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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

ITER Neutronic Challenges for Upper Port 14

The future of magnetically confined nuclear fusion facilities is extremely promising, yet there are extensive challenges that still have to be addressed. These challenges range significantly, but one of the more difficult challenges is nuclear shielding and nuclear safety. The challenge of adequately shielding neutrons is especially prevalent at ITER. Throughout the lifetime of ITER there will be large quantities of high energy (14.1- MeV) neutrons generated, with the expected neutron lifetime fluence to be around 0.3 MW.a/m{sup 2}. For such a large neutron fluence, shielding is a very important factor for the operation of ITER. Not only the shielding of the people that will be supporting the facility, but also to the components that surround the plasma. Without the many diagnostics that are planned for ITER operation, ITER would not be able to function. But with the inclusion of the diagnostics and the many apertures required, an interesting neutronics environment is created. This environment is like nothing we have seen or are currently able to create on the earth. Due to this unknown, significant time and effort is being placed on the research, design, and analysis of the ITER neutronics environment. A neutron has an interesting life after birth within the plasma on ITER. There are many different paths and reactions that can take place; so many, that trying to simplify the problem has brought significant unknowns into the analysis. Due to unique material compositions, neutron streaming paths, and neutronics requirements, the simplest of ITER neutronics calculations have to be thought out in great detail. Placing a new hole for a diagnostic creates a new streaming path; internal components must survive a high fast neutron flux, while external components face a completely different neutron energy spectrum of a high flux; material activation within human accessible areas must be maintained to an allowable level. As the complexity of the ITER device grows, so does the complexity of the neutronics challenges.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Evaluation of a new commercial Monte Carlo dose calculation algorithm for electron beams

Purpose: In this report the authors present the validation of a Monte Carlo dose calculation algorithm (XiO EMC from Elekta Software) for electron beams. Methods: Calculated and measured dose distributions were compared for homogeneous water phantoms and for a 3D heterogeneous phantom meant to approximate the geometry of a trachea and spine. Comparisons of measurements and calculated data were performed using 2D and 3D gamma index dose comparison metrics. Results: Measured outputs agree with calculated values within estimated uncertainties for standard and extended SSDs for open applicators, and for cutouts, with the exception of the 17 MeV electron beam at extended SSD for cutout sizes smaller than 5 × 5 cm{sup 2}. Good agreement was obtained between calculated and experimental depth dose curves and dose profiles (minimum number of measurements that pass a 2%/2 mm agreement 2D gamma index criteria for any applicator or energy was 97%). Dose calculations in a heterogeneous phantom agree with radiochromic film measurements (>98% of pixels pass a 3 dimensional 3%/2 mm γ-criteria) provided that the steep dose gradient in the depth direction is considered. Conclusions: Clinically acceptable agreement (at the 2%/2 mm level) between the measurements and calculated data for measurements in water are obtained for this dose calculation algorithm. Radiochromic film is a useful tool to evaluate the accuracy of electron MC treatment planning systems in heterogeneous media.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Idaho National Laboratory CY 2024 National Emission Standards for Hazardous Air Pollutants Analysis, Methodology, and Results for Radionuclides

This report documents the methodology and results for calculating the effective dose equivalent (EDE) to the maximally exposed individual (MEI) from atmospheric radionuclide emissions from Idaho National Laboratory (INL) sources in Calendar Year (CY) 2024. The calculations were performed in accordance with requirements in Code of Federal Regulations (CFR), Title 40, “Protection of the Environment,” Part 61, “National Emission Standards for Hazardous Air Pollutants (NESHAPs),” Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” (40 CFR 61, Subpart H). UDFs were calculated using the computer model CAP88-PC for unit (1 Ci/yr) emission rates at INL Site facilities and INL in-town (Idaho Falls) facilities and stored in Microsoft Access databases. The UDFs—in this case, mrem/Ci—were then combined with radionuclide-specific release (emission) rates for each facility-specific source to compute doses at predetermined public receptor locations, including the MEI locations. This report contains the dose results and a description of the methodology and tools used to compute the doses.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2014

WE-G-19A-01: Radiologists and Medical Physicists: Working Together to Achieve Common Goals

It is vitally important that medical physicists understand the clinical questions that radiologists are trying to answer with patient images. Knowledge of the types of information the radiologist needs helps medical physicists configure imaging protocols that appropriately balance radiation dose, time, and image quality. The ability to communicate with radiologists and understand medical terminology, anatomy, and physiology is key to creating such imaging protocols. In this session, radiologists will present clinical cases and describe the information they are seeking in the clinical images. Medical physicists will then discuss how imaging protocols are configured. Learning Objectives: Understand the types of information that radiologists seek in medical images. Apply this understanding in configuring the imaging equipment to deliver this information. Develop strategies for working with physician colleagues.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

MO-DE-BRA-04: The CREATE Medical Physics Research Training Network: Training of New Generation Innovators

Purpose: Over the past century, physicists have played a major role in transforming scientific discovery into everyday clinical applications. However, with the increasingly stringent requirements to regulate medical physics as a health profession, the role of physicists as scientists and innovators has become at serious risk of erosion. These challenges trigger the need for a new, revolutionized training program at the graduate level that respects scientific rigor, attention for medical physics-relevant developments in basic sciences, innovation and entrepreneurship. Methods: A grant proposal was funded by the Collaborative REsearch and Training Experience program (CREATE) of the Natural Sciences and Engineering Research Council (NSERC) of Canada. This enabled the creation of the Medical Physics Research Training Network (MPRTN) around two CAMPEP-accredited medical physics programs. Members of the network consist of medical device companies, government (research and regulatory) and academia. The MPRTN/CREATE program proposes a curriculum with three main themes: (1) radiation physics, (2) imaging & image processing and (3) radiation response, outcomes and modeling. Results: The MPRTN was created mid 2013 (mprtn.com) and features (1) four new basic Ph.D. courses; (2) industry participation in research projects; (3) formal job-readiness training with involvement of guest faculty from academia, government and industry. MPRTN activities since 2013 include 22 conferences; 7 workshops and 4 exchange travels. Three patents were filed or issued, nine awards/best papers were won. Fifteen journal publications were accepted/published, 102 conference abstracts. There are now 13 industry partners. Conclusion: A medical physics research training network has been set up with the goal to harness graduate student’s job-readiness for industry, government and academia in addition to the conventional clinical role. Two years after inception, significant successes have been booked, but the true challenge will be to demonstrate that with this training philosophy CREATE scholars gain access to a much broader job market. Supported by the Natural Sciences and Engineering Research Council (NSERC) Canada.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program

Criticality safety at TA-55 relies on nuclear material containers that are water resistant to prevent significant amounts of water from coming into contact with fissile material in the event of a fire that causes a breach of glovevbox confinement and subsequent fire water ingress. A “water tight container” is a container that will not allow more than 50ml of water ingress when fully submerged, except when under sufficient pressure to produce structural discontinuity. There are many types of containers, welded containers, hermetically sealed containers, filtered containers, etc.

61 RADIATION PROTECTION AND DOSIMETRY↗
NASA NTRSDate not supplied

Operational Health Physics for Radiation Source Payloads

Spacecraft payloads processed at the Kennedy Space Center (KSC) frequently include ionizing and nonionizing radiation sources that require controls to protect the health of the workers involved in the ground operations. At KSC, a robust radiation protection program ensures that radiation hazards are identified, and appropriate controls are implemented using a radiation use authorization (RUA) process and Radiation Protection Committee review and approval. The Health Physics Office provides monitoring and oversight to ensure compliance. A brief overview of health physics activities in support of the Mars 2020 mission will be described to illustrate the overall management of radiation sources during payload ground processing.

Kurt R Geber↗
DOE OSTI2012

Pacific Northwest National Laboratory Site Radionuclide Air Emissions Report for Calendar Year 2011

This report documents radionuclide air emissions that result in the highest effective dose equivalent (EDE) to a member of the public, referred to as the maximally exposed individual (MEI). The report has been prepared in compliance with the Code of Federal Regulations (CFR), Title 40, Protection of the Environment, Part 61, National Emission Standards for Hazardous Air Pollutants (NESHAP), Subpart H, National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities and Washington Administrative Code (WAC) Chapter 246-247, Radiation ProtectionAir Emissions. The EDE to the PNNL Site MEI due to routine emissions in 2011 from PNNL Site sources was 1.7E 05 mrem (1.7E-7 mSv) EDE. No nonroutine emissions occurred in 2011. The total radiological dose for 2011 to the MEI from all PNNL Site radionuclide emissions was more than 10,000 times smaller than the federal and state standard of 10 mrem/yr, to which the PNNL Site is in compliance.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Education and training in the field of nuclear instrumentation and measurement: CEA/INSTN (National Institute for Nuclear Sciences and Technologies) strategy to improve and develop new pedagogical tools and methods

Part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) is a higher education institution whose mission is to provide students and professionals a high level of scientific and technological qualification in all disciplines related to nuclear energy applications. In this frame, INSTN carries out education and training (E and T) programs in nuclear instrumentation and radioprotection. Its strategy has always been to complete theoretical courses by training courses and laboratory works carried out on an extensive range of training tools that includes a large panel of nuclear instrumentation as well as software applications. Since its creation in 1956, the INSTN has conducted both education and vocational programs on ionizing radiation detection. An extensive range of techniques have commonly been used during practical works with students and employees of companies who need to get the knowledge and specialization in this field. Today, the INSTN is mainly equipped with usual detectors and electronics in large numbers in order to be able to accommodate up to 48 trainees at the same time in two classrooms, with only two trainees for one workstation in order to optimize their learning. In the field of the neutron detection systems, the INSTN has strongly developed its offer taking advantage of the use of research reactors, such as ISIS reactor (700 kW) at Saclay. The implementation of neutron detection systems specific to the courses offers a unique way of observing and analysing the signal coming from neutron detectors, as well as learning how to set the parameters of the detection system in real conditions. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. Another field of activity of the INSTN is the development of new teaching tools using software applications. We describe hereafter two applications that have recently been developed by the institute, O.S.I.R.I.S and DOSIMEX. The INSTN and the company OREKA have developed an innovative teaching tool named O.S.I.R.I.S. The tool is built on a virtual 3D environment in which users operate in a totally free way. The action is located in a pressurised water reactor in the environment of a steam generator building. The users are students or professionals who want to learn and train on concrete situations on how to protect the workers against radiation. Through the use of this serious game, users must: establish a predictive dose evaluation, implement the principles of radiation protection, perform the dose result of the operation, analyse the gap between the predictive collective dose and the collective dose achieved, and decide about the lessons to provide feedback. Taking into account the increasing constraints related to radiation protection, it is essential to use computer codes to evaluate the dose rates generated by sources of ionizing radiation. Many powerful codes exist but are often relatively complex to implement and are dedicated to experts. In addition, these codes are often 'black boxes' that does not allow the user to understand the underlying physics. The objective of DOSIMEX codes developed by the INSTN is to give, by the means of a VBA code with a simple interface, the ability to easily conduct a wide range of radiological exposure situations: calculation of gamma, beta, neutron equivalent dose rate, calculation of internal contamination and atmospheric transfer.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Workload and transmission data for the installation of a digital breast tomosynthesis system

Purpose: Digital breast tomosynthesis (DBT) differs from conventional mammography in target/filter, kVp range, and imaging geometry. The aim of this study was to assess the breast input exposure of a DBT system by completing a workload survey of DBT installations, and to determine the parameters {alpha}, {beta}, and {gamma} in the Archer equation for the primary radiation generated by the clinical workload distributions. Methods: The authors conducted a retrospective survey of the x-ray breast imaging performed between September 2011 and September 2012 in three clinical DBT rooms equipped with Selenia Dimensions systems (Hologic Inc., Bedford, MA). A total of 343 examinations were analyzed to calculate the workload (mA-minute) and the primary air kerma at 1 m from the source (K{sup 1}). Transmission curves were calculated for the primary radiation generated by the workload distributions of the DBT rooms, and were fitted to the Archer equation. Results: There were large variations in patient volume and workload in the three examination rooms. In all these rooms, the average tube voltage (kVp) was about 31, the average K{sup 1} per patient was 16-21 mGy, and the average mA-minute per patient was 1.4-2.2 times higher than that of the mammography room described in NCRP Report No. 147. Most DBT screening examinations consisted of four two-dimensional mammographic views plus four tomosynthesis scans; the numbers of views acquired in diagnostic examinations varied widely. Tomosynthesis scans contributed about 30% of total mA-minute and about 50% of K{sup 1}. For the primary radiation generated by the clinical workload distributions, {alpha} was similar to that of 40-45 kVp W/Al (target/filter), and {alpha}+{beta} was similar to that of 30 kVp W/Al. Conclusions: The workload (mA-minute and K{sup 1}) distributions of mammographic examinations with DBT differ from conventional mammography. A field survey of patient volume and x-ray tube usage is important for the shielding design of DBT facilities. The clinical workloads and transmission data presented in this paper may be useful for clinical medical physicists.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

SU-E-T-599: Patient Safety Enhancements Through a Study of R&V System Override Data

Purpose: Record and verify (R&V) software systems include safety checks that compare actual machine parameters with prescribed values for a patient’s treatment, such as treatment couch position, linac, energy, and MUs. The therapist is warned of a mismatch with a pop-up and prompted to approve an override in order to continue without changes. Override approval is often legitimate, but the pop-up can also genuinely indicate a problem that would Result in the wrong treatment. When there are numerous pop-up warnings, human nature leads us to approve any override without careful reading, undermining the effectiveness of the safety mechanism. Methods: Override data was collected from our R&V system for all patients treated between October 8 and 29, 2012, on four linacs and entered into a spreadsheet. Additional data collected included treatment technique, disease site, immobilization, time, linac, and whether localization images were obtained. Data were analyzed using spreadsheet tools to reveal trends, patterns and associations that might suggest appropriate process changes that could decrease the total number of overrides. Results: 76 out of 113 patients had overrides. Out of the 944 treatments, 599 override items were generated. The majority were due to couch positions. 74 of the 84 overrides on a linac equipped with a 6D couch were due to the use of the rotational corrections and the fact that the 6D couch control does not communicate with the R&V system; translations required to rotate the couch appear to the R&V system as translations outside the tolerance range. Conclusion: Many findings were interesting but did not suggest a process change. Proposed process changes include creating site-specific instead of just technique-specific tolerance tables for couch shifts. Proposed improvements to the vendor are to facilitate direct communication between the 6D couch and the R&V system to eliminate those override warnings related to lack of communication.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Sci-Fri AM: Quality, Safety, and Professional Issues 01: CPQR Technical Quality Control Suite Development including Quality Control Workload Results

A close partnership between the Canadian Partnership for Quality Radiotherapy (CPQR) and the Canadian Organization of Medical Physicist’s (COMP) Quality Assurance and Radiation Safety Advisory Committee (QARSAC) has resulted in the development of a suite of Technical Quality Control (TQC) Guidelines for radiation treatment equipment, that outline specific performance objectives and criteria that equipment should meet in order to assure an acceptable level of radiation treatment quality. The framework includes consolidation of existing guidelines and/or literature by expert reviewers, structured stages of public review, external field-testing and ratification by COMP. The adopted framework for the development and maintenance of the TQCs ensures the guidelines incorporate input from the medical physics community during development, measures the workload required to perform the QC tests outlined in each TQC, and remain relevant (i.e. “living documents”) through subsequent planned reviews and updates. This presentation will show the Multi-Leaf Linear Accelerator document as an example of how feedback and cross-national work to achieve a robust guidance document. During field-testing, each technology was tested at multiple centres in a variety of clinic environments. As part of the defined feedback, workload data was captured. This lead to average time associated with testing as defined in each TQC document. As a result, for a medium-sized centre comprising 6 linear accelerators and a comprehensive brachytherapy program, we evaluate the physics workload to 1.5 full-time equivalent physicist per year to complete all QC tests listed in this suite.

60 APPLIED LIFE SCIENCES↗
DOE OSTI1995

Nuclear Safety [Vol. 36, No. 1, January-June 1995]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. THE CHORNOBYL ACCIDENT: 1 The Chornobyl Accident Revisited, Part II: The State of the Nuclear Fuel Located Within the Chornobyl Sarcophagus, A A. Borovoi and A. R. Sich; GENERAL SAFETY CONSIDERATIONS: 33 Nuclear Power Safety in Central and Eastern Europe, R. Wilson; 46 Safety of Nuclear Power Reactors in the Former Eastern European Countries, S. Chakraborty; 53 Technical Note: On the Definition of Common-Cause Failures, H. Paula; ACCIDENT ANALYSIS: 58 Modeling and Analysis of Core-Debris Recriticality During Hypothetical Severe Accidents in the Advanced Neutron Source Reactor, S.-H. Kim, V. Georgevich, D. B. Simpson, C. O. Slater, and R. P. Taleyarkhan; 68 Ignitability of Hydrogen/Oxygen/Diluent Mixtures in the Presence of Hot Surfaces, R. K. Kumar and G. W. Koroll; 94 Coupled RELAP5 and CONTAIN Accident Analysis Using PVM, K. A. Smith, A. J. Baratta, and G. E. Robinson; CONTROL AND INSTRUMENTATION: 109 Application of Fuzzy Logic in Nuclear Reactor Control Part I: An Assessment of State-of-the-Art, A. S. Heger, N. K. Alang-Rashid, and M. Jamshidi; DESIGN FEATURES: 122 Twenty-Third DOE/NRC Nuclear Air-Cleaning and Treatment Conference, R. R. Bellamy, J. J. Hayes, and M. W. First; ENVIRONMENTAL EFFECTS: 135 Atmospheric Dispersion and the Radiological Consequences of Normal Airborne Effluents from a Nuclear Power Plant, D. Fang, C. Z. Sun, and L. Yang; 142 Calculation of Distribution Coefficients for Radionuclides in Soils and Sediments, I. Puigdomenech and U. Bergstrom: OPERATING EXPERIENCES: 155 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 158 Operating Experience Feedback Report—Reliability of Safety-Related Steam Turbine-Driven Standby Pumps Used in U.S. Commercial Nuclear Power Plants, J. R. Boardman; 166 Turbine Building Hazards, H. L Ornstein; RECENT DEVELOPMENTS: 169 Reports, Standards, and Safety Guides, D. S. Queener; 175 Proposed Rule Changes as of Dec. 31,1994; ANNOUNCEMENTS: 32 Harvard School of Public Health In-Place Filter Testing Workshop; 134 International Conference on Advances in the Operational Safety of Nuclear Power Plants; 193 30th Tennessee Industries Week; 193 DOE Technical Standards Program 1995 Workshop; 194 Multiphase Flow Experiments and Instrumentation; 180 The Authors; 185 Indexes to Nuclear Safety, Volumes 34 and 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Application of Ultrasonic for Decontamination of Contaminated Soil - 13142

The trials of soil decontamination were carried out with the help of a pilot ultrasonic installation in different modes. The installation included a decontamination bath equipped with ultrasonic sources, a precipitator for solution purification from small particles (less than 80 micrometer), sorption filter for solution purification from radionuclides washing out from soil, a tank for decontamination solution, a pump for decontamination solution supply. The trials were carried out on artificially contaminated sand with specific activity of 4.5 10{sup 5} Bk/kg and really contaminated soil from Russian Scientific Center 'Kurchatovsky Institute' (RSC'KI') with specific activity of 2.9 10{sup 4} Bk/kg. It was established that application of ultrasonic intensify the process of soil reagent decontamination and increase its efficiency. The decontamination factor for the artificially contaminated soil was ∼200 and for soil from RSC'KI' ∼30. The flow-sheet diagram has been developed for the new installation as well as determined the main technological characteristics of the equipment. (authors)

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

Nuclear analysis generic requirements for fusion applications of radiation shielding - 25639

Radiation shielding engineering and associated nuclear analysis plays an important role in design, operation, maintenance and decommissioning of nuclear fusion devices. Requirements adopted in current analyses are investigated versus basic principles of requirement management and verification and validation aspects. An assessment is proposed to derive applicable product and process requirements from safety principles and transversal function requirements. V and V aspects are extended to cover input and output data. It is concluded that beyond the progress on computational capabilities and technical quality of fusion neutronics shielding analysis, there is the need to develop the underlying procedures and requirement management to satisfy configuration management, quality assurance and safety demonstration needs. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

A modern Monte Carlo investigation of the TG-43 dosimetry parameters for an {sup 125}I seed already having AAPM consensus data

Purpose: To investigate potential causes for differences in TG-43 brachytherapy dosimetry parameters in the existent literature for the model IAI-125A{sup 125}I seed and to propose new standard dosimetry parameters. Methods: The MCNP5 code was used for Monte Carlo (MC) simulations. Sensitivity of dose distributions, and subsequently TG-43 dosimetry parameters, was explored to reproduce historical methods upon which American Association of Physicists in Medicine (AAPM) consensus data are based. Twelve simulation conditions varying{sup 125}I coating thickness, coating mass density, photon interaction cross-section library, and photon emission spectrum were examined. Results: Varying{sup 125}I coating thickness, coating mass density, photon cross-section library, and photon emission spectrum for the model IAI-125A seed changed the dose-rate constant by up to 0.9%, about 1%, about 3%, and 3%, respectively, in comparison to the proposed standard value of 0.922 cGy h{sup −1} U{sup −1}. The dose-rate constant values by Solberg et al. [“Dosimetric parameters of three new solid core {sup 125}I brachytherapy sources,” J. Appl. Clin. Med. Phys. 3, 119–134 (2002)], Meigooni et al. [“Experimental and theoretical determination of dosimetric characteristics of IsoAid ADVANTAGE™ {sup 125}I brachytherapy source,” Med. Phys. 29, 2152–2158 (2002)], and Taylor and Rogers [“An EGSnrc Monte Carlo-calculated database of TG-43 parameters,” Med. Phys. 35, 4228–4241 (2008)] for the model IAI-125A seed and Kennedy et al. [“Experimental and Monte Carlo determination of the TG-43 dosimetric parameters for the model 9011 THINSeed™ brachytherapy source,” Med. Phys. 37, 1681–1688 (2010)] for the model 6711 seed were +4.3% (0.962 cGy h{sup −1} U{sup −1}), +6.2% (0.98 cGy h{sup −1} U{sup −1}), +0.3% (0.925 cGy h{sup −1} U{sup −1}), and −0.2% (0.921 cGy h{sup −1} U{sup −1}), respectively, in comparison to the proposed standard value. Differences in the radial dose functions between the current study and both Solberg et al. and Meigooni et al. were <10% for r ≤ 5 cm, and increased for r > 5 cm with a maximum difference of 29% at r = 9 cm. In comparison to Taylor and Rogers, these differences were lower (maximum of 2% at r = 9 cm). For the similarly designed model 6711 {sup 125}I seed, differences did not exceed 0.5% for 0.5 ≤ r ≤ 10 cm. Radial dose function values varied by 1% as coating thickness and coating density were changed. Varying the cross-section library and source spectrum altered the radial dose function by 25% and 12%, respectively, but these differences occurred at r = 10 cm where the dose rates were very low. The 2D anisotropy function results were most similar to those of Solberg et al. and most different to those of Meigooni et al. The observed order of simulation condition variables from most to least important for influencing the 2D anisotropy function was spectrum, coating thickness, coating density, and cross-section library. Conclusions: Several MC radiation transport codes are available for calculation of the TG-43 dosimetry parameters for brachytherapy seeds. The physics models in these codes and their related cross-section libraries have been updated and improved since publication of the 2007 AAPM TG-43U1S1 report. Results using modern data indicated statistically significant differences in these dosimetry parameters in comparison to data recommended in the TG-43U1S1 report. Therefore, it seems that professional societies such as the AAPM should consider reevaluating the consensus data for this and others seeds and establishing a process of regular evaluations in which consensus data are based upon methods that remain state-of-the-art.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

TU-H-CAMPUS-TeP2-05: Selective Protection of Normal Tissue by Cerium Oxide Nanoparticles During Radiation Therapy

Purpose: Cerium oxide nanoparticles (CONPs) have unique pH dependent properties such that they act as a radical modulator. These properties may be used in radiation therapy (RT) to protect normal tissue. This work investigates the selective radioprotection of CONPs in-vitro and potential for in-situ delivery of CONPs in prostate cancer RT. Methods: i) Normal human umbilical vein endothelial cells (HUVEC) and human prostate cancer cells (PC-3) were treated with 0 or 2 ng/mL CONPs (NP size: 5 nm). 2 Gy of 100 kVp radiation was delivered to the cells 4 hours after the CONP treatment. Cell viability was checked 48 hours later using MTS assays. ii) A prostate tumor was modeled as a 2-cm diameter sphere. CONPs were proposed to be loaded in a hollow radiotherapy fiducial marker. The concentration profile for the CONPs within the tumor was modeled with a previously validated diffusion equation employed in other studies for nanoparticles 10 nm or less. Results: i) Without radiation, cell viability was above 90% when treated with 2 ng/mL CONPs for both HUVEC and PC-3. After irradiation, a slightly higher viability was observed in HUVEC with CONPs than the ones without CONPs, and this effect was not observed in PC-3. ii) Based on the calculations, 2 ng/mL of CONPs could be delivered to normal cells by diffusion with a 1 µg/mL initial concentration within two weeks. Conclusion: We conclude that CONPs can provide selective radioprotection. The delivery of needed concentrations of CONPs is feasible via in-situ release from radiotherapy biomaterials (e.g. fiducials) loaded with the CONPs.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2013

Validation of nuclear models used in space radiation shielding applications

A program of verification and validation has been undertaken to assess the applicability of models to space radiation shielding applications and to track progress as these models are developed over time. In this work, simple validation metrics applicable to testing both model accuracy and consistency with experimental data are developed. The developed metrics treat experimental measurement uncertainty as an interval and are therefore applicable to cases in which epistemic uncertainty dominates the experimental data. To demonstrate the applicability of the metrics, nuclear physics models used by NASA for space radiation shielding applications are compared to an experimental database consisting of over 3600 experimental cross sections. A cumulative uncertainty metric is applied to the question of overall model accuracy, while a metric based on the median uncertainty is used to analyze the models from the perspective of model development by examining subsets of the model parameter space.

61 RADIATION PROTECTION AND DOSIMETRY↗