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SU-F-18C-08: A Validation Study of a Commercially Available Software Package's Absorbed Dose Estimates in a Physical Phantom

Purpose: This study assesses the accuracy of the absorbed dose estimates from CT scans generated by Monte Carlo (MC) simulation using a commercially available radiation dose monitoring software program. Methods: Axial CT studies of an anthropomorphic abdomen phantom with dose bores at a central location and 4 peripheral locations were conducted using a fixed tube current at 120 kV. A 100 mm ion chamber and a 0.6 cc ion chamber calibrated at diagnostic energy levels were used to measure dose in the phantom at each of the 5 dose bore locations. Simulations using the software program's Monte Carlo engine were run using a mathematical model of the anthropomorphic phantom to determine conversion coefficients between the CTDIvol used for the study and the dose at the location of the dose bores. Simulations were conducted using both the software's generic CT beam model and a refined model generated using HVL and bow tie filter profile measurements made on the scanner used for the study. Results: Monte Carlo simulations completed using the generalized beam model differed from the measured conversion factors by an absolute value average of 13.0% and 13.8% for the 100 mm and 0.6 cc ion chamber studies, respectively. The MC simulations using the scanner specific beam model generated conversion coefficients that differed from the CTDIvol to measured dose conversion coefficients by an absolute value average of 7.3% and 7.8% for the 100 mm and 0.6 cc ion chamber cases, respectively. Conclusion: A scanner specific beam model used in MC simulations generates more accurate dose conversion coefficients in an anthropomorphic phantom than those generated with a generalized beam model. Agreement between measured conversion coefficients and simulated values were less than 20% for all positions using the universal beam model.

07 ISOTOPE AND RADIATION SOURCES↗

Fission Product and Actinide Distribution in the SiC Layer of Fissioned Triso Fuel Particles

Tri-structural Isotropic (TRISO) encapsulated fuel particles are used in next generation high-temperature gas-cooled nuclear reactors. The integrity of the fuel over a wide range of temperatures is integral to the safety of an operating nuclear reactor. It has been observed that fission products are released at high temperatures that simulate hypothesized accident scenarios. The particles under study were from the AGR-1 fuel development and qualification program. They consist of an enriched UCO fuel kernel surrounded by a dense carbon buffer layer (100 {mu}m), an inner pyrolytic carbon (IPyC) layer (40 {mu}m), a silicon carbide (SiC) layer (35 {mu}m), and an outer pyrolytic carbon (OPyC) layer (40 {mu}m). The SiC layer acts as the main diffusion barrier to fission products, but has been demonstrated to fail due to release of Ag through intact particles and local corrosion from metallic fission products such as Pd. In this work, we use transmission electron microscopy and diffraction techniques to study the structure and dispersion of release products within the SiC of TRISO particles as a function of annealing temperature used to simulate margin temperatures for in-reactor accident scenarios. In order to improve our understanding of fission product retention as a function of temperature, we have the unique opportunity to examine irradiated, fissioned TRISO particles to study post-irradiation effects on these materials. This data will contribute to current models used to predict the behavior of fission products in the SiC layer of TRISO fuels.

07 ISOTOPE AND RADIATION SOURCES↗

SU-E-T-495: Neutron Induced Electronics Failure Rate Analysis for a Single Room Proton Accelerator

Purpose: To determine the failure rate as a function of neutron dose of the range modulator's servo motor controller system (SMCS) while shielded with Borated Polyethylene (BPE) and unshielded in a single room proton accelerator. Methods: Two experimental setups were constructed using two servo motor controllers and two motors. Each SMCS was then placed 30 cm from the end of the plugged proton accelerator applicator. The motor was then turned on and observed from outside of the vault while being irradiated to known neutron doses determined from bubble detector measurements. Anytime the motor deviated from the programmed motion a failure was recorded along with the delivered dose. The experiment was repeated using 9 cm of BPE shielding surrounding the SMCS. Results: Ten SMCS failures were recorded in each experiment. The dose per monitor unit for the unshielded SMCS was 0.0211 mSv/MU and 0.0144 mSv/MU for the shielded SMCS. The mean dose to produce a failure for the unshielded SMCS was 63.5 ± 58.3 mSv versus 17.0 ±12.2 mSv for the shielded. The mean number of MUs between failures were 2297 ± 1891 MU for the unshielded SMCS and 2122 ± 1523 MU for the shielded. A Wilcoxon Signed Ranked test showed the dose between failures were significantly different (P value = 0.044) while the number of MUs between failures were not (P value = 1.000). Statistical analysis determined a SMCS neutron dose of 5.3 mSv produces a 5% chance of failure. Depending on the workload and location of the SMCS, this failure rate could impede clinical workflow. Conclusion: BPE shielding was shown to not reduce the average failure of the SMCS and relocation of the system outside of the accelerator vault was required to lower the failure rate enough to avoid impeding clinical work flow.

07 ISOTOPE AND RADIATION SOURCES↗

Comparison of Burnup Credit Uncertainty Quantification Methods

The recent revision of Interim Staff Guidance (ISG-8 revision 3), issued by the US Nuclear Regulatory Commission, on burnup credit for criticality safety analysis of pressurized water reactor spent fuel in transportation and storage casks expands the nuclides that can be credited in the fuel composition to include minor actinides and fission products. The technical basis for development of the uncertainties associated with calculated nuclide concentrations in spent nuclear fuel that are recommended in ISG-8 was established through the use of measured isotopic data acquired from destructive radiochemical assay data for spent fuel samples. These data provide a direct experimental basis for determining uncertainties in the calculated nuclide concentrations. To provide additional insight into code uncertainties resulting from nuclear data and also the level of conservatism associated with using experimental assay benchmark data for uncertainty evaluation, a new uncertainty analysis tool in the SCALE code system that performs stochastic sampling of nuclear data has been applied to representative burnup credit calculations for spent fuel pool and dry cask storage applications methodology. The nuclear data uncertainty analysis capabilities in Sampler can account for uncertainty contributions from cross sections, fission yields, and decay data by using correlated sampling of covariance files developed for all nuclear data used by SCALE. These uncertainties associated with nuclear data evaluations are compared to corresponding values developed in ISG-8 that are based on destructive assay data, as well as to values obtained using an alternate methodology that was developed by the Electric Power Research Institute based on reactor core simulation data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Belgian program for low and intermediate short lived waste management: from 1985 to license application

On 31 January 2013, ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, has introduced the construction and operation license application for the surface disposal facility for category A waste at Dessel. The objective of this paper is to present the development of this national programme. First the key milestones and their impact on the development of the project will be highlighted. This includes in particular: 1) the turning point in 1994, when ONDRAF/NIRAS converts its purely technic-scientific work method into an approach of participation that allows for societal aspects to be taken into account and 2) the decision of the government in 2006 in favour of the surface disposal of category A waste in Dessel. Then the paper presents the pillars of the long-term safety of a surface disposal and the principles for site selection and how they were applied to the surface disposal facility for category A waste at Dessel. The pillars of the long-term safety of a surface disposal are: (1) the passive confinement and isolation provided by the Engineering Barrier System (2) the contribution to this passive isolation and confinement by the site (3) the limitation of the radioactive source term, and (4) the control and monitoring measures. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tracking radioactivity content in the spent fuel pool in PWRS - 15150

Over the last decade, significant progress has been made in managing radioisotopes generated in the PWR primary systems. These include limiting the use of cobalt containing materials, along with improved fuel performance and zinc injection to passivate the corrosion layers to limit the accumulation of Co-58 and Co-60 on primary coolant surfaces. Attention here is directed to the spent fuel pool where accumulation of spent fuel and activated material represents a growing source of mobile radioactivity supplemented with periodic exchanges of water from the reactor cooling system. Characterization of this source is increasingly important to effective management of disposal options. This paper examines the radioactive resin source term as determined from routine sampling of the coolant and fuel pool liquids. Targeted resin stream sampling and radiochemical analyses conducted at Diablo Canyon serve to inform and provide context for this examination. This examination is supplemented with computer modelling using the DW James Consulting computer program, 3R-SCAN. At most operating plants daily reactor coolant samples are counted using gamma spectroscopy and recorded. These samples serve to track fuel performance and maintain water chemistry. A secondary use for this data is to maintain a day to day tracking of releases of radioisotopes important to waste disposal. Given the coolant concentrations, plant operating history, and corroborative mathematical models and sampling, accumulation of these radioisotopes in coolant cleanup system ion-exchangers can be predicted. Fuel pool sampling is less frequent, but at the same time, things change more slowly there. The fuel pool contains about ten times the volume of the reactor coolant but its cleanup system process rate is comparable to reactor coolant letdown, i.e. 5.047 - 8.833 liters per second [lps] (80-140 gallons per minute [gpm]). A spent fuel pool clean-up demineralizer flow rate is typically about 5.047 lps (80 gpm). Concentrations of longer lived radioisotopes in the spent fuel pool can be comparable to those in reactor coolant. Releases of corrosion products and fission products from the stored fuel occur at a low rate but given the overall inventory can be significant. Recognized scaling factor ratios, particularly those to Co-60, are skewed by the aging of the material in the pool. An additional consideration stems from refueling operations themselves. Prior to refueling, it is common practice to conduct a forced oxidation of the corrosion films within the reactor. This spikes the activity within the reactor coolant. Most of this activity is removed by the reactor coolant purification system prior to removing the reactor head. This paper examines the accumulation of activity in the fuel pool purification ion exchange resins and how it impacts the overall classification of those ion exchange resins. The following discussion focuses on the transport of nickel and cobalt isotopes particularly as it relates to determination of Ni-63 which has become increasingly prominent in disposal classification. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Project 57 Air Monitoring Report: January 1 through December 31, 2017

On April 24, 1957, the Atomic Energy Commission (AEC) (now the Department of Energy [DOE]) conducted the Project 57 safety experiment in western Emigrant Valley northeast of the Nevada National Security Site (NNSS, formerly the Nevada Test Site) on lands withdrawn by the Department of Defense (DOD) for the Nevada Test and Training Range (NTTR). The test was undertaken to develop (1) a means of estimating plutonium distribution resulting from a non-nuclear detonation; (2) biomedical evaluation techniques for use in plutonium-laden environments; (3) methods of surface decontamination; and (4) instruments and field procedures for prompt estimation of alpha contamination (Shreve, 1958). Although the test did not result in the fission of nuclear materials, it did disseminate plutonium across the land surface. Following the experiment, the AEC fenced the contaminated area and returned control of the surrounding land to the DOD. Various radiological surveys were performed in the area and in 2007, the DOE expanded the demarked Contamination Area (CA) by posting signs 200 to 400 ft (60 to 120 m) outside of the original fence. Plutonium in soil attaches preferentially to smaller particles (Tamura, 1985; Friesen, 1992; Murarik et al., 1992; and Misra et al., 1993). Therefore, redistribution of soil particles by wind (dust) and water are the mechanisms most likely to transport plutonium beyond the boundary of the Project 57 CA. Monitoring was implemented in 2011 by Desert Research Institute (DRI) to determine if radionuclide contamination was detectable in samples of airborne dust and to characterize meteorological and environmental parameters that influence dust transport. Initially, two monitoring stations consisting of radiological, meteorological, and dust sampling equipment were installed near the southeast and northeast corners of the CA. In January 2015, the original monitoring stations were dismantled and moved farther to the west along the CA boundary. This move was made to place the monitoring stations downwind of ground zero and the High Contamination Area (HCA) during the dominant northerly and southerly winds. Samples of particles suspended in the air are collected every two weeks and submitted for laboratory assessments of gross alpha and gross beta radioactivity, and for determinations of gamma-emitting radionuclides. In calendar year (CY) 2017, the mean gross alpha concentration at monitoring station P57-4 was approximately 4.25 times the value at monitoring station P57-3. In CY2015 and CY2016, the difference in mean gross alpha concentration at the two stations was approximately 1.5. The mean gross alpha concentrations at P57-3 and P57-4 were approximately 1.9 and 7.9 times, respectively, the mean gross alpha concentrations at the surrounding Community Environmental Monitoring Program (CEMP) stations. The minimum gross alpha concentrations reported for the Project 57 stations are in the range of the minimum values observed at the CEMP stations. However, maximum gross alpha concentrations are greater than the maximum concentrations observed at the surrounding CEMP stations. Gamma spectroscopy analyses resulted in the detection of seven radionuclides. Except for americium-241 (Am-241), all reported radionuclides occur naturally in the environment. The Am-241 detection by gamma spectroscopy was confirmed by alpha spectroscopy. During each quarter of CY2017, four suspended particulate matter samples were selected for alpha spectroscopy analysis. Two were selected each quarter from the regular biweekly samples collected at each of the two monitoring stations. Of the eight samples from P57-3, one was reported to have plutonium-238 (Pu-238) above the minimum detectable activity (MDA) and five were reported to have plutonium 239/240 (Pu-239/240) above the MDA. Of the eight samples from P57-4, three were reported to have Pu-238 above the MDA and six were reported to have Pu-239/240 above the MDA. Soil material is also transported by saltation, which is a wind driven phenomena that bounces sand-sized soil particles across the land surface because they are too heavy to be suspended in the air. Samples of particles transported by saltation were collected downwind and upwind of the CA at both monitoring stations in May 2017 after a 190-day deployment and again in September 2017 after a 114-day deployment. Mass of the saltation samples was determined after oven drying. The September sample had greater mass than the May sample. The same relationship was observed for the suspended particulate matter samples during periods of saltation sample collection. The difference in the collected mass appears to be because of the amount of soil moisture. Higher soil moisture exerts additional isotopic bonds between soil particles and water molecules, and between water molecules that envelop soil particles, making it more difficult for wind to raise the particles as dust. In general, samples collected from saltation traps facing south contained more material than samples collected from traps facing north. This suggests that although saltation material may have been moving back and forth in conjunction with the two dominant wind directions, there was a net trend for saltation material to be transported toward the north. Radiological analyses of the saltation samples show that the smaller particle size faction (<63 μm) generally produced the highest activity levels of Am-241, Pu-238, and Pu-239/240. In general, the isotope activity levels decreased as the particle size increased. These results are consistent with the expectation that the greatest potential for transport of radiological contamination is with the smaller particles. Thermoluminescent dosimeters (TLDs) indicated that the average annual external radioactivity dose at the monitoring stations is slightly higher than the dose determined at the surrounding CEMP stations, but approximately half of the estimated national average dose received by the general public from exposure to natural sources. The TLDs at the Project 57 monitoring stations are exposed to both natural sources (terrestrial and cosmic) and radioactive releases from the Project 57 CA. Winds in excess of approximately 15 mph (24 km/hr) begin to generate dust movement by saltation or suspension, as determined by the monitoring instruments. Saltated particles, PM10 (i.e., inhalable dust), and PM2.5 (i.e., fine particulate dust) exhibit an approximately exponential increase with increasing wind speed. The greatest concentrations of dust occurred for winds exceeding 20 mph (32 km/hr). When the wind/dust analysis was performed for winds separated into the dominant northerly and southerly wind directions, it was evident that the northerly winds generated the higher PM10 concentrations and had the dominate influence on the PM10 concentration determined for all wind speeds when all wind directions were evaluated together. During the year, winds in excess of 20 mph (32 km/hr) occurred approximately one percent of the time. Although winds sufficient to generate dust occur at the Project 57 site, they are infrequent and of short duration. A preliminary assessment of individual wind events suggests that dust generation in response to specific wind conditions is highly variable. Wind speeds that would be expected to generate noticeable dust may not do so in every instance. Additionally, saltation transport in response to wind conditions was less common than expected. These variations were likely because of the influence of meteorological and environmental parameters other than wind.

54 ENVIRONMENTAL SCIENCES↗

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluating Optical Techniques to Characterize Solid State Samples for the Pu-238 Supply Program

This report documents steps that have been taken to establish advanced characterization capabilities for solid-state samples to support the 238 Pu Supply Program. The program has already invested significantly in spectroscopy and online monitoring development for liquid processes. Many of the data, analytical methods, and equipment could also be leveraged for the analysis of Np/Pu oxides and Np pellets to help classify material properties and link them to operations. The techniques described in this work include diffuse reflectance spectroscopy (DRS), Raman spectroscopy, and laser-induced breakdown spectroscopy (LIBS). Each technique is sensitive to various chemical and/or elemental species and can provide complimentary information to more standard methods such as powder x-ray diffraction and scanning electron microscopy. These options are remotely deployable via fiber-optic cables. DRS and Raman spectroscopy are already approved for glove box and hot cell work, but LIBS requires US Department of Energy approval for incorporation. Raman spectroscopy does require laser shielding, which means that DRS is the easiest method to implement. However, each technique provides complimentary data, and the most robust sample characterization could be achieved when all three, among other methods, are harnessed together. This report documents some of the progress made testing these capabilities, some of the challenges, and next steps to optimize the technology and integration strategies.

07 ISOTOPE AND RADIATION SOURCES↗

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↗

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↗

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↗

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↗

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↗

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↗

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↗

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↗

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 ↗