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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Implementation of Information Management System for Radiation Safety of Personnel at the Russian Northwest Center for Radioactive Waste Management 'SevRAO' - 13131

The report is an overview of the information-analytical system designed to assure radiation safety of workers. The system was implemented in the Northwest Radioactive Waste Management Center 'SevRAO' (which is a branch of the Federal State Unitary Enterprise 'Radioactive Waste Management Enterprise RosRAO'). The center is located in the Northwest Russia. In respect to 'SevRAO', the Federal Medical-Biological Agency is the regulatory body, which deals with issues of radiation control. The main document to regulate radiation control is 'Reference levels of radiation factors in radioactive wastes management center'. This document contains about 250 parameters. We have developed a software tool to simplify control of these parameters. The software includes: input interface, the database, dose calculating module and analytical block. Input interface is used to enter radiation environment data. Dose calculating module calculates the dose on the route. Analytical block optimizes and analyzes radiation situation maps. Much attention is paid to the GUI and graphical representation of results. The operator can enter the route at the industrial site or watch the fluctuations of the dose rate field on the map. Most of the results are presented in a visual form. Here we present some analytical tasks, such as comparison of the dose rate in some point with control levels at this point, to be solved for the purpose of radiation safety control. The program helps to identify points making the largest contribution to the collective dose of the personnel. The tool can automatically calculate the route with the lowest dose, compare and choose the best route. The program uses several options to visualize the radiation environment at the industrial site. This system will be useful for radiation monitoring services during the operation, planning of works and development of scenarios. The paper presents some applications of this system on real data over three years - from March 2009 to February 2012. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2019

Cs-137 Body-burdens in Wild Pigs on the Department of Energy's Savannah River Site: Determination of an Appropriate Framework for Spatiotemporal Analysis of Hunter Dose Tracking System Data

The Savannah River Site (SRS) is a U.S. Department of Energy facility with an area of 310 square miles. Since its opening in 1951, SRS has produced and processed nuclear materials for national defense and industrial purposes. These activities have resulted in the release of radiocesium (Cs-137) which has led to contamination of associated streams, floodplains, man-made reservoirs, soil and vegetation. Contamination is spread to wild pigs as they move about SRS, consume vegetation, drink water and wallow in soil. SRS provides annual deer hunts open to the public in which wild pigs may be taken as well. The Cs-137 concentration of each animal is determined before it is released to the hunter for consumption and a cumulative dose is retained for each individual hunter. These tasks are accomplished using the Hunter Dose Tracking System (HDTS). The data collected on SRS by the HDTS for White-Tailed deer has been extensively analyzed. A repeatable method for calculating the background radiocesium body burden for the SRS deer population was established by Gaines and Novak (2009,2016). This model was framed using Cs-137 contamination obtained via gamma over-flight data, the decay rate of radiocesium and the historical behavior observed in the SRS deer population. These parameters were used in a spatiotemporal analysis to estimate background Cs-137 body burdens (from global fallout, etc.), as well as the effect SRS activities may have. The purpose of this project is to modify the framework to represent the behavior of the wild pig population on SRS such that HDTS data collected may be analyzed similarly. The 2018 data were collected via the HDTS in the field to screen wild pigs dispatched by Forest Service - Savannah River. These data will provide insight for an analysis of seasonal variance in the body burdens of the SRS wild pig population. The HDTS quickly provided Cs-137 concentrations in the field, as well as automatically updated the electronic database. Space: Historically, the average home range of wild pigs on the Savannah River Site has been reported at approximately six square miles. This range was confirmed accurate as of July 2018, with the observed minimum and maximum at one and ten square miles respectively. This range was geoprocessed using ArcMap to project buffers based on 1991 gamma over-flight data. The buffers were constructed to originate at all Cs-137 contamination point coordinates and protrude outward in all directions for 1.5, 2.45 and 6 square miles. While the 2.45 mile buffer is theoretically representative of potential hog exposure based on the average home range, it does not consider the possibility of home ranges not uniform in shape. Therefore, the 1.5 and 6 mile buffers were also constructed as visual aids for determining the appropriate hunt compartments for background level considerations. The SRS is divided into numbered hunt compartments of varying size and shape. These compartment numbers are used during sponsored hunts to record the location in which a wild pig was dispatched. Time: The Gaines report determined deer analysis should be carried out at the half-decade level for the best representation of true body burdens. This is due to the impact natural factors, such as drought or flood, can have on the bioaccumulation of Cs-137. To determine the appropriate time scale for hog analysis, all available data was totaled and averaged at 1, 2, 3 and 7 year increments. The annual averages and running average are provided. The appropriate compartments for the determination of the SRS hog population background level should be: 2, 3, 4, 9, 10, 14 and 15. Hogs dispatched within other compartments are most likely exposed to Cs-137 contamination and should not be included in the background level. The current dataset is small, but was still able to demonstrate the relationship between where hogs on SRS and the Cs-137 body burdens they incur. Geoprocessing the data using the average home range of wild pigs highlighted the importance of collecting additional samples from areas surrounding SRS if an accurate background is to be determined. When the average concentration of Cs-137 in pigs from background compartments is compared against that of pigs from contaminated compartments, a significant difference can be seen. A Table provides a summary of the analyzed data from the 15 years processed thus far.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2007

Anti-radiation vaccine: Immunologically-based Prophylaxis of Acute Toxic Radiation Syndromes Associated with Long-term Space Flight

Protecting crew from ionizing radiation is a key life sciences problem for long-duration space missions. The three major sources/types of radiation are found in space: galactic cosmic rays, trapped Van Allen belt radiation, and solar particle events. All present varying degrees of hazard to crews; however, exposure to high doses of any of these types of radiation ultimately induce both acute and long-term biological effects. High doses of space radiation can lead to the development of toxicity associated with the acute radiation syndrome (ARS) which could have significant mission impact, and even render the crew incapable of performing flight duties. The creation of efficient radiation protection technologies is considered an important target in space radiobiology, immunology, biochemistry and pharmacology. Two major mechanisms of cellular, organelle, and molecular destruction as a result of radiation exposure have been identified: 1) damage induced directly by incident radiation on the macromolecules they encounter and 2) radiolysis of water and generation of secondary free radicals and reactive oxygen species (ROS), which induce chemical bond breakage, molecular substitutions, and damage to biological molecules and membranes. Free-radical scavengers and antioxidants, which neutralize the damaging activities of ROS, are effective in reducing the impact of small to moderate doses of radiation. In the case of high doses of radiation, antioxidants alone may be inadequate as a radioprotective therapy. However, it remains a valuable component of a more holistic strategy of prophylaxis and therapy. High doses of radiation directly damage biological molecules and modify chemical bond, resulting in the main pathological processes that drive the development of acute radiation syndromes (ARS). Which of two types of radiation-induced cellular lethality that ultimately develops, apoptosis or necrosis, depends on the spectrum of incident radiation, dose, dose rate, and functional conditions of impacted cells/organisms. The administration of an experimental anti-radiation vaccine may provide an immunologically based, adjunct method of prevention or prophylaxis against clinical ARS. The administration of experimental anti-radiation serum (ARS) and the use of the blood dialysis methods, such as immune plasma-sorption, may assist in the clearance of radiation-specific toxins and may enhance established strategies for the mitigation of the biological effects leading to ARS, and should be evaluated for use on exploration-class space missions.

Popov, Dmitri↗
DOE OSTI2014

Probabilistic risk analysis toward cost-effective 3S (safety, safeguards, security) implementation

Probabilistic Risk Analysis (PRA) has been introduced for several decades in safety and nuclear advanced countries have already used this methodology in their own regulatory systems. However, PRA has not been developed in safeguards and security so far because of inherent difficulties in intentional and malicious acts. In this paper, probabilistic proliferation and risk analysis based on random process is applied to hypothetical reprocessing process and physical protection system in nuclear reactor with the Markov model that was originally developed by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) in Generation IV International Framework (GIF). Through the challenge to quantify the security risk with a frequency in this model, integrated risk notion among 3S to pursue the cost-effective installation of those countermeasures is discussed in a heroic manner.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Early-Stage Radiation Safety Analysis for the Spallation Neutron Source Second Target Station Bunker Operations

The Second Target Station project at Oak Ridge National Laboratory will develop a cold neutron source to meet growing experimental needs. This paper describes calculations of the residual dose rates associated with the monolith shield plug and the beamline bunker, two key conventional operations and radiation safety features. While neutron production is active, the instrument hall outside the bunker must be generally accessible with dose rates of less than 0.25 mrem/h. When neutron production is halted, the bunker must be accessible for hands-on maintenance operations. Further, these two requirements form the cause for the assessments reported herein of residual dose rates caused by the monolith shield plug and residual dose rates in the bunker. The monolith shield plug was shown to not produce significant dose rates inside the bunker after a 20-year lifetime, and the residual dose rates inside the bunker for the case of an operating beamline were shown to reasonably allow for hands-on maintenance. These calculations are based on preliminary design models of the relevant systems. Additionally, an example showing the significance of considering neutron supermirror physics in transport calculations that track nuclide production and destruction rates to produce gamma sources for residual dose rate calculations is included. The example shows that if neutron supermirror physics is not considered, dose rate fields may be significantly underpredicted.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Modeling Plutonium Decorporation in a Female Nuclear Worker Treated with Ca-DTPA after Inhalation Intake

The present work models plutonium (Pu) biokinetics in a female former nuclear worker. Her bioassay measurements are available at the US Transuranium and Uranium Registries. The worker was internally exposed to a plutonium-americium mixture via acute inhalation at a nuclear weapons facility. She was medically treated with injections of 1 g Ca-DTPA on days 0, 5, and 14 after the intake. Between days 0 and 20, fecal and urine samples were collected and analyzed for 239 Pu and 241 Am. Subsequently, she was followed up for bioassay monitoring over 14 y, with additional post-treatment urine samples collected and analyzed for 239 Pu. The uniqueness of this dataset is due to the availability of: (1) both early and long-term bioassay data from a female with plutonium intake; (2) data on chelation therapy for a female; and (3) fecal measurement results. Chelation therapy with Ca- and/or Zn-salts of DTPA is known to aid in reducing the internal radiation dose by enhancing the excretion of plutonium and americium from the body. Such enhancement affects plutonium biokinetics in the human body, posing a challenge to the internal dose assessment. The current radiation dose assessment practice is to exclude the data affected by Ca-DTPA from the analysis. The present analysis is the first to explicitly model the chelation-affected bioassay data in a female by using a newly developed chelation model. Thus, the bioassay data collected during and after the Ca-DTPA administrations were used for biokinetic modeling and dose assessment. The Markov Chain Monte Carlo method was used to investigate model parameter uncertainty, based on the bioassay data and assumed prior probability distributions. A χ 2 /nData (number of data points) ≈ 1 was observed in this study, which indicates self-consistency of the data with the model. Results of this study show that the worker’s 239 Pu intake was 12 Bq, with a committed effective dose to the whole-body of 1.2 mSv and a committed equivalent dose to the bone surfaces, liver, and lungs of 37.8, 9.1, and 0.8 mSv, respectively. This study also discusses the worker’s dose reduction due to chelation treatment.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Development and optimization of a beam shaper device for a mobile dedicated IOERT accelerator

Purpose: The aim of this study was to design and build a prototype beam shaper to be used on a dedicated mobile accelerator that protects organs at risk within the radiation field and conforms the beam to the target geometry during intraoperative electron radiotherapy (IOERT). A dosimetric characterization of the beam shaper device was performed based on Monte Carlo (MC) simulations, as well as experimental data, at different energies, field sizes, and source to skin distances. Methods: A mobile light intraoperative accelerator (LIAC{sup Registered-Sign }, Sordina, Italy) was used. The design of the beam shaper prototype was based on MC simulations (BEAMnrc/OMEGA and DOSXYZnrc code) for a selection of materials and thicknesses, as well as for dosimetric characterization. Percentage depth dose (PDD) and profile measurements were performed using a p-type silicon diode and a commercial water phantom, while output factors were measured using a PinPoint ion chamber in a PMMA phantom. Planar doses in planes of interest were carried out using radiochromic films (Gafchromic{sup TM} EBT and EBT2) in PMMA and in a Solid Water{sup Registered-Sign} phantom. Several experimental set-ups were investigated with the beam shaper device fixed on the top of the phantom, varying both the short side of the rectangular field and the air gap between the device and the phantom surface, simulating the clinical situation. The output factors (OFs) were determined using different geometrical set-ups and energies. Results: The beam shaper prototype consists of four blades sliding alongside each other and mounted on a special support at the end of the 10 cm diameter PMMA circular applicator. Each blade is made of an upper layer of 2.6 cm of Teflon{sup Registered-Sign} and a lower layer of 8 mm of stainless steel. All rectangles inscribed in a 5 cm diameter can be achieved in addition to any 'squircle-shaped' field. When one side of the rectangular field is held constant and the second side is reduced, both R{sub 50} and R{sub max} move towards the phantom surface. Comparing the PDDs obtained with the 5 cm circular applicator and with a 4.4 Multiplication-Sign 4.4 cm{sup 2} square field (that is the equivalent square of the 5 cm circular field) obtained with the beam shaper, a different behavior was observed in the region extending from the surface to a depth of 50% of the maximum dose. Isodoses measured for rectangular fields used for clinical cases (i.e., 4 Multiplication-Sign 9 cm{sup 2} 8 MeV) are shown, with different air gaps. For each energy investigated, the normalized OFs slowly increase, when the length of the side decreases down to about 4 cm, and then rapidly decreases for smaller field widths. MC simulation showed an excellent agreement with experimental data (<2%). Conclusions: The beam shaper device is able to provide square/rectangular/squircle fields with adequate dose homogeneity for mobile dedicated accelerators, thus allowing conformal treatment with IOERT. Monte Carlo simulation can be a very useful tool to simulate any clinical set up and can be used to create a data set to calculate MUs, thereby increasing the accuracy of the delivered dose during IOERT procedures.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2014

Comparison of 2D and 3D gamma analyses

Purpose: As clinics begin to use 3D metrics for intensity-modulated radiation therapy (IMRT) quality assurance, it must be noted that these metrics will often produce results different from those produced by their 2D counterparts. 3D and 2D gamma analyses would be expected to produce different values, in part because of the different search space available. In the present investigation, the authors compared the results of 2D and 3D gamma analysis (where both datasets were generated in the same manner) for clinical treatment plans. Methods: Fifty IMRT plans were selected from the authors’ clinical database, and recalculated using Monte Carlo. Treatment planning system-calculated (“evaluated dose distributions”) and Monte Carlo-recalculated (“reference dose distributions”) dose distributions were compared using 2D and 3D gamma analysis. This analysis was performed using a variety of dose-difference (5%, 3%, 2%, and 1%) and distance-to-agreement (5, 3, 2, and 1 mm) acceptance criteria, low-dose thresholds (5%, 10%, and 15% of the prescription dose), and data grid sizes (1.0, 1.5, and 3.0 mm). Each comparison was evaluated to determine the average 2D and 3D gamma, lower 95th percentile gamma value, and percentage of pixels passing gamma. Results: The average gamma, lower 95th percentile gamma value, and percentage of passing pixels for each acceptance criterion demonstrated better agreement for 3D than for 2D analysis for every plan comparison. The average difference in the percentage of passing pixels between the 2D and 3D analyses with no low-dose threshold ranged from 0.9% to 2.1%. Similarly, using a low-dose threshold resulted in a difference between the mean 2D and 3D results, ranging from 0.8% to 1.5%. The authors observed no appreciable differences in gamma with changes in the data density (constant difference: 0.8% for 2D vs 3D). Conclusions: The authors found that 3D gamma analysis resulted in up to 2.9% more pixels passing than 2D analysis. It must be noted that clinical 2D versus 3D datasets may have additional differences—for example, if 2D measurements are made with a different dosimeter than 3D measurements. Factors such as inherent dosimeter differences may be an important additional consideration to the extra dimension of available data that was evaluated in this study.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

SU-E-I-10: Automatic Monitoring of Accumulated Dose Indices From DICOM RDSR to Improve Radiation Safety in X-Ray Angiography

Purpose: To investigate the potential benefits of automatic monitoring of accumulated patient and staff dose indicators, i.e., CAK and KAP, from DICOM Radiation Dose Structured Reports (RDSR) in x-ray angiography (XA). Methods: Recently RDSR has enabled the convenient aggregation of dose indices and technique parameters for XA procedures. The information contained in RDSR objects for three XA systems, dedicated to different types of clinical procedures, has been collected and aggregated in a database for over one year using a system developed with open-source software at the Karolinska University Hospital. Patient weight was complemented to the RDSR data via an interface with the Hospital Information System (HIS). Results: The linearly approximated trend in KAP over a time period of a year for cerebrovascular, pelvic/peripheral vascular, and cardiovascular procedures showed a decrease of 12%, 20%, and 14%, respectively. The decrease was mainly due to hardware/software upgrades and new low-dose imaging protocols, and partially due to ongoing systematic radiation safety education of the clinical staff. The CAK was in excess of 3 Gy for 15 procedures, and exceeded 5 Gy for 3 procedures. The dose indices have also shown a significant dependence on patient weight for cardiovascular and pelvic/peripheral vascular procedures; a 10 kg shift in mean patient weight can result in a dose index increase of 25%. Conclusion: Automatic monitoring of accumulated dose indices can be utilized to notify the clinical staff and medical physicists when the dose index has exceeded a predetermined action level. This allows for convenient and systematic follow-up of patients in risk of developing deterministic skin injuries. Furthermore, trend analyses of dose indices over time is a valuable resource for the identification of potential positive or negative effects (dose increase/decrease) from changes in hardware, software, and clinical work habits.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

SU-E-T-469: Implementation of VAs Web-Based Radiotherapy Incident Reporting and Analysis System (RIRAS)

Purpose: This Web-based Radiotherapy Incident Reporting and Analysis System (RIRAS) is a tool to improve quality of care for radiation therapy patients. This system is an important facet of continuing effort by our community to maintain and improve safety of radiotherapy.Material and Methods: VA’s National Radiation Oncology Program office has embarked on a program to electronically collect adverse events and good-catch data of radiation treatment of over 25,000 veterans treated with radiotherapy annually. This VA-Intranet based software design has made use of dataset taxonomies and data dictionaries defined in AAPM/ASTRO reports on error reporting. We used proven industrial and medical event reporting techniques to avoid several common problems faced in effective data collection such as incomplete data due to data entry fatigue by the reporters, missing data due to data difficult to obtain or not familiar to most reporters, missing reports due to fear of reprisal etc. This system encompasses the entire feedback loop of reporting an incident, analyzing it for salient details, and developing interventions to prevent it from happening again. The analysis reports with corrective, learning actions are shared with the reporter/facility and made public to the community (after deidentification) as part of the learning process. Results: Till date 50 incident/good catches have been reported in RIRAS and we have completed analysis on 100% of these reports. This is done due to the fact that each reported incidents is investigated and a complete analysis/patient-safety-work-product report is generated by radiation oncology domain-experts. Conclusions Because of the completeness of the data, the system has enabled us to analyze process steps and track trends of major errors which in the future will lead to implementing system wide process improvement steps and safe standard operating procedures for each radiotherapy treatment modality/technique and fulfills our goal of “Effecting Quality While Treating Safely”. RIRAS developed and copyrighted by TSG Innovations Inc.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

SU-G-IeP4-05: Experience with a Practical Approach to the Release of Radioactive Patients from Radiation Safety Isolation

Purpose: The radioactive patient releases of the nuclear medicine department in a large cancer center were analyzed to compare the estimated doses to others to the regulatory limit and institutional targets and to gauge how onerous the patients’ instructions were. Methods: In-house software that implements the NUREG 1556 approach and further considers fellow travelers, children and pregnant women and members of the public following Siegel, et al., is used to calculate release timing and to provide patients with individualized instructions. A retrospective analysis of the 218 releases in 2015 that required instructions included all I-131 and Lu-177 treatments. Results: There were 87 male and 131 female patients. They were 45.5±16.5 (8-84) years old, weighed 83.3±21.8 kg and stood 168±10 cm tall with a body mass index of 29.3±6.7 kg/m2. 195 patients received 2.96±2.23 GBq of I-131-NaI for differentiated thyroid carcinoma and eight received 636±231 MBq of I-131-NaI for hyperthyroidism. There were eight administrations of 18.5 GBq of I-131-mIBG and seven administrations of 7.4 GBq of Lu-177-DOTATATE. The dose to the most exposed person was 1.67±1.05 (mode=0.60, median=1.15) mSv. The dose to a fellow traveler was 0.31±0.31 (mode=0.1, median=0.17) mSv. The dose to children and pregnant women was 0.51±0.34 (mode=1.0, median=0.42) mSv. The duration of isolation was 7.7±17.3 (mode=2, median=1.7) hours. The duration of sleeping alone was 60.4±127 (mode=0, median=15.5) hours. The time to delay traveling was 12.2±11 (mode=median=0) hours. The time to completely avoid children and pregnant women was 23.3±26.5 (0–254, mode=median=24) hours followed by limited contact for 20.6±68.1 (0–491, mode=median=0) hours. The time to avoid others (e.g., to wait to return to work) was 12.1±101 (0–1465, mode=median=0) hours. Conclusion: The estimated doses to others were often well below the limits set by regulation, 5 mSv, or prudent practice, 1 mSv. Most patients’ instructions were not hard to comply with.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

WASTE ISOLATION PILOT PLANT (WIPP): THE NATIONS' SOLUTION TO NUCLEAR WASTE STORAGE AND DISPOSAL ISSUES

In the southeastern portion of my home state of New Mexico lies the Chihuahauan desert, where a transuranic (TRU), underground disposal site known as the Waste Isolation Pilot Plant (WIPP) occupies 16 square miles. Full operation status began in March 1999, the year I graduated from Los Alamos High School, in Los Alamos, NM, the birthplace of the atomic bomb and one of the nation’s main TRU waste generator sites. During the time of its development and until recently, I did not have a full grasp on the role Los Alamos was playing in regards to WIPP. WIPP is used to store and dispose of TRU waste that has been generated since the 1940s because of nuclear weapons research and testing operations that have occurred in Los Alamos, NM and at other sites throughout the United States (U.S.). TRU waste consists of items that are contaminated with artificial, man-made radioactive elements that have atomic numbers greater than uranium, or are trans-uranic, on the periodic table of elements and it has longevity characteristics that may be hazardous to human health and the environment. Therefore, WIPP has underground rooms that have been carved out of 2,000 square foot thick salt formations approximately 2,150 feet underground so that the TRU waste can be isolated and disposed of. WIPP has operated safely and successfully until this year, when two unrelated events occurred in February 2014. With these events, the safety precautions and measures that have been operating at WIPP for the last 15 years are being revised and improved to ensure that other such events do not occur again.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

International Radiological/Nuclear Training for Emergency Response - Major Public Events Virtual Workshop: Radiation Detection and Emergency Response Equipment (Day 2) [Slides]

The objective of this presentation is to familiarize participants with the different types of radiation detection systems and their practical applications for radiological emergency response. The specific goals are for participants to: (1) Understand the Three Step Process for Radiological Response of (i) Search and/or Survey, (ii) Radioisotope Identification, and (iii) Source Recovery, (2) Recognize the types of radiation detection equipment and their applications, and (3) View examples of common radiation detection instrumentation with operational videos.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Summary Report for the 2019 Radiation Detection for Nuclear Security Summer School

The Pacific Northwest National Laboratory (PNNL) hosted students from across the United States at the 6th Radiation Detection for Nuclear Security Summer School 17-28 June 2019. The summer school provided students with a unique understanding of nuclear security challenges faced in the field and exposed them to the technical foundations, analyses, and insight that will be required by future leaders in technology development and implementation. The course heavily emphasized laboratory and field demonstrations including measurements of special nuclear material. Student evaluations and feedback from student advisors indicates that the summer school achieved its objectives of 1) exposing students to the range of nuclear security applications for which radiation detection is necessary, 2) articulating the relevance of student research into the broader context, and 3) growing student interest in careers in nuclear security.

61 RADIATION PROTECTION AND DOSIMETRY↗