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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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At least 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

SU-E-P-09: Radiation Transmission Measurements and Evaluation of Diagnostic Lead-Based and Lead-Free Aprons

Purpose: This study was conducted to ensure that various lead shield apron manufacturers provided accurate attenuation factors regardless of whether the apron was made of lead-based or lead-free equivalent material. Methods: A calibrated ionization survey meter was placed at chest height and 36 cm horizontally away from a solid water phantom on a simulator couch. Measurements were done with or without apron. Radiation field was set to 24cmx24cm with the phantom at 100cm source-to-surface distance. Irradiation time was set for 1 minute at voltages of 60, 80, 100 and 120 kVp. Current was set at 6mA. Results: Between 60 kVp and 120 kVp, the transmission through 0.50 mm of lead-based apron was between 1.0% and 6.5% with a mean value of 3.2% and a standard deviation (s.d.) of 1.4%. The transmissions through the 0.50 mm lead-free aprons were 1.0 % to 12.0% with a mean value of 6.1% and s.d. of 2.6%. At 120 kVp, the transmission value was 6.5% for 0.50 mm lead-based apron and 11.1% to 12.0% for 0.50 mm lead-free aprons. The radiation transmissions at 80 kVp, measured in two different 0.5 mm lead-free aprons, were 4.3% each. However, only 1.4% transmission was found through the lead-based apron. Overall, the radiation transmitted through the lead-based apron was 1/3 transmission of lead-free at 80kVp, and half value of lead-free aprons at 100 and 120 kVp. Conclusion: Even though lead-based and lead-free aprons all claimed to have the same lead equivalent thickness, the transmission might not be the same. The precaution was needed to exercise diligence in quality assurance program to assure adequate protection to staff who wear it during diagnostic procedures. The requirement for aprons not only should be in certain thickness to meet state regulation but also to keep reasonably achievable low exposure with the accurate labeling from manufacturers.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Extraction of Pm-147 from savannah river site nuclear material management programs

The Savannah River Site (SRS) processes used-nuclear fuel to recover uranium. There are many interesting fission products lost in the raffinate from this separation including 147 Pm. Here, the goal of this work was to recover 147 Pm from SRS waste solutions. A radiochemical separation method typically used for the determination of 147 Pm and 151 Sm was modified to purify significant quantities of 147 Pm from the H Canyon waste solutions. The developed scheme was tested on a small aliquot of waste solution from the processing of used fuel from the High Flux Isotope Reactor at Oak Ridge National Laboratory.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Cooperation on Improved Isotopic Identification and Analysis Software for Portable, Electrically Cooled High-Resolution Gamma Spectrometry Systems Final Report

Under a 2006 agreement between the Department of Energy (DOE) of the United States of America and the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) of France, the National Nuclear Security Administration (NNSA) within DOE and IRSN initiated a collaboration to improve isotopic identification and analysis of nuclear material [i.e., plutonium (Pu) and uranium (U)]. The specific aim of the collaborative project was to develop new versions of two types of isotopic identification and analysis software: (1) the fixed-energy response-function analysis for multiple energies (FRAM) codes and (2) multi-group analysis (MGA) codes. The project is entitled Action Sheet 4 – Cooperation on Improved Isotopic Identification and Analysis Software for Portable, Electrically Cooled, High-Resolution Gamma Spectrometry Systems (Action Sheet 4). FRAM and MGA/U235HI are software codes used to analyze isotopic ratios of U and Pu. FRAM is an application that uses parameter sets for the analysis of U or Pu. MGA and U235HI are two separate applications that analyze Pu or U, respectively. They have traditionally been used by safeguards practitioners to analyze gamma spectra acquired with high-resolution gamma spectrometry (HRGS) systems that are cooled by liquid nitrogen. However, it was discovered that these analysis programs were not as accurate when used on spectra acquired with a newer generation of more portable, electrically cooled HRGS (ECHRGS) systems. In response to this need, DOE/NNSA and IRSN collaborated to update the FRAM and U235HI codes to improve their performance with newer ECHRGS systems. Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL) performed this work for DOE/NNSA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

AMPX [Slides]

This presentation discusses and announces that AMPX is now available as open source. It additionally provides an overview about ENDF and GNDS reading in AMPX. It also discusses thermal scattering law updates and the CE and MG data libraries for SCALE 6.3.0 and SCALE 7.0.0. In addition, the presentation discusses the photonuclear sub library processing capability of AMPX.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2026

Quality Assurance of Legacy Post-Irradiation Examination Data for Metallic Fuels

The U.S. DOE-NE’s Advanced Reactor Technologies (ART) Fast Reactor Program (FRP) and the NE-4 Advanced Fuels Campaign (AFC) have jointly undertaken the qualification of the legacy post-irradiation examination (PIE) data held in the Fuels Irradiation & Physics Database (FIPD), covering metallic fuel experiments conducted in the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). In FY26 this effort reached a milestone: six major types of PIE data—contact profilometry, isotopic gamma scan, fission gas chemistry, fission gas release, laser profilometry, and neutron radiography—have been qualified for all available experiments in FIPD, and the U.S. nuclear industry can now draw on them with confidence in licensing activities for metallic fuel-based advanced fast reactors. This report documents the qualification process and the resulting status of the PIE data.

Mo, Kun↗
DOE OSTI2017

Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification

The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup, under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes. The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.

Jacobsson-Svard, Staffan↗
DOE OSTI2023

FRIB Transition to User Operations, Power Ramp Up, and Upgrade Perspectives

After project completion on scope, on cost, and ahead of schedule, the Facility for Rare Isotope Beams began operations for scientific users in May of 2022. During the first 12 months of user operations, the FRIB accelerator complex delivered 5250 beam hours, including 1528 hours to nine science experiments conducted with primary beams of 36Ar, 48Ca, 70Zn, 82Se, 124Xe, and 198Pt at beam energies >200 MeV/u; 2724 hours for beam developments, studies, and tuning; and 998 hours to industrial users and non-scientific programs using the FRIB Single Event Effect (FSEE) beam line. The ramp-up to a beam power of 400 kW is planned over a six-year period; 1 kW was delivered for initial user runs from in 2022, and 5 kW was delivered as of February 2023. Upgrade plans include doubling the primary-beam energy to 400 MeV/nucleon for enhanced discovery potential (¿FRIB 400¿). This talk reports on FRIB status and progress since SRF2021, emphasizing lessons learned during the transition from beam commissioning to machine operations, challenges and resolutions for the power ramp-up, progress with accelerator improvements, and R&D for the energy upgrade.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Aluminum Clad SNF: Technical Considerations and Challenges for Extended (>50 years) Dry Storage - 1900y

The U.S. Department of Energy (DOE) manages a significant inventory (>13 MTHM) of aluminum clad spent nuclear fuel (ASNF) using a combination of wet and dry storage at three primary sites; Hanford, Idaho National Laboratory (INL), and Savannah River Site (SRS). The majority of this ASNF is generated from U.S. and international research test reactors (RTR), examples include Advanced Test Reactor (ATR) at INL and High-Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Obama Administration suspension of the Yucca Mountain repository, issuance of the January 2013 DOE 'Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste,' delays in implementing the 'regionalization' SNF approach recommended in the 1995 'Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement,' and limited ASNF processing capability at SRS H-Canyon means that interim storage has become the de-facto strategy for managing ASNF. The 1995 Idaho Settlement Agreement requires the DOE to transfer all SNF stored at INL to interim dry storage prior to December 31, 2023 and wet storage capacity at the SRS is nearing capacity. This situation was recognized by the DOE Spent Nuclear Fuel Working Group (SNFWG) and at its June 2016 meeting formed a ASNF sub-working group comprised subject matter experts to evaluate this area and provide a report of findings. The ASNF sub-working group met many times over several month period to review the engineering and technical basis of existing and planned DOE ASNF dry storage facilities and additional studies on the behavior of ASNF under different conditions, including during reactor service. The ASNF sub-working group focused its review on RTR SNF. Based on these reviews, the ASNF sub-working group identified five knowledge gaps and technical data needs that should be addressed to ensure viability and safety of extended, greater than 50 years, sealed or vented interim dry storage of ASNF. These five areas include; behavior/chemistry of aluminum oxyhydroxide layer in dry storage, radiolytic gas generation properties of aluminum oxyhydroxide layers, effect of episodic breathing and radiolytic generation of corrosive gases during storage, performance of ASNF in existing dry storage systems, and effects of high temperature (>1000 C) drying on chemistry and behavior of aluminum oxyhydroxide layers. The ASNF sub-working group also recommended development of an action plan to identify technical and engineering activities needed to address the knowledge gaps and technical data needs. Results of the ASNF sub-working group are documented in 2017 report 'Aluminum Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 years) Dry Storage,' DOE-ID/RPT-1575. The INL assembled a team comprised of twenty INL and Savannah River National Laboratory SNF management and subject matter experts to develop this action plan. Over the course of 2 months the team worked to develop the action plan. The plan identified and recommended six technical and engineering tasks needed to address the previously identified knowledge gaps and technical data needs. DOE Environmental Management Office of Technology Development is providing funding to the INL to execute these tasks. This paper will present and discuss the results of the ASNF sub-working group, tasks identified in the action plan, and experimental activities underway at INL and SRNL to address the action plan. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

SU-E-T-173: Evaluation of the MicroSTAR Reader for the OSLD System for Remote Dosimetry Audits at the IROC HQAC

Purpose: Analysis of the performance of the microSTARii reader for optically stimulated luminescence dosimeters (OSLD) used by the IROC Houston Quality Assurance Center (IROC HQAC) for external audits compare to the characteristics of the Microstar reader. Methods: The IROC HQAC uses the Microstar reader for the OSLD program for verification of output of photon, electron and proton beams. The calculation of dose from the OSLD system is based on a group of factors defined at time of the commissioning of a batch of detectors. Factors like system sensitivity (SS), depletion (KD), element correction factor (ECF), linearity (KL), energy correction (KE). The new microSTARii unit presents many hardware and software upgrades that were considered useful for this program. Based on these changes many factors, that were considered reader dependent, were revised in order to analyze the effect of the new reading process. The SS, KD, ECF and KL were evaluated and compared with data defined based on reading done on Microstar reader. Results: The SS is reader specific and specified at 100 cGy dose level. This value is define per reading session and monitored over time. The KD factor was found to be different because of reading procedure are different. The ECF values changed for a group of nano dots compare to values defined based on reading done on the Microstar reader. The KL was defined for the reader. Conclusion: The new microSTARii reader presents new features that improve the efficiency of the OSLD program at the IROC HQAC. New characterization is needed before final implementation is done to match the requirements of the existing OSLD system defined for the Microstar reader. Changes in uncertainty of the results has not been analized.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Enhancement of STAYSL_PNNL with IRDFF/V1.05 to 60 MeV

The STAYSL PNNL program performs a least-squares adjustment of the neutron spectrum based on the IRDFF cross section library and user-defined activation data and a starting neutron spectrum. All known covariances are included. Required corrections are included for the irradiation history, gamma absorption, neutron burn-up, and neutron self-shielding using separate utility programs prior to the execution of STAYSL PNNL. Several new neutron cross section libraries based on IRDFF V1.05 to 60 MeV are provided with the program including energy group structures of 69 (WIMS), 129 (14 MeV), 140 (prior 100 groups extended to 60 MeV), 175 (Vitamin J), 640 (old SANDII to 20 MeV), and 725 (old Sand II extended to 60 MeV) energy groups. Twelve new reactions were included in IRDFF V1.05. At higher neutron energies, calculations demonstrate the need for activation cross sections from natural targets such as Ti(n,x)46Sc rather than only isotopic targets such as 46Ti(n,p)46Sc and 47Ti(n,x)46Sc. The effects of such reactions increase dramatically with increasing neutron energies and cannot be neglected. Some critical reactions such as 56Fe(n,x)54Mn are not currently included in IRDFF and prevent the use of the 54Fe(n,p)54Mn reaction at higher neutron energies. It is recommended that natural element cross sections be included in future releases of IRDFF for a list of reactions that suffer from reactions on multiple isotopes that lead to the same activation product. Spectral adjustments are illustrated in Be(d,n) thick target neutron fields at deuteron energies of 30 and 40 MeV.

reactor dosimetry, neutron spectral adjustment↗
DOE OSTI2019

Atom Probe Tomography for understanding microstructural and chemical changes in nuclear materials

Atom Probe Tomography (APT) is routinely employed to determine 3D microstructure and chemistry changes in structural materials used in reactor to understand irradiation effect. These findings help to correlate changes to mechanical and physical properties of nuclear materials which are subjected to extreme reactor conditions. With safety and economics of a reactor in mind, it is essential to be able to predict the changes in the fuel composition and microstructure as a function of burnup. Also, there is current need to develop advanced fuels with improved properties and performance which requires the experimental data on changes in microstructure, physical and chemical properties after irradiation for feedback process. A novel approach is presented to determine the local burnup in irradiated fuels using isotopic quantification obtained by Atom Probe Tomography (APT). Considering the volume of sample used (<100?µm3) for APT experiments using the lift-out process in a scanning electron microscope equipped with a Focused Ion Beam (FIB), the presented method determines the local burnup from a nuclear fuel, where a minimal amount of waste is produced. In this work, three samples were analyzed with different burnup conditions to quantify the isotopes of 235U, 236U, and 238U for burnup calculation in the irradiated metallic U–7Mo dispersion fuel. The calculated values were found in excellent agreement with MCNP calculations. With regard to structural materials, one of the primary factors limiting the utilization of fuel rods of Zr alloys is hydrogen pick up during oxidation. Past experimental studies have shown that the hydrogen pickup in Zirconium alloys is influenced by the alloy composition as well by irradiation temperature and neutron dose. Oftentimes, it is not feasible to carry out comprehensive studies on understanding the microstructural evolution in neutron irradiated materials since in reactor irradiation programs are extremely costly, lengthy and logistically complicated. The use of ion irradiation as a substitute has grown in recent years for several reasons including less radioactivity, faster irradiations and lower costs. In the framework of the MUZIC 3 project, samples of Zircaloy-4 have been proton and Zr2+ irradiated at the Michigan Ion Beam Lab (MIBL) facility with aim to create a microstructure that is similar to that observed after in-reactor service. The temperature of the environment is adjusted to compensate for the higher dose rate, as has been done previously. As a first step in this project the irradiated microstructures have been characterized using Transmission Electron Microscopy and Atom Probe Tomography to elucidate the microstructural changes, including precipitate amorphization and dissolution and the formation of dislocation loops. This presentation plans to highlight initial finding of this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

LANL Environmental ALARA Program Status Report for CY 2015

Los Alamos National Laboratory (LANL) ensures that radiation exposures to members of the public and the environment from LANL operations, past and present, are below regulatory thresholds and are as low as reasonably achievable (ALARA) through compliance with DOE Order 458.1 Radiation Protection for the Public and the Environment, and LANL Policy 412 Environmental Radiation Protection. In 2007, a finding (RL.2-F-1) and observation (RL.2-0-1) in the NNSA/ LASO report, September 2007, Release of Property (Land) Containing Residual Radioactive Material Self-Assessment Report, indicated that LANL had no policy or documented process in place for the release of property containing residual radioactive material. In response, LANL developed PD410, Los Alamos National Laboratory Environmental ALARA Program. The most recent version of this document became effective on September 28, 2011. The document provides program authorities, responsibilities, descriptions, processes, and thresholds for conducting qualitative and quantitative ALARA analyses for prospective and actual radiation exposures to the public and t o the environment resulting from DOE activities conducted on the LANL site.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

S&TR September 2025: Computing Grand Challenge Turns 20

Livermore’s Computing Grand Challenge Program enters its 20th year with more unclassified high-performance computing (HPC) power than ever before. This unique, peer-reviewed competition awards HPC allocations on top supercomputers to multidisciplinary teams with high-impact projects. The Grand Challenge encourages researchers to innovate, pushes scientific discovery to new heights, improves the Laboratory’s HPC capabilities, and extends HPC accessibility to collaborators. Awardees must adapt to successive generations of HPC hardware and learn to run simulations at scale. The feature article spotlights three Grand Challenge teams whose research broke new ground in key scientific pursuits—the essence of dark matter, explosion-generated seismic waves, and protein interactions linked to cancer—while underscoring the importance of academic partnerships and considering the program’s future.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Personnel Qualification and Certification Program in Pressure Decay Leak Testing on the RT-100 Type (B) Cask - 17229

The requirements for qualifications and certification of leak test personnel on Type (B) shipping casks have been more clearly defined with the recent release of ANSI N14.5-2014, Leak Tests on Radioactive Packages for Shipment. The latest version creates a standardized approach for qualifying and certifying personnel to conduct pre-shipment leakage tests on Type (B) packages. The U.S Nuclear Regulatory Commission (NRC) is currently updating the Regulatory Guide R.G 7.4, Leakage Tests on Packages for shipment of Radioactive Material, to endorse ANSI N14.5-2014. According to ANSI N14.5-2014, leakage testing shall be performed in accordance with written procedures by qualified and certified personnel in accordance with the requirements of ASNT SNT-TC-1A-2006, Recommended Practice for Personnel Qualification and Certification in Nondestructive Testing. This is a significant modification to the old standard, ANSI N14.5-1997, which simply states that pressure leak testing shall be conducted by trained and certified personnel without reference to any specific certification standards or regulations. SNT-TC-1A published by the American Society for Nondestructive Testing (ASNT) has several advantages to employers; namely, it allows for a Limited NDT Level II certification program. Limited certifications are especially valuable for personnel who possess limited experience and wish to become a certified Limited NDT Level II in a specific technique on a specific cask using a specific procedure and equipment. Robatel Technologies recently developed a comprehensive qualification and training program based on the recommendations of SNT-TC-1A for Exelon Corporation to certify their employees to a Limited NDT Level II in Pressure Decay Leak Testing on the RT-100 Cask. The project scope included the development of a training and qualification program (Written practice), writing a pressure decay leak test procedure, the design of a pressure decay leak test manifold, and the delivery of classroom training. The written practice outlines the requirements to ensure that personnel participating in the program have the necessary qualification, experience, and training to become certified in the leak test method. The program will give Exelon the liberty to conduct their own leak testing on the RT- 100 without the need to rely on outside resources for leak testing. When fully implemented, the program is expected to improve productivity and eliminate schedule delays, along with substantial economic benefits. Following the appropriate standards and industry best practices, the project was successfully completed and scheduled to be fully implemented by Exelon's Nuclear Stations by the end of 2016. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Microstructure of In-Core Molten Salt Corrosion Hastelloy N{sup R} and 316 Stainless Steel

Since 2011, many materials have been discussed for potential application in Fluoride salt-cooled High-temperature nuclear Reactors (FHRs), which are emerging as leading reactor concept for the next generation of nuclear reactors thanks to the benefits of a high degree of passive safety, high thermal efficiency, and low spent fuel volume. Among these discussed materials, Hastelloy N{sup R} and 316L stainless steel have been selected as the most likely candidate alloys for FHRs. To provide direct experience in integrating a real system using molten FLiBe salt for reactor operation, these two candidate alloys were experimentally tested in Li-7 enriched Li{sub 2}BeF{sub 4} (FLiBe) molten salt in the core of MIT Research Reactor (MITR) at 700 deg. C for 1000 hours. After in-core corrosion testing, some basic post-irradiation examinations (PIE) including weight change per unit area and surface profile as well as surface morphology were conducted in the MIT Nuclear Reactor Laboratory (NRL) and compared with out-of-reactor corrosion samples. The deep understanding of these alloys in high-temperature molten salt with strong neutron irradiation requires further microstructural analysis such as electron microscopy and energy dispersive spectroscopy on the surface and cross-section of the in-core corrosion tested samples. The challenge in the microstructural analysis is due to the high radioactivity from the gamma emitting isotopes formed in alloys during neutron irradiation. Additionally, the facilities that are authorized for radioactive materials are very limited on campus. In this study, the microstructure of the in-core corrosion tested Hastelloy N{sup R} and 316L stainless steel specimen was analyzed by using the facilities at the Center for Advanced Energy Studies (CAES) at Idaho National Laboratory (INL) through the Nuclear Science User Facilities (NSUF) Rapid Turnaround Experiment (RTE) program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Analysis of Radiological Release From Fueled Irradiation Experiments During Manual Handling (Slides)

Irradiation experiments are manually handled at the Advanced Test Reactor (ATR) by qualified operators using long handled tools. Standard handling evolutions include insertion and removal from the reactor vessel, transfer to/from a storage location or cask in the ATR canal, and handling/reconfigurations at a canal working tray. Such routine handling has the potential to result in mechanical damage to the experiment boundary which is credited to retain fission products. Since damage can occur due to operator error, this is an anticipated occurrence. This work determines the radiological consequences to receptors inside the ATR facility, as well as public receptors. Given the wide variety of fuel types tested at the ATR, a generic approach to the analysis is taken. The radiological inventory is determined to bound a variety of fuel types (e.g., ceramic and molten fuel matrices) and fission powers on the U-235 enrichment spectrum. The source term analysis considers different release fractions to bound different fuel types and burnups. The postulated handling events occur underwater; thus, retention of the released isotopic content is considered within the canal water column. Retention of radionuclides in the water column is determined using a modified approach of United States (US) Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.183. Radiological dose to the facility receptor is determined using a compartment release model. Dose to the public receptor is determined using atmospheric dispersion models using site specific atmospheric conditions with the use of the Radiological Safety Analysis Computer (RSAC) program version 7.2.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2014

Sci—Fri PM: Topics — 05: Experience with linac simulation software in a teaching environment

Medical linear accelerator education is usually restricted to use of academic textbooks and supervised access to accelerators. To facilitate the learning process, simulation software was developed to reproduce the effect of medical linear accelerator beam adjustments on resulting clinical photon beams. The purpose of this report is to briefly describe the method of operation of the software as well as the initial experience with it in a teaching environment. To first and higher orders, all components of medical linear accelerators can be described by analytical solutions. When appropriate calibrations are applied, these analytical solutions can accurately simulate the performance of all linear accelerator sub-components. Grouped together, an overall medical linear accelerator model can be constructed. Fifteen expressions in total were coded using MATLAB v 7.14. The program was called SIMAC. The SIMAC program was used in an accelerator technology course offered at our institution; 14 delegates attended the course. The professional breakdown of the participants was: 5 physics residents, 3 accelerator technologists, 4 regulators and 1 physics associate. The course consisted of didactic lectures supported by labs using SIMAC. At the conclusion of the course, eight of thirteen delegates were able to successfully perform advanced beam adjustments after two days of theory and use of the linac simulator program. We suggest that this demonstrates good proficiency in understanding of the accelerator physics, which we hope will translate to a better ability to understand real world beam adjustments on a functioning medical linear accelerator.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Beyond Californium—A Neutron Generator Alternative for Dosimetry and Instrument Calibration in the U.S.

Many evaluations of neutron survey instruments, area monitors and personal dosimeters rely on neutron reference radiations from 252Cf, a source of spontaneous fission neutrons that produces high dose rates from an almost point source geometry. Its characteristics are generally more favorable when compared to (a,n), photoneutron and reactor-based sources of neutrons. The source is typically used in two standardized configurations: unmoderated, to yield a fission energy spectrum, or placed within a heavy-water moderating sphere to produce a softened spectrum that is more appropriate for evaluating devices used in nuclear power plant work environments. The U.S. Department of Energy (DOE) 252Cf Loan/Lease Program was terminated in 2009, and the replacement of high-activity sources is increasingly cost prohibitive for laboratories that formerly benefited from that program, including the Pacific Northwest National Laboratory (PNNL) and National Institute of Standards and Technology (NIST). Neutron generators, based on the D-T and D-D fusion reactions, are becoming economically competitive with 252Cf and are recognized internationally as important calibration and test standards. Researchers from NIST and PNNLare jointly considering the practicality and technical challenges of implementing neutron generators as calibration standards in the U.S. This article reviews characteristics of isotope-based sources, leading to rationale behind increasing favor of electronically-generated neutron options. The evaluation of a D-T system at PNNL has revealed primary challenges in adapting generators to the task of calibration and testing in which neutron characteristics must be accurately determined. Concepts for modifying the generated neutron spectra to achieve particular targeted spectra, simulating 252Cf or workplace environments, are discussed.

Calibration, californium-252, Dosimeter Irradiatio↗