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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.

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DOE OSTI2016

MO-DE-BRA-01: Flipped Physics Courses Within a Radiologic Technologist Program: Video Production and Long Term Outcomes

Purpose: To determine if the flipped class format has an effect on examination results for a radiologic technologist (RT) program and discuss benefits from creating video resources. Methods: From 2001–2015, students had taken both a radiological physics and quality control (QC) class as a part of their didactic training. In 2005/2006, the creation of videos of didactic lectures and QC test demonstrations allowed for a flip where content was studied at home while exercises and reviews were done in-class. Final examinations were retrospectively reviewed from this timeframe. 12 multiple choice physics questions (MCP) and 5 short answer QC questions (SAQC) were common to pre and post flip exams. The RT program’s ARRT exam scores were also obtained and compared to national averages. Results: In total, 36 lecture videos and 65 quality control videos were created for the flipped content. Data was ∼2.4GB and distributed to students via USB or CD media. For MCP questions, scores improved by 7.9% with the flipped format and significance (Student’s t-test, p<0.05) was found for 3 of the 12 questions. SAQC questions showed improvement by 14.6% and significance was found for 2 of the 5 questions. Student enrollment increased from ∼14 (2001–2004) to ∼23 students (2005–15). Content was continuously added post-flip due to the efficiency of delivery. The QC class in 2003 covered 45 test setups in-class while 65 were covered with video segments in 2014. Flipped materials are currently being repurposed. In 2015, this video content was restructured into an ARRT exam review guide and in 2016, the content was reorganized for fluoroscopy training for physicians. Conclusion: We believe that flipped classes can improve efficiency of content delivery and improve student performance even with an increase in class size. This format allows for flexibility in learning as well as re-use in multiple applications.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

SU-E-T-103: Development and Implementation of Web Based Quality Control Software

Purpose: Historically many radiation medicine programs have maintained their Quality Control (QC) test results in paper records or Microsoft Excel worksheets. Both these approaches represent significant logistical challenges, and are not predisposed to data review and approval. It has been our group's aim to develop and implement web based software designed not just to record and store QC data in a centralized database, but to provide scheduling and data review tools to help manage a radiation therapy clinics Equipment Quality control program. Methods: The software was written in the Python programming language using the Django web framework. In order to promote collaboration and validation from other centres the code was made open source and is freely available to the public via an online source code repository. The code was written to provide a common user interface for data entry, formalize the review and approval process, and offer automated data trending and process control analysis of test results. Results: As of February 2014, our installation of QAtrack+ has 180 tests defined in its database and has collected ∼22 000 test results, all of which have been reviewed and approved by a physicist via QATrack+'s review tools. These results include records for quality control of Elekta accelerators, CT simulators, our brachytherapy programme, TomoTherapy and Cyberknife units. Currently at least 5 other centres are known to be running QAtrack+ clinically, forming the start of an international user community. Conclusion: QAtrack+ has proven to be an effective tool for collecting radiation therapy QC data, allowing for rapid review and trending of data for a wide variety of treatment units. As free and open source software, all source code, documentation and a bug tracker are available to the public at https://bitbucket.org/tohccmedphys/qatrackplus/.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

SU-E-T-434: Evaluation of the Machine Performance Check (MPC) Program for Truebeam 2.0

Purpose: Machine Performance Check (MPC) is a software application used to verify that the TrueBeam machine is operating within major specifications prior to treatment. Used in combination with a phantom named Isocal, it verifies beam output, beam uniformity, treatment isocenter size, coincidence of treatment and imaging isocenters, positioning accuracy of kV and MV imaging systems, accuracy of collimator and gantry rotation angle, positioning accuracy of jaws and MLC leafs, and couch positioning. The tests can be performed semi-automatically and requires approximately 10 minutes of machine time. It is the purpose of this study to report the performance of this program. Methods: A pre-release version of the MPC tool was installed on a Truebeam linac with 6D couch at our center. Baseline beam output measurements were taken for 5 photon beams (6–15 MV, 6 FFF, 10 FFF) and 5 electron beams (6–20 MeV). Deviations from the baseline output were subsequently recorded for several days and compared against independent measurements from a PTW farmer chamber and our daily QA device (Fluke Biomedical Tracker) as part of an ongoing evaluation. Results: The beam output deviations between the MPC and the PTW chamber measurements were within ±0.7% for photons beams and ±1.0% for electrons beams. This was similar to the tracker performance. There were some isolated incidents where the MPC measurements had unexplained spikes (>3%) that disappeared on a repeat measurement. MPC was also able to detect maximum positioning errors in the jaws (1.12 mm), MLCs (1.14 mm), and couch roll (0.11°). Conclusion: Overall, the ability of the MPC to monitor linac output stability was comparable to that of ionization chamber-based measurements. MPC also provided fast daily mechanical tests not currently available in the clinic. How best to utilize this previously unavailable data is still under investigation.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 1

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore fuel product. The long-term EnCore ® fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 2

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore ® fuel product. The long-term EnCore fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15 N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Analysis of Synergistic Fuel Cycle Options with Thorium and Heavy Water Reactors

Fuel cycle options (FCOs) with thorium-based fuels have characteristics that may have benefits for meeting certain objectives, such as transuranic actinide inventory management and overall natural resource utilization. Neutron irradiation of thorium and uranium-233 results in negligible production of transuranic actinides. This advantage is compromised, to a degree, when thorium fuels are used in conjunction with uranium-238; however, strategic reprocessing and fuel use (e.g., recycle of plutonium and/or minor actinides) can be used to manage this problem. Furthermore, thorium FCOs with reprocessing can extend the availability of natural nuclear resources by breeding new fissile material, recovering and recycling unused material, and achieving high burnups. In instances where natural thorium is more readily available than uranium, FCOs which use thorium may be preferable from sustainability and energy security perspectives. However, many hybrid uranium-thorium FCOs struggle to perform well in this regard, since large amounts of external fissile material (and therefore natural uranium) are required to breed the required uranium-233. DOE's evaluation and screening effort reports that for every metric ton (MT) of U-233 bred in a light water reactor for reuse, about 1000 MT of natural uranium are required to support such breeding, much of which is used inefficiently due to enrichment. The combination of HWRs and thorium-based fuels has the potential to reduce this consumption of natural uranium resources. Because heavy water moderates neutrons without significant absorption, fissile material requirements are lower compared to those of LWRs. This has enabled conventional 'once-through' HWRs to operate with only slightly-enriched or even natural uranium. Also, HWRs can be combined with LWRs in multistage FCOs where used LWR fuel can be processed and the remaining fissile content is sufficient to support use as fuel in HWRs in spite of the presence of neutron poisons (e.g., uranium-236); this concept has been demonstrated at a research scale in Korea. Thus, additional energy is extracted from the same initial natural uranium feed and uranium enrichment effort. Another consideration with using thorium-based fuels in HWRs is the current limited operational experience of the reactors and associated fuel cycle facilities using these fuels. However, work continues internationally towards gaining further experience. For example, at equilibrium the third stage of India's planned Three-Stage Nuclear Power Program uses thorium/uranium-233 or thorium/plutonium fuels with a fissile supply from second-stage fast reactors. Further, China plans to experiment with low-enriched uranium (LEU)-Th CANFLEX fuel bundles in a once-through configuration in its HWRs and a long-term objective is to transition to thorium use in a closed HWR FCO. Vanderbilt University (VU) and Oak Ridge National Laboratory (ORNL) are characterizing multi-stage, thermal-reactor FCOs that incorporate thorium and examining their potential performance. This paper presents insights from the examination of FCOs which consider the synergy of HWRs and thorium. Two multi-stage FCOs which incorporate thorium-based fuels in HWRs have been examined for their potential, one which uses recycled PWR uranium oxide fuel in conjunction with thorium in an HWR, and another that uses recycled HWR fuel in an MSR which also includes thorium. Both of the systems studied offer resource utilization improvements as compared to a once-through LWR fueled with low-enrichment uranium and highlight the potential synergy of thorium and HWRs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A Verification of Flux Sensitivity Estimates Using the MCNP Tally Perturbation Tool

Nuclear data is commonly used in applications such as nuclear nonproliferation, safeguards, and criticality safety. More specifically, nuclear data is used in predictive simulation codes like the Monte-Carlo N-Particle (MCNP ® ) transport code, Serpent, and similar radiation transport codes. The improvement of nuclear data enables more precise and accurate simulations, which result in higher fidelity designs and reduced operational/procedural costs. Therefore, the improvement of nuclear data is of paramount importance across the nuclear community. Nuclear data is improved and validated through integral benchmark experiments. The design of benchmark experiments is an extensive process; therefore, these experiments are often optimized on multiple characteristics, including sensitivity to the nuclear data, during the design process. Sensitivity is a measure of how much a quantity changes due to changes in independent variables such as experimental configuration. An experimental design that has a larger sensitivity to the nuclear data of interest will have a larger impact on the accuracy and precision of the validated data. Past integral benchmark experiments have primarily used the effective multiplication factor ($k_{eff}$) as the predominant measured quantity; however, experiments designed with other quantities in mind would be able to optimize on validating different areas of the nuclear data. A primary goal of the EUCLID project is to design, constrain, and reduce compensating errors in experiments focused on quantities other than $k_{eff}$ to better validate nuclear data across the board. Currently, there is a capability in MCNP to easily calculate the sensitivity of $k_{eff}$ to specific nuclear data of numerous reactions types and isotopes (KSEN card); however, the sensitivity of other quantities must be estimated in more strenuous manners. For example, the perturbation feature (PERT card) of MCNP can be used to estimate first-order sensitivities of some response in fixed source simulations. A recent announcement revealed that the first- and second-order perturbation features in previous releases of MCNP contained a bug. It was identified that particles were being scored into the wrong energy bin. The bug is in the most recent public release (MCNP6.2); however, a patch has been added to the most up to date version (MCNP6.2.2) that has not been released publicly. A direct comparison of the PERT card results for an F4 (neutron flux averaged over a cell) tally before and after the patch are shown in figure 1. All simulations used in the sensitivity estimates in this report were performed with MCNP6.2.2. This work verifies the patched MCNP perturbation tool by comparing first order sensitivities made using the PERT card to estimates made using manual perturbation of the compact ENDF (ACE) files.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Poster — Thur Eve — 28: Enabling trajectory-based radiotherapy on a TrueBeam accelerator with the Eclipse treatment planning system

The TrueBeam linear accelerator platform has a developer's mode which permits the user dynamic control over many of the machine's mechanical and radiation systems. Using this research tool, synchronous couch and gantry motion can be programmed to simulate isocentric treatment with a shortened SAD, with benefits such as smaller projected MLC leaf widths and an increased dose rate. In this work, water tank measurements were used to commission a virtual linear accelerator with an 85 cm SAD in Eclipse, from which several arc-based radiotherapy treatments were generated, including an inverse optimized VMAT delivery. For each plan, the pertinent treatment delivery information was extracted from control points specified in the Eclipse-exported DICOM files using the pydicom package in Python, allowing construction of an XML control file. The dimensions of the jaws and MLC positions, defined for an 85 cm SAD in Eclipse, were scaled for delivery on a conventional SAD linear accelerator, and translational couch motion was added as a function of gantry angle to simulate delivery at 85 cm SAD. Ionization chamber and Gafchromic film measurements were used to compare the radiation delivery to dose calculations in Eclipse. With the exception of the VMAT delivery, ionization chamber measurements agreed within 3.3% of the Eclipse calculations. For the VMAT delivery, the ionization chamber was located in an inhomogeneous region, but gamma evaluation of the Gafchromic film plane resulted in a 94.5% passing rate using criteria of 3 mm/3%. The results indicate that Eclipse calculation infrastructure can be used.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Recent development of the Multi-Grid detector for large area neutron scattering instruments

Most of the Neutron Scattering facilities are committed in a continuous program of modernization of their instruments, requiring large area and high performance thermal neutron detectors. Beside scintillators detectors, {sup 3}He detectors, like linear PSDs (Position Sensitive Detectors) and MWPCs (Multi-Wires Proportional Chambers), are the most current techniques nowadays. Time Of Flight instruments are using {sup 3}He PSDs mounted side by side to cover tens of m{sup 2}. As a result of the so-called '{sup 3}He shortage crisis{sup ,} the volume of 3He which is needed to build one of these instruments is not accessible anymore. The development of alternative techniques requiring no 3He, has been given high priority to secure the future of neutron scattering instrumentation. This is particularly important in the context where the future ESS (European Spallation Source) will start its operation in 2019-2020. Improved scintillators represent one of the alternative techniques. Another one is the Multi-Grid introduced at the ILL in 2009. A Multi-Grid detector is composed of several independent modules of typically 0.8 m x 3 m sensitive area, mounted side by side in air or in a vacuum TOF chamber. One module is composed of segmented boron-lined proportional counters mounted in a gas vessel; the counters, of square section, are assembled with Aluminium grids electrically insulated and stacked together. This design provides two advantages: First, magnetron sputtering techniques can be used to coat B{sub 4}C films on planar substrates, and second, the neutron position along the anode wires can be measured by reading out individually the grid signals with fast shaping amplifiers followed by comparators. Unlike charge division localisation in linear PSDs, the individual readout of the grids allows operating the Multi-Grid at a low amplification gain, hence this detector is tolerant to mechanical defects and its production accessible to laboratories equipped with standard equipment. Prototypes of different configurations and sizes have been developed and tested. A demonstrator, with a sensitive area of 0.8 m x 3 m, has been studied during the CRISP European project; it contains 1024 grids, and a surface of isotopically enriched B{sub 4}C film close to 80 m{sup 2}. Its size represented a challenge in terms of fabrication and mounting of the detection elements. Another challenge was to make the gas chamber mechanically compatible with operation in a vacuum TOF chamber. Optimal working condition of this detector was achieved by flushing Ar-CO{sub 2} at a pressure of 50 mbar, and by applying 400 Volts on the anodes. This unusual gas pressure allows to greatly simplifying the mechanics of the gas vessel in vacuum. The detection efficiency has been measured with high precision for different film thicknesses. 52% has been measured at 2.5 Angstrom, in good agreement with the MC simulation. A high position resolution has been achieved by centre of gravity measurement of the TOT (Time-Over-Threshold) signals between neighbouring grids. These results, as well as other detection parameters, including gamma sensitivity and spatial uniformity, will be presented. (author)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Evaluating Age Effects on HEPA Filter Performance

Nuclear grade high efficiency particulate air (HEPA) filters are routinely employed as the final control element in confinement ventilation systems. The traditional HEPA filter utilizes a pleated filter medium of fibrous glass that has been produced in a manner analogous to manufacturing paper. These filters are the physically weakest element in a containment ventilation system. Containment ventilation systems are designed to condition the air-stream (temperature, relative humidity, moisture, particle size distribution of aerosols, and aerosol concentration) to levels/conditions that fibrous glass HEPA filters can handle. Approximately 6000 nuclear grade HEPA filters are purchased each year for use in the DOE/NNSA complex. Additional filters are purchased for use in commercial nuclear power generation and ventilation systems of other nuclear applications. These filters are designed and tested in accordance with the American Society of Mechanical Engineers (ASME) AG-1 Code Nuclear Air and Gas Treatment. This code establishes design requirements and qualification procedures. Qualification testing of axial flow filters is described in paragraphs 5000 and following of AG-1. The series of qualification tests are intended to describe the operating envelope of HEPA filters and verify the reliability of a filter design. AG-1 qualification also confirms the pedigree of components used in manufacture for that filter design. The collective set of qualification tests establishes the assumed operating envelope for AG-1 filters. The effects of age on HEPA filter performance have been a concern for at least two decades. A combination of physical evaluation of aged filter media and a review of filter failure records resulted in Appendix C of the Nuclear Air Cleaning Handbook (NACH), Determination of HEPA Filter Life. This states that HEPA filters are not to be used beyond 10 years from the date of manufacture. Additionally, filters put into service in locations where they may become wet are to be taken out of service five years from date of manufacture. Filters that have become wet are to be taken out of service immediately. Finally, it should also be pointed out that filters held in Level B storage are subject to the same 10-year service life. Very limited bench-scale testing in the 1990's raised concern that aging HEPA filters do not have the strength to withstand an accident scenario. In May of 1999, the DNFSB released a Technical Report 23 entitled HEPA Filters Used in the Department of Energy's Hazardous Facilities. This report expressed concerns for the potential vulnerability of HEPA filters in vital safety systems. Concerns and uncertainty associated with degradation of HEPA filter performance over time led the DOE sites to limit HEPA filter service life to 10 years from the date of manufacture or five years in cases where filters may become wet. Establishment of a conservative service life needs to be based on data from a structured series of tests comparing the performance envelope of new and aged full-scale filters. A service life that is insufficiently conservative endangers workers and the public. One that is excessively conservative costs can cause hundreds of otherwise unnecessary filter changes. This increases exposure of employees, disrupts facility operations, and increase disposal costs by millions of dollars annually. Conclusive data are needed to resolve uncertainty associated with the damaging effects of aging on durability of HEPA filters. Therefore, this study needs to provide a sufficient body of evidence to allow DOE and site professionals to make prudent decisions. An updated version of ASME AG-1 Section FC included a non-mandatory appendix that would include the DOE guidance for service life of HEPA filters. Balloting of this revision generated significant resistance to inclusion of any kind of service life language citing the lack of supporting experimental data. Efforts to resolve negative votes failed and resistance to acceptance of the language increased. The result has been to not include service language in Section FC and requests to remove it from sections that already include such language. The recent activity of the ASME AG-1 Committee on Nuclear Air and Gas Treatment effectively de-emphasizes following DoE's service life guidelines. The DOE Office of Nuclear Safety under its NSR and D program provided the Mississippi State University Institute for Clean Energy Technology (ICET) finding to initiate a study to determine the effects of aging on AG-1 axial flow filters just prior to actions taken within the AG-1 committee regarding service life. The NSR and D funding facilitated timely development of a national technical working group comprised of members from the regulatory community, DOE complex, and commercial power to resolve uncertainty associated with effects of aging. The Department of Energy Office of Environmental Management (DOE-EM) has been funding the ICET to evaluate the performance envelope for AG-1 HEPA filters. ICET is one of DOE-EM's qualified suppliers. Synergy has been achieved by associating the NSR and D aged filter activities to testing activities evaluating the performance envelopes of separator and separator-less AG-1 HEPA filters funded by DOE-EM. The DOE-EM funded activities establish baseline data for the aged filter study and future EM funding will provide long term support for testing additional aged filters. EM funding of new filters includes analytical evaluation of media tensile strength by thermal analysis to establish baseline values and determine mass loss of moisture resistance and acrylic binders due to use. All of these values can be used for comparison of test results for aged filters. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

On-Site Inspection RadioIsotopic Spectroscopy (Osiris) System Development

We have designed and tested hardware and software for the acquisition and analysis of high-resolution gamma-ray spectra during on-site inspections under the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The On-Site Inspection RadioIsotopic Spectroscopy—Osiris—software filters the spectral data to display only radioisotopic information relevant to CTBT on-site inspections, e.g.,132I. A set of over 100 fission-product spectra was employed for Osiris testing. These spectra were measured, where possible, or generated by modeling. The synthetic test spectral compositions include non-nuclear-explosion scenarios, e.g., a severe nuclear reactor accident, and nuclear-explosion scenarios such as a vented underground nuclear test. Comparing its computer-based analyses to expert visual analyses of the test spectra, Osiris correctly identifies CTBT-relevant fission product isotopes at the 95% level or better.The Osiris gamma-ray spectrometer is a mechanically-cooled, battery-powered ORTEC Transpec-100, chosen to avoid the need for liquid nitrogen during on-site inspections. The spectrometer was used successfully during the recent 2014 CTBT Integrated Field Exercise in Jordan. The spectrometer is controlled and the spectral data analyzed by a Panasonic Toughbook notebook computer. To date, software development has been the main focus of the Osiris project. In FY2016-17, we plan to modify the Osiris hardware, integrate the Osiris software and hardware, and conduct rigorous field tests to ensure that the Osiris system will function correctly during CTBT on-site inspections. The planned development will raise Osiris to technology readiness level TRL-8; transfer the Osiris technology to a commercial manufacturer, and demonstrate Osiris to potential CTBT on-site inspectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

SU-G-TeP2-15: Feasibility Study of Fiber-Optic Cerenkov Radiation Sensors for in Vivo Measurement: Dosimetric Characterization and Clinical Application in Proton Beams

Purpose: To evaluate the possibility of a fiber-optic Cerenkov radiation sensor (FCRS) for in vivo dose verification in proton therapy. Methods: The Cerenkov radiation due to the proton beam was measured using a homemade phantom, consisting of a plastic optical fiber (POF, PGSCD1001-13-E, Toray, Tokyo, Japan) connected to each channel of a multianode photomultiplier tube (MAPMT:H7546, Hamamatsu Photonics, Shizuoka, Japan). Data were acquired using a multi-anode photomultiplier tube with the NI-DAQ system (National Instruments Texas, USA). The real-time monitoring graphic user interface was programmed using Labview. The FCRS was analyzed for its dosimetrics characteristic in proton beam. To determine the accuracy of the FCRS in proton dose measurements, we compared the ionization chamber dose measurements using a water phantom. We investigated the feasibility of the FCRS for the measurement of dose distributions near the superficial region for proton plans with a varying separation between the target volume and the surface of 3 patients using a humanoid phantom. Results: The dose-response has good linearity. Dose-rate and energy dependence were found to be within 1%. Depth-dose distributions in non-modulated proton beams obtained with the FCRS was in good agreement with the depth-dose measurements from the ionization chamber. To evaluate the dosimetric accuracy of the FCRS, the difference of isocenter dose between the delivery dose calculated by the treatment planning system and that measured by the FCRS was within 3%. With in vivo dosimetry using the humanoid phantom, the calculated surface doses overestimated measurements by 4%–8% using FCRS. Conclusion: In previous study, our results indicate that the performance of the array-type FCRS was comparable to that of the currently used a multi-layer ion chamber system. In this study, we also believe that the fiber-optic Cerenkov radiation sensor has considerable potential for use with in vivo patient proton dosimetry.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

MO-FG-CAMPUS-IeP1-05: New Ionization Chamber Dosimetry of Absorbed Dose to Water in Diagnostic KV X-Ray Beams

Purpose: To develop new ionization chamber dosimetry of absorbed dose to water in diagnostic kV x-ray beams, by using a beam quality conversion factor, kQ, for Co-60 to kV x-ray and an ionization conversion factor for a water-substitute plastic phantom. Methods: kQ was calculated for aluminum half value-layers (Al-HVLs) of 1.5 mm to 8 mm which were generated by kV x-ray beams of 50 to 120 kVp. Twenty-two energy spectra for ten effective energies (Eeff) were calculated by a SpecCalc program. Depth doses in water were calculated at 5 × 5 to 30 × 30 cm{sup 2} fields. Output factors were also obtained from the dose ratio for a 10 × 10 cm{sup 2} field. kQ was obtained for a PTW30013 Former ion chamber. In addition, an ionization conversion factor of the PWDT phantom to water was calculated. All calculations were performed with EGSnrc/cavity code and egs-chamber codes. Results: The x-ray beam energies for 1.5 mm to 8 mm Al-HVLs ranged in Eeff of 25.7 to 54.3 keV. kQ for 1.5 mm to 8 mm Al-HVLs were 0.831 to 0.897, at 1 and 2 cm depths for a 10 × 10 cm2 field. Similarly, output factors for 5 × 5 to 30 × 30 cm{sup 2} fields were 0.937 to 1.033 for 25.7 keV and 0.857 to 1.168 for 54.3 keV. The depth dose in a PWDT phantom decreased up to 5% compared to that in water at depth of ten percent of maximum dose for 1.5 mm Al-HVL. The ionization ratios of water/PWDT phantoms for the PTW30013 chamber were 1.012 to 1.007 for 1.5 mm to 8 mm Al-HVLs at 1 cm depth. Conclusion: It became possible to directly measure the absorbed dose to water with the ionization chamber in diagnostic kV x-ray beams, by using kQ and the PWDT phantom.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Modernization efforts for the R -Matrix code SAMMY [Abstract]

The R-Matrix code SAMMY is a widely used nuclear data evaluation code focused on the resolved range, which includes corrections for experimental effects. The code is still mostly written in Fortran 77, and uses a memory management system suitable for the time of its initial writing (1984). A modernization effort is under way to bring the code in-line with modern software development practices. A continuous-integration testing framework was added, automating the large existing set of test cases. It is run on every commit. The memory management was updated to current standard practices suitable for modern software analysis tools. The code can be obtained from https://code.ornl.gov/RNSD/SAMMY. The resonance parameters and covariance information are now stored in C++ objects shared by SAMMY and AMPX, the processing code that generates nuclear data libraries for SCALE. This allows for easier maintenance and access to the resonance parameters inside and outside of SAMMY. This feature is already used by accessing and changing parameters in memory in the Bayesian Monte Carlo Evaluation Framework for Cross Sections Nuclear Data and Integral Benchmark Experiments project, Further plans include the switch to the ENDF reading and writing routines in AMPX, as these routines are more robust, easier to maintain, and support more features. Of note here is support for the new GNDS format. Previously it wasn’t easy to share the full covariance matrix for evaluations containing more than one isotope due to limitations on the ENDF format; this is now supported in GNDS. The data are currently available in a binary SAMMY format and can be exported to GNDS to make them more widely available and sharable. The next step will be to use the same resonance processing code at 0K in AMPX and SAMMY as one of the available Reich-Moore R-Matrix formalism. The first step toward this goal is to isolate the reconstruction into a module that takes resonance parameters as its input and does not depend on SAMMY global parameters. This goal has been achieved and it should now be possible to more easily change the resonance formalism and add enhancements as the Phenomenological R-Matrix parameterization of direct, doorway, and compound nuclear reactions discussed elsewhere on this conference. This concerted modernization and enhancement effort provides multiple advantages to the nuclear data community. It will allow parameter optimization using enhanced formalisms, including experimental effects, that better match complex experimental data. Then those evaluated parameters can immediately be passed off to AMPX to be reconstructed with the exact same cross section model and be put into a data library for subsequent testing using SCALE and the Valid Benchmark suite or other suitable benchmark suites.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Validation of Jezebel Reactivity Coefficients and Sensitivity Analysis

Nuclear data validation is often performed today using criticality measurements. The gold standard for criticality measurements is the International Criticality Safety Benchmark Experiment Project (ICSBEP). The validation specifically focuses on the effective multiplication factor ($k_{eff}$). $K_{eff}$ is a relatively easy parameter to infer and has reduced uncertainty due to being at or above critical. However, while $k_{eff}$ is the most documented parameter and its uncertainties and sensitivities have been evaluated in great detail, it cannot be used as a standalone metric to determine inaccuracies in nuclear data (e.g., cross section data, PFNS, nu), which is based on theory, physics, and differential measurements. The Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) project aims to identify compensating errors in specific isotope nuclear data by optimally designing experiments that are, or are not sensitive to a suite of measurement parameters beyond $k_{eff}$. By identifying parameters that are sensitive to each other, oppositely sensitive, or have substantial magnitude differences in sensitivity, experiments can be designed to constrain questionable nuclear data. One sensitivity that is of particular interest to this project includes the sensitivity of reactivity coefficients. Reactivity coefficients compare reactivity, which is related to $k_{eff}$ at two different states therefore being sensitive to small changes in the system. The most common type of reactivity coefficient measurements is comparison to void for a small sample within the assembly. It is key that the sample sizes are small enough to not affect the flux of the full system. Reactivity coefficients were evaluated for many early experiments to better understand transport corrected cross sections. In fact, ICSBEP includes reactivity coefficient results as “Supplemental Measurements” in appendices for a handful of older benchmarks. One of those benchmarks is Jezebel, the bare Pu critical assembly. This paper compares new simulations of reactivity coefficients for Jezebel, and explores the sensitivity of reactivity coefficients to small changes in nuclear data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

TH-A-19A-04: Latent Uncertainties and Performance of a GPU-Implemented Pre-Calculated Track Monte Carlo Method

Purpose: Assessing the performance and uncertainty of a pre-calculated Monte Carlo (PMC) algorithm for proton and electron transport running on graphics processing units (GPU). While PMC methods have been described in the past, an explicit quantification of the latent uncertainty arising from recycling a limited number of tracks in the pre-generated track bank is missing from the literature. With a proper uncertainty analysis, an optimal pre-generated track bank size can be selected for a desired dose calculation uncertainty. Methods: Particle tracks were pre-generated for electrons and protons using EGSnrc and GEANT4, respectively. The PMC algorithm for track transport was implemented on the CUDA programming framework. GPU-PMC dose distributions were compared to benchmark dose distributions simulated using general-purpose MC codes in the same conditions. A latent uncertainty analysis was performed by comparing GPUPMC dose values to a “ground truth” benchmark while varying the track bank size and primary particle histories. Results: GPU-PMC dose distributions and benchmark doses were within 1% of each other in voxels with dose greater than 50% of Dmax. In proton calculations, a submillimeter distance-to-agreement error was observed at the Bragg Peak. Latent uncertainty followed a Poisson distribution with the number of tracks per energy (TPE) and a track bank of 20,000 TPE produced a latent uncertainty of approximately 1%. Efficiency analysis showed a 937× and 508× gain over a single processor core running DOSXYZnrc for 16 MeV electrons in water and bone, respectively. Conclusion: The GPU-PMC method can calculate dose distributions for electrons and protons to a statistical uncertainty below 1%. The track bank size necessary to achieve an optimal efficiency can be tuned based on the desired uncertainty. Coupled with a model to calculate dose contributions from uncharged particles, GPU-PMC is a candidate for inverse planning of modulated electron radiotherapy and scanned proton beams. This work was supported in part by FRSQ-MSSS (Grant No. 22090), NSERC RG (Grant No. 432290) and CIHR MOP (Grant No. MOP-211360)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Quantified abundance of magnetofossils at the Paleocene–Eocene boundary from synchrotron-based transmission X-ray microscopy

The Paleocene–Eocene boundary (~55.8 million years ago) is marked by an abrupt negative carbon isotope excursion (CIE) that coincides with an oxygen isotope decrease interpreted as the Paleocene–Eocene thermal maximum. Biogenic magnetite (Fe 3 O 4 ) in the form of giant (micron-sized) spearhead-like and spindle-like magnetofossils, as well as nano-sized magnetotactic bacteria magnetosome chains, have been reported in clay-rich sediments in the New Jersey Atlantic Coastal Plain and were thought to account for the distinctive single-domain magnetic properties of these sediments. Uncalibrated strong field magnet extraction techniques have been typically used to provide material for scanning and transmission electron microscopic imaging of these magnetic particles, whose concentration in the natural sediment is thus difficult to quantify. In this study, we use a recently developed ultrahigh-resolution, synchrotron-based, full-field transmission X-ray microscope to study the iron-rich minerals within the clay sediment in their bulk state. We report we are able to estimate the total magnetization concentration of the giant biogenic magnetofossils to be only ~10% of whole sediment. Along with previous rock magnetic studies on the CIE clay, we suggest that most of the magnetite in the clay occurs as isolated, near-equidimensional nanoparticles, a suggestion that points to a nonbiogenic origin, such as comet impact plume condensates in what may be very rapidly deposited CIE clays.

58 GEOSCIENCES↗
DOE OSTI2014

SU-E-T-294: Simulations to Investigate the Feasibility of ‘dry’ Optical-CT Imaging for 3D Dosimetry

Purpose: To perform simulations investigating the feasibility of “dry” optical-CT, and determine optimal design and scanning parameters for a novel dry tank telecentric optical-CT 3D dosimetry system. Such a system would have important advantages in terms of practical convenience and reduced cost. Methods: A Matlab based ray-tracing simulation platform, ScanSim, was used to model a telecentric system with a polyurethane dry tank, cylindrical dosimeter, and surrounding fluid. This program's capabilities were expanded for the geometry and physics of dry scanning. To categorize the effects of refractive index (RI) mismatches, simulations were run for several dosimeter (RI = 1.5−1.48) and fluid (RI = 1.55−1.33) combinations. Additional simulations examined the effect of increasing gap size (1–5mm) between the dosimeter and tank wall, and of changing the telecentric lens tolerance (0.5°−5°). The evaluation metric is the usable radius; the distance from the dosimeter center where the measured and true doses differ by less than 2%. Results: As the tank/dosimeter RI mismatch increases from 0–0.02, the usable radius decreases from 97.6% to 50.2%. The fluid RI for matching is lower than either the tank or dosimeter RI. Changing gap sizes has drastic effects on the usable radius, requiring more closely matched fluid at large gap sizes. Increasing the telecentric tolerance through a range from 0.5°–5.0° improved the usable radius for every combination of media. Conclusion: Dry optical-CT with telecentric lenses is feasible when the dosimeter and tank RIs are closely matched (<0.01 difference), or when data in the periphery is not required. The ScanSim tool proved very useful in situations when the tank and dosimeter have slight differences in RI by enabling estimation of the optimal choice of RI of the small amount of fluid still required. Some spoiling of the telecentric beam and increasing the tolerance helps recover the usable radius.

07 ISOTOPE AND RADIATION SOURCES↗