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IER 538: International Dosimetry Intercomparison Exercise with Godiva-IV [Slides]

Presentation on International Dosimetry Intercomparison Exercise with Godiva-IV. This presentation details Department of Energy (DOE) requirements and irradiation. Graphs show Neutron dose conversion factors, Neutron dose characterization, and Photon dose characterization. Results of 2m, 3m, and 4m tests are then compared to DOE standards. Findings show photon dose measurements were limited by instrumentation not available. Additionally, there has been a turnover of staff in dosimetry. However, funding is being provided for Universal Nuclear Accident Dosimeter (U-NAD).

61 RADIATION PROTECTION AND DOSIMETRY↗

AFRRI TRIGA Reactor Neutron and Gamma Dose Characterization Preliminary Results

Integral Experiment Request (IER) 484 is part of a series of dose characterization and nuclear accident dosimeter (NAD) exercises performed under the Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). The Armed Forces Radiobilogy Research Institute (AFRRI) 1.1 MW Training, Research, Isotopes, General Atomics (TRIGA) Mark-F reactor is the third facility to undergo a dose study to establish reference dose values. A preliminary measurement was performed earlier this year, where a variety of foils were irradiated and activation gamma rays were measured using a high-purity germanium (HPGe) detector. The data was taken and analyzed to calculate the neutron spectrum in the irradiation facility. This paper will present preliminary results of the neutron dose measured. The characterization of the AFRRI dose field is planned in August.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Near-Zero Power MOF-Based Sensors for NO 2 Detection

Detection and capture of toxic nitrogen oxides (NO x ) is important for emissions control of exhaust gases and general public health. The ability to directly electrically detect trace (0.5–5 ppm) NO 2 by a metal–organic framework (MOF)-74-based sensor at relatively low temperatures (50 °C) is demonstrated via changes in electrical properties of M-MOF-74, M = Co, Mg, Ni. The magnitude of the change is ordered Ni > Co > Mg and explained by each variant's NO 2 adsorption capacity and specific chemical interaction. Ni-MOF-74 provides the highest sensitivity to NO 2 ; a 725× decrease in resistance at 5 ppm NO2 and detection limit <0.5 ppm, levels relevant for industry and public health. Furthermore, the Ni-MOF-74-based sensor is selective to NO 2 over N 2 , SO 2 , and air. Linking this fundamental research with future technologies, the high impedance of MOF-74 enables applications requiring a near-zero power sensor or dosimeter, with the active material drawing <15 pW for a macroscale device 35 mm 2 with 0.8 mg MOF-74. Finally, this represents a 10 4 –10 6 × decrease in power consumption compared to other MOF sensors and demonstrates the potential for MOFs as active components for long-lived, near-zero power chemical sensors in smart industrial systems and the internet of things.

47 OTHER INSTRUMENTATION↗

X-ray Inspection Model Validation with Physical Dosimetry

A non-functional printed circuit board assembly was developed using ULTIboard and a simple computational transport model was created for use with the Monte Carlo software MCNP. EBT 3 film radiography was used to compare with Monte Carlo simulation to validate the board representation. The computational transport model was then revised to include connection pins, solder balls, highly attenuating internal structures, and copper trace distribution. Results from the revised model were compared to the EBT 3 films to observe improvements to dose profiles. It was found that this method was useful in verifying the placement of components, as the dose profiles were observed to follow the same trends. The experiment was then repeated using XRQA 2 films to achieve the same level of contrast with 1% the dose of EBT 3 films. It was found that high contrast may be achieved using these films to identify major issues with the model geometry, at a cost of dose profile accuracy. A second validation method was applied to the model using 37 CaF 2 thermoluminescent dosimeters (TLDs). TLD measurements were compared with the simplified and complex transport models to identify the features that have the greatest impact on simulation accuracy. The TLD calibration to CaF 2 was found to be accurate within 5.6%, while calibration to dose in Si was found to be accurate within 4.7%. Finally, it was observed that the accurate representation of solder balls and proper modeling of highly attenuating internal structures had the greatest impact on simulation accuracy.

61 RADIATION PROTECTION AND DOSIMETRY↗

Verification and Validation Tests of Gamma Library of MC2-3 for Coupled Neutron and Gamma Heating Calculation

For the accurate assessment of the heat generation rate in fast reactors, the gamma library of MC2 -3 and the MC2 -3 + GAMSOR procedure employing the coupled neutron and gamma heating calculation has been thoroughly verified against Monte Carlo results and validated using the ZPPR-15D gamma dose measurement data. NJOY outputs are post-processed in a consistent way with the NJOY procedure to avoid any missing data or double counting of data. Prompt heating for 379 out of 391 isotopes in the gamma library was verified against MCNP6.2 to within 1% relative error in total heating for most isotopes. For both a simple one-dimensional slab problem representing a sodium cooled fast reactor and the Experimental Breeder Reactor II (EBR-II) benchmark problem, the root-mean-square values of assembly power error were less than 0.5% for fuel assemblies, ~1% in blankets and ~1 to ~3% in reflectors compared to MCNP6.2 results. The most plausible cause for the 3% error in a reflector assembly is believed to be the error in the multigroup neutron cross sections for the reflector assembly. For validation, gamma doses measured with thermoluminiscent dosimeters (TLDs) in the ZPPR-15D experiment were calculated using GAMSOR. Due to the uncertainty in the TLD measurement with regards to the energy deposition of photons and neutrons, the validation data leads to a 12.7% uncertainty on the experimental measurement. With this uncertainty bound, the calculated doses all fell within one standard deviation of the measured value. Combined with the accurate calculation of reaction rate distributions and neutron spectrum measurements, these results indicate good agreement for the neutron and gamma heating calculations that were performed.

coupled neutron and gamma heating↗

Fundamentals of wildlife dosimetry and lessons learned from a decade of measuring external dose rates in the field

Methods for determining the radiation dose received by exposed biota require major improvements to reduce uncertainties and increase precision. We share our experiences in attempting to quantify external dose rates to free-ranging wildlife using GPS-coupled dosimetry methods. The manuscript is a primer on fundamental concepts in wildlife dosimetry in which the complexities of quantifying dose rates are highlighted, and lessons learned are presented based on research with wild boar and snakes at Fukushima, wolves at Chornobyl, and reindeer in Norway. GPS-coupled dosimeters produced empirical data to which numerical simulations of external dose using computer software were compared. Our data did not support a standing paradigm in risk analyses: Using averaged soil contaminant levels to model external dose rates conservatively overestimate the dose to individuals within a population. Following this paradigm will likely lead to misguided recommendations for risk management. The GPS-dosimetry data also demonstrated the critical importance of how modeled external dose rates are impacted by the scale at which contaminants are mapped. When contaminant mapping scales are coarse even detailed knowledge about each animal’s home range was inadequate to accurately predict external dose rates. Importantly, modeled external dose rates based on a single measurement at a trap site did not correlate to actual dose rates measured on free ranging animals. These findings provide empirical data to support published concerns about inadequate dosimetry in much of the published Chernobyl and Fukushima dose-effects research. Furthermore, our data indicate that a huge portion of that literature should be challenged, and that improper dosimetry remains a significant source of controversy in radiation dose-effect research.

61 RADIATION PROTECTION AND DOSIMETRY↗

Copper-nickel electroplating of 3D-printed acrylonitrile butadiene styrene for interference and radiation shielding applications

Here, in this study, electromagnetic and radiation shielding structures were fabricated by 3D printing acrylonitrile butadiene styrene (ABS) structures via fused filament fabrication and electroplating with copper (Cu) and nickel (Ni) metals. Prior to electroplating, the printed ABS materials were made conductive by surface electrodeposition of polypyrrole (PPy). ABS adsorbs the pyrrole monomer, which subsequently polymerizes via oxidation forming conductive PPy layer on the printed structures. The successful anchoring of PPy layer can be attributed to its hydrogen bonding and electrostatic interactions with ABS. Thermal analysis, optical, and electron microscopy equipped with energy dispersive X-ray were employed to characterize the printed shielding structures, while the interference and radiation shielding effectiveness were evaluated using a spectrum analyzer and dosimeter, respectively. The interference shielding was found to be effective at 60 ± 5 dB over relatively small frequency range. Shielding was also apparent against gamma and X-ray ionizing rays, but more significantly effective against beta rays. Furthermore, the Cu/Ni coating was able to reduce any heat-induced dimensional changes for the 3D printed ABS substrate without compromising the mechanical integrity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Nuclear Data Sheets for A=64

The evaluated experimental data are presented and evaluated for 13 known nuclides of mass 64 (Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se). For each nuclide, the best values combining all available data are recommended for spectroscopic properties. No excited states have been identified in 64 Ti, 64 As, and 64 Se. Only one excited state in 64 V as an isomer, and three in 64 Cr have been identified. Data for excited states in 64 Mn, 64 Fe, and 64 Co remain limited. 64 Ni, 64 Cu, and 64 Zn are the most studied nuclides through various reactions and decays, followed by 64 Ga and 64 Ge. In the opinion of the evaluators, there are several incomplete or discrepant aspects of the high-spin portion of the level scheme for 64 Zn above ≈5 MeV excitation which need to be resolved in further experiments using large γ-detector arrays. The decay schemes of 64 Ti β - , 64 As ε and 64 Se ε are unknown, while very little information is available for 64 V β - and 64 Cr β - . The decay schemes of 64 Mn β - , 64 Fe β - , 64 Co β - and 64Ge ε are somewhat better known, but still considered incomplete by evaluators. The decay scheme of 64 Ga ε decay is known in detail, however there is the possibility of additional levels populated above 4713 keV, as the Q(ε) value is 7171 keV. The β - , β + and ε decay modes of 64 Cu are well known, as this radionuclide is of great importance in applications, for example, as a dosimeter for neutron flux determination in nuclear reactors, and as a radiopharmaceutical for positron emission tomography (PET). This work supersedes earlier full evaluations of A=64 by 2007Si04, 1996Si12, 1991Si03, 1979Ha35 and 1974Au04.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Design and commissioning of an e-beam irradiation beamline at the Upgraded Injector Test Facility at Jefferson Lab

We report the Upgraded Injector Test Facility (UITF) at Jefferson Lab is a continuous-wave superconducting linear accelerator capable of providing an electron beam with energy up to 10 MeV. A beamline for electron-beam irradiation has been designed, installed and successfully commissioned at this facility, aimed at the degradation study of 1,4-dioxane and per- and polyfluoroalkyl substances (PFAS) in wastewater treatment. A solenoid with a peak axial magnetic field of up to 0.28 T and a set of raster coils were used to obtain a Gaussian beam profile with a transverse standard deviation of ~ 15.0 mm at the target location. Monte-Carlo simulations using FLUKA were carried out to calculate the total absorbed dose and the dose distribution in the sample volume inside the target cell. The simulations were benchmarked experimentally by dosimetry mapping using optichromic dosimeters. The results of the irradiation experiments showed a ~ 95% reduction of 1,4-dioxane in ultra-pure water for a dose of 1 kGy, demonstrating the potential of electron-beam irradiation towards addressing growing challenges in environmental remediation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Evaluation of neutron dosimetry capabilities with the MC-15 portable multiplicity counter

This work proposes a preliminary neutron dose rate estimation method for a neutron multiplicity detector through measurement- and simulation-based analyses. Uncharacterized neutron-emitting sources may be encountered in situations such as nuclear emergency response, safeguards, and treaty verification. These circumstances may present irradiation risk to personnel conducting field assay, search, and characterization measurements. It is therefore of interest to provide a field neutron dosimetry capability with the existing neutron multiplicity counting (NMC) capabilities. To date, no commercially-available neutron detection systems are capable of both accurate NMC and real-time neutron dosimetry. This work will focus on estimating dose rate using input from a single fielded NMC called the MC-15. The energy-dependent neutron detection efficiency response of the MC-15 was quantified in monoenergetic neutron simulations and evaluated in response to two neutron-emitting sources and to a polyethylene-moderated source. The results were compared to existing neutron dosimeters and established the proof of concept for further investigation of the MC-15 for dose estimation. Measurement results were also replicated in simulations; additional simulations were then conducted to expand upon the limited empirical data. The initial empirical results provided a conversion factor appropriate for use when measuring 252 Cf neutrons that is independent of polyethylene shielding presence and thickness. The simulated data sets were then used to evaluate a fit equation allowing estimation of the neutron dose rate for less restricted geometric configurations and dependent only on the distance between the source and the detector. Additionally, the energy dependence of the efficiency response indicates that further empirical evaluations could provide energy-dependent conversion factors for broader neutron dosimetry capabilities with a wider range of neutron-emitting sources.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Integrating HPC simulations and physical experiments to characterize the effects of gamma radiation on seismic protective devices

Seismic protective systems, composed of seismic isolators and dampers, can substantially reduce the effects of earthquake shaking on nuclear power plants and components therein. To enable the use of these devices to protect equipment inside a plant and close to a source of radiation, the U.S. Department of Energy (DOE) funded a project at the Idaho National Laboratory (INL) and the University at Buffalo to characterize the effects of absorbed gamma dose on their mechanical properties. An early task in the project was to determine the exposure time required in the INL Foss Therapy Services (FTS) 60 Co gamma irradiator to achieve a target absorbed dose in the materials used to construct isolators and dampers, including fluids, polymers, composites, and metals. This task required the novel integration of high-performance computing (HPC), Monte Carlo N-Particle (MCNP) simulations, and irradiation experiments using Fricke dosimetry. An MCNP model of the FTS irradiator at INL was developed and validated using Fricke dosimetry. Simulations of three experiments in the irradiator, two with Fricke vials only and one with Fricke vials and a large-size isolator, predicted the Fricke-measured absorbed dose rate to within 15% in all three cases, providing high confidence in the calculation of the gamma dose absorbed in the materials comprising the seismic protective devices. The simulations demonstrated that the effects of photon scattering on absorbed dose rate in the FTS irradiator are negligible for test articles installed close to the cobalt sources and near the rear of the irradiator. The validated MCNP model of the FTS irradiator is being used to support ongoing DOE-funded experiments on seismic protective devices and could be applied to future, non-seismic-related experiments. In conclusion, the novel validation process successfully deployed for the FTS irradiator at INL could be applied to other irradiators, requiring new MCNP models and simulations, and irradiation experiments using dosimeters.

42 - ENGINEERING↗

Optically stimulated luminescence dosimetry on mobile phone back protective glass

Protective glasses from the back of modern smartphones were tested as potential emergency dosimeters using optically stimulated luminescence (OSL). The glasses were shown to exhibit several favorable dosimetric properties including a strong sensitivity to radiation, a linear dose-response relationship from 0.1 to 8 Gy, a 3% reproducibility, and a minimum detectable dose in the low tens of milligray, even after significant fading. The OSL signal was observed to lose ~78% and ~90% of the original intensity over 1 h and 1 week, respectively; however, a fading curve was developed to correct for any such signal loss. A dose-recovery test demonstrated that reconstructed doses were capable of accuracies within 20%.

61 RADIATION PROTECTION AND DOSIMETRY↗

Infrared optically stimulated luminescence of rare earth doped CaF 2 and co-doped with Al or Li produced by combustion synthesis

The aim of this paper is to characterize the IRSL response of rare earth doped CaF 2 and co-doped with Al or Li, produced by the solution combustion synthesis (SCS). The results showed a high intensity of the infrared optically stimulated luminescence (IRSL) response for the samples of CaF 2 doped with Tb and co-doped with Li. For the samples of CaF 2 doped with Ce and co-doped with Li present better IRSL response than the samples co-doped with Al. It was also observed that the samples of CaF 2 :Ce,Li present a high fading with 50% reduction of the IRSL signal in a period of 60 min. In case of the CaF 2 :Tb,Li there is no significant fading during this period. For clinical dosimetry the fast fading of the dosimeter is not adequate. However, this feature can be exploited for other applications such as tamper-proof passive systems that detect and timestamp the movement of undeclared nuclear materials, by the estimation of the OSL response of materials with different fading rates.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Measuring very low radiation doses in PTFE for nuclear forensic enrichment reconstruction

Every country that has made nuclear weapons has used uranium enrichment to do so. Despite the centrality of this technology to international security, there is still no reliable physical marker of past enrichment in the open literature that can be used to perform forensic verification of historically produced weapons on gas centrifuges. We show that the extremely low radioactivity from uranium alpha emissions during enrichment leaves detectable and irreversible calorimetric signatures in the common enrichment gasket material PTFE, allowing for historical reconstruction of past enrichment activities at a sensitivity better than one weapon’s quantity of highly enriched uranium. Fast scanning calorimetry also enables the measurement of recrystallization enthalpies of sequentially microtomed slices, confirming the magnitude and the type of radiation exposure while also providing detection of tampering and a method for analyzing field samples useful for treaty verification. Furthermore, this work opens the door for common items to be turned into precise dosimeters to detect the past presence of radioactivity, nuclear materials, and related activities with high confidence.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

A neutron fluence map of the Los Alamos National Laboratory Godiva IV critical assembly

A neutron fluence map and a total ionizing dose map of the Los Alamos National Laboratory Godiva IV fast burst critical assembly was generated using passive reactor dosimetry, comprised of sulfur pellets and thermoluminescent dosimeters. Godiva IV is an unmoderated, fast burst, critical assembly constructed of approximately 65 kg of highly enriched uranium fuel alloyed with 1.5 % molybdenum for strength. The mapping was performed during a single 75.6 ºC temperature rise burst operation, with the top and sides of the cylindrical Godiva-IV Top Hat covered in passive dosimetry. Dosimetry was placed in a symmetric pattern around the Top Hat, with higher concentrations near the control rods and burst rod. A specific portion of the lower quadrant of the burst rod was mapped to confirm a testing region where the neutron fluence varied by no more than ± 5%. The results will be used to assess the neutron, gamma, and total ionizing dose environment in three-dimensional space around the assembly for higher fidelity experiment placement, active dosimetry positioning, and radiation field characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Using personal monitoring data to derive organ doses for medical radiation workers in the Million Person Study—considerations regarding NCRP Commentary no. 30

The study of low dose and low-dose rate exposure is of central importance in understanding the possible range of health effects from prolonged exposures to radiation. The One Million Person Study of Radiation Workers and Veterans (MPS) of low-dose health effects was designed to evaluate radiation risks among healthy American workers and veterans. The MPS is evaluating low-dose and dose-rate effects, intakes of radioactive elements, cancer and non-cancer outcomes, as well as differences in risks between women and men. Medical radiation workers make up a large group of individuals occupationally exposed to low doses of radiation from external x-ray/gamma exposures. For the MPS, about 100 000 United States medical radiation workers have been selected for study. The approach to the complex dosimetry circumstances for such workers over three to four decades of occupation were initially and broadly described in National Council on Radiation Protection and Measurements (NCRP) Report No. 178. NCRP Commentary No. 30 provides more detail and describes an optimum approach for using personal monitoring data to estimate lung and other organ doses applicable to the cohort and provides specific precautions/considerations applicable to the dosimetry of medical radiation worker organ doses for use in epidemiologic studies. The use of protective aprons creates dosimetric complexity. It is recommended that dose values from dosimeters worn over a protective apron be reduced by a factor of 20 for estimating mean organ doses to tissues located in the torso and that 15% of the marrow should be assumed to remain unshielded for exposure scenarios when aprons are worn. Conversion coefficients relating personal dose equivalent, H p (10) in mSv, to mean absorbed doses to organs and tissues, D T in mGy, for females and males for six exposure scenarios have been determined and presented for use in the MPS. This Memorandum summarises several key points in NCRP Commentary No. 30.

Environmental Sciences & Ecology↗

Reconstructing Hanford worker external doses from photons for epidemiology

The accurate reconstruction of external photon doses is essential for credible radiation epidemiology. This article presents the methodology used to derive dose estimates for 37 012 Hanford Site workers included in the Million Person Study. The approach employs historical dose records from the Hanford Radiation Exposure database and a previous epidemiology study. Bias correction factors specific to dosimeter type and period of use were applied and missing annual doses were estimated using a hierarchical nearby method to estimate deep dose equivalent for each worker. For early years with limited detection sensitivity, missed doses were quantified based on expected time-period-specific, low-dose statistical distributions. The revised dose estimates resulted in lower median and mean career doses than unadjusted data, while increasing the number of person-years with nonzero dose. Sensitivity analyses assessed the influence of bias in dosimetry measurements, missed doses and gap years on dose estimates. Differences in cumulative dose estimates between unadjusted and revised annual estimates are most prominent in the early operational years due to the highest bias during that time period.

dose reconstruction↗

12 C(p,p') 12 C Reaction ( E p =19.5–30 MeV) for Active Interrogation of Special Nuclear Material

Passive detection of special nuclear material (SNM) is challenging due to its inherently low rate of spontaneous emission of penetrating radiation, the relative ease of shielding, and the fluctuating and frequently overwhelming background. Active interrogation, the use of external radiation to increase the emission rate of characteristic radiation from SNM, has long been considered to be a promising method to overcome those challenges. Current active-interrogation systems that incorporate radiography tend to use bremsstrahlung beams, which can deliver high radiation doses. Low-energy ion-driven nuclear reactions that produce multiple monoenergetic photons may be used as an alternative. The 12 C ( p , p ' ) 12 C reaction is one such reaction that could produce large yields of highly penetrating 4.4- and 15.1-MeV gamma rays. This reaction does not directly produce neutrons below the approximately 19.7 MeV threshold, and the 15.1-MeV gamma-ray line is well matched to the photofission cross section of 235 U and 238 U . In this article, we report the measurements of thick-target gamma-ray yields at 4.4 and 15.1 MeV from the 12 C ( p , p ' ) 12 C reaction at proton energies of 19.5, 25, and 30 MeV. Measurements are made with two 3 -in. EJ-309 cylindrical liquid scintillation detectors and thermoluminescent dosimeters placed at 0 ° and 90 ° , with an additional 1.5 -in. Na I ( Tl ) cylindrical scintillation detector at 0 ° . We estimate the highest yields of the 4.4- and 15.1-MeV gamma rays of 1.65 × 10 10 and 4.47 × 10 8 sr - 1 μ C - 1 at a proton energy of 30 MeV, respectively. The yields in all experimental configurations are greater than in a comparable deuteron-driven reaction that produces the same gamma-ray energies— 11 B ( d , n γ ) 12 C . However, a significant increase of the neutron radiation dose accompanies the proton energy increase from 19.5 to 30 MeV.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗