation and solvers in the USim simulation code.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

A Case Study in the Application of TSUNAMI-3D - Part 1, Multigroup

The use of the sensitivity and uncertainty tool suite within SCALE, known as the TSUNAMI suite, has grown following its initial introduction more than a decade ago. Proper use of these tools provides a more complete understanding of system responses to perturbations, especially in the area of nuclear cross section uncertainties. Recent advances within TSUNAMI have added generalized perturbation theory to the k{sub eff} sensitivity capability that was included in TSUNAMI originally. The generation of sensitivity data is encouraged in the review process for new critical experiment requests sponsored by the US Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP) to demonstrate that the proposed experiment will satisfy the identified data need. Many individuals participate in this process, but many analysts have not received proper training in the use of the TSUNAMI tools. The adjoint perturbation theory used to calculate sensitivities is not typically included in nuclear engineering curricula, leaving inexperienced engineers particularly at risk for incorrect use of the tools. The first step in the use of the TSUNAMI tools is the generation of a sensitivity data file (SDF), which contains all of the energy-dependent sensitivity data generated for a model. Most often, the SDF is generated from a three-dimensional (3D) model using either the KENO V.a or KENO-VI code within the TSUNAMI-3D sequence. The sensitivity data generated in TSUNAMI-3D must be accurate for correct quantification of system similarity or any other use of the sensitivity data. The sensitivities calculated by TSUNAMI can be affected by many of the input parameters, cross section processing models, and geometry modeling techniques used within the TSUNAMI-3D input. Fortunately, the total sensitivity of the most important nuclides can be confirmed simply by using direct perturbation (DP) calculations. Confirmation of the total sensitivity of these important nuclides provides confidence that the flux and importance function calculated within the TSUNAMI sequence will yield accurate sensitivities for all nuclides, reactions, and energies of interest. This process is essentially identical for the multigroup (MG) methods and the continuous-energy (CE) methods introduced in SCALE 6.2. This paper discusses the use of DP calculations and provides a case study to confirm that accurate sensitivities are generated by TSUNAMI-3D for a particular benchmark experiment. A companion paper provides a similar case study applying the CE TSUNAMI capability to the same critical experiment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗