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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Measuring Very Low Radiation Doses in PTFE for Nuclear Forensic Enrichment Reconstruction

Every country that has made nuclear weapons has used uranium enrichment. Despite the centrality of this technology to international security, there is still no reliable physical marker of past enrichment that can be used to perform forensic verification of historically produced weapons. 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 a detection of tampering and a method for analyzing field samples useful for treaty verification. 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.

Connick, Rachel↗
DOE OSTI2020

Test GammaPix™ Radiation Detection Software with Smartphones

GammaPix™ is a smartphone application that uses smartphone cameras to detect ionizing radiation. Brookhaven National Laboratory (BNL) evaluated the performance of the software in detecting radioactive materials. For this purpose, a series of tests were conducted at the Radiation Calibration Facility Annex at BNL. The experiments included dose measurements with three types of NIST-traceable calibration sources, 60Co, 241Am, and 137Cs. The recorded results by GammaPix™ were compared with the values measured by calibrated dosimeters to determine the detection thresholds. Several selected models of iPhones and Android cell phones with different pixel sizes and resolution were tested. In this paper, we present the test procedures and discuss the test results.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Low power sensor for NO x detection

Detection and capture of toxic nitrogen oxides (NO x ) is important for emissions control of exhaust gases and general public health. The low power sensor provides direct electrically detection of trace (0.5-5 ppm) NO 2 at relatively low temperatures (50° C.) via changes in the electrical properties of nitrogen-oxide-capture active materials. For example, the high impedance of MOF-74 enables applications requiring a near-zero power sensor or dosimeter, such as for smart industrial systems and the internet of things, with 0.8 mg MOF-74 active material drawing <15 pW for a macroscale sensor 35 mm 2 area.

Small, Leo J.↗
DOE OSTI2023

Godiva-IV Dosimetry Exercise 2022 Preliminary Results

Integral Experiment Request (IER) 538 is part of a series of dose characterization and nuclear accident dosimetry (NAD) exercises performed under the Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). This is the second NAD exercise using the Godiva-IV critical assembly and the third NAD exercise overall. The participating laboratories provided their own dosimeters that were mounted on the Lawrence Livermore National Laboratory (LLNL) BOttle Manikin ABsorption (BOMAB) phantoms and aluminum plates. The BOMABs and plates were placed at two, three, and four meters away from the center of Godiva. Alongside the NADs, there was a LLNL Passive Neutron Spectrometer (PNS), Atomic Weapons Establishment (AWE) PNS, and Y-12 Sphere present to measure the neutron dose from Godiva. Two irradiations were conducted to test the NAD performance from each laboratory and assesses their performance to the DOE-STD-1098-2017 part 515 criteria. Neutron and gamma doses were measured prior to this exercise. This work presents a model for the neutron and gamma dose respectively to serve as the reference value. A code written in C/C++/ROOT was used to fit the measured neutron and gamma dose with the new models. It was assumed that the neutron and gamma doses are proportional to the change in temperature of Godiva after a burst irradiation. Uncertainties for the reference values were calculated using error propagation of the model’s parameters. Preliminary results (within twenty-four hours) and final results were compared for each laboratory. On average of all the participating laboratories, 32% of neutron doses and 78% of gamma doses were outside the DOE standards. One laboratory did not report their dose readings and were not included in this average. There is a bias for a lower neutron dose and a higher gamma dose based on the distribution of results. In comparison with the past Godiva-IV NAD exercise, there is an improvement in neutron dose readings by 20%.

BOttle Manikin ABsorption (BOMAB)↗
DOE OSTI2020

Analysis of Exposure during Transportation of Waste to the Nevada National Security Site (NNSS) - 20445

Safety is a priority for the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program. This commitment to safety is integrated into requirements and work procedures at DOE sites across the U.S. that generate low-level radioactive waste, mixed low-level radioactive waste, and classified waste that is transported to the Nevada National Security Site (NNSS) for permanent disposal. Generators of this waste are responsible for ensuring it is packaged and transported in a safe and compliant manner. These generators and their contracted commercial carriers must comply with U.S. Department of Transportation (DOT) regulations; the Nevada National Security Site Waste Acceptance Criteria (NNSSWAC); and other applicable federal, state, and local regulations and requirements. This includes packaging and shipping the waste to meet DOT regulatory radiation dose limits to ensure public safety during transportation. NNSS waste generators have a solid safety record for complying with these stringent regulations as demonstrated by the fact that no shipments exceeding DOT dose rate limits were received at the NNSS Area 5 Radioactive Waste Management Complex (RWMC) during FY 2014 through FY 2018. A 2005 study conducted by the Desert Research Institute validated shipments to the NNSS comply with DOT radiation dose limits. In response to recent Nevada stakeholder requests for an updated study, an analysis was conducted to verify and provide evidence that the public continues to be safe from exposure to radioactive waste shipment to the NNSS. Radiation exposure from radioactive waste shipments in transit to the NNSS was evaluated. Three studies were performed: 1. Confirm compliance of waste shipments to the DOT radiation limits using dose rate surveys taken at the generator site and at the disposal site. 2. Estimate the dose to the public from waste shipments using the dosimetry results for disposal site personnel by comparing time of exposure for various individuals/scenarios. 3. Estimate the truck driver dose using the receipt dose rate surveys and time of exposure, and from dosimetry results. Study 1 confirmed compliance with the DOT dose rate limits for a sample of waste shipments (123 of 1,038) received during fiscal year (FY) 2017. Study 2 showed that the average annual radiation equivalent dose for disposal site personnel during FY 2017 was approximately 6.7% [0.415 mSv (41.5 millirem)] of the annual average dose that a person living in the U.S. receives from all sources [6.2 mSv (620 mrem)]. Maximum dose to a member of the public, a person resting/sleeping in an adjacent truck at a rest area to a waste shipment, was 0.0166 mSv (1.66 mrem), which is less than the dose received from a person taking a round trip from the east coast to the west coast of the continental U.S. on an airplane [0.07 mSv (7 mrem) per round trip]. Study 3 showed that the truck driver who made the most shipments of radioactive waste to the NNSS, who drove the most hours (estimated time of exposure 2,520 hours), during FY 2017 had a total dose of 0.15 mSv (15 mrem), which was obtained from their personal dosimeter. The annual dose recorded for disposal site personnel, the dose estimates for the public, and the truck driver dose were all a small percentage of the average radiation dose a person receives from all sources (e.g., background and medical radiation). At these levels, risk should only be described qualitatively. Radiation exposure from radioactive waste shipments to the NNSS were shown to be compliant with the DOT dose rate limits, safe, and very low (i.e., within background levels). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

The National Criticality Experiments Research Center: Capability Expansion and Experiments in the Last Three Years

The National Criticality Experiments Research Center (NCERC) is a general purpose criticality experiments facility located inside the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Critical experiments containing any special nuclear material, any enrich ment/separation, most physical forms, and any configuration are possible within the constraints of the defined safety basis. NCERC draws upon physical assets and experimental knowledge to solve some of the most difficult problems with respect to criticality safety, reactor physics, and reactor kinetics. In terms of physical assets, NCERC houses hundreds of kilograms of special nuclear material with a majority consisting of highly enriched uranium (HEU) and weapons grade plutonium (WGPu). NCERC is home to four critical assembly machines: Comet, Planet, Flattop, and Godiva IV. To support various derivative diagnostics on fissioning systems, NCERC houses a count room to measure irradiated samples and dosimeters. This paper will step through each of these capabilities explaining recently completed work and upgrades.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Conceptual Design of Neutron Sensor Qualification Device

The Advanced Sensors and Instrumentation Program at Idaho National Laboratory has been formulating strategies to qualify sensors for use in nuclear environments, particularly in irradiation experiments and advanced reactors. When qualifying neutron sensors for use in high-temperature environments, the wide range of neutron flux levels and representative energy spectra presents significant challenges. This paper discusses the development of the Neutron Sensor Qualification Device (NQD), which is designed to test neutron sensors in high temperature controlled environments with known neutron spectra, addressing the spatial and spectral complexities of neutron fluxes in reactor cores. The proposed NQD will be situated in the exposure room at the Armed Forces Radiobiology Research Institute, thus affording a unique capability to expose sensors to high neutron and gamma fluxes. To achieve thermal control, the device will utilize a radiation-hardened tube furnace, accommodating multiple sensors and neutron activation dosimetry wires. Titanium, iron, and cobalt dosimeter wires are chosen from the American Society for Testing and Materials and International Reactor Dosimetry and Fusion File libraries as references for providing energy-dependent fluence measurements. The design ensures precise sensor positioning to minimize mutual shielding and flux perturbation, which are evaluated via Monte Carlo N particle Transport Code (MCNP) simulations. These simulations have informed the development of guidelines on sensor placement within the NQD. The NQD is essential to the qualification of neutron sensors for advanced reactor technologies. It enables controlled testing of a statistically significant number of sensors, thereby supporting assessments of sensor performance across various neutron flux levels and temperatures. This paper highlights the detailed planning for the NQD prototype, along with its inaugural irradiation (scheduled for fiscal year [FY] 2025). The results from this initial testing will be fundamental in evaluating the device’s performance and establishing measurement uncertainty for in-pile neutron sensor measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

A317_Fast and Thermal Neutron Spectrum Dosimetry Measurements in the Advanced Test Reactor large-B and small-I Positions Following the Sixth Core Internals Change-out

The Advanced Test Reactor (ATR) has a wide variety of irradiation positions that have had experiments that were developed by users from around the world. Most experiment irradiations rely on thoroughly benchmarked numerical models. However, some irradiation positions in ATR are not as well- characterized and are complicated by spectral perturbations from control cylinder orientation. The “small I” and “large-B” irradiation positions are located nearby control cylinders and suffer from these flux perturbations from control cylinder orientations that change during an irradiation cycle to maintain the desired core power distribution. Models predicted that these types of position would exhibit both spectral shifts and amplitude changes in neutron flux, but few measurements have been conducted to benchmark these predictions. Recently, requalification testing was performed to confirm the operational readiness of the ATR following the completion of the Core Internals Change-out (CIC). These tests provided a unique opportunity to validate the analytical methods used to simulate the ATR because nearly all components in the reactor were in a clean as-built state, significantly reducing modelling uncertainties. One subset of the testing included characterization of the fast and thermal neutron flux in the “small-I” and “large-B” positions using silver, cobalt, and nickel neutron dosimetry. In contrast to typical irradiation cycles, the control cylinders were held in position during the post-CIC nuclear testing. The specific activity of these dosimeter wires was measured following two low-power tests with different control cylinder positions.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

Comparison of the performance of TLD, OSL and RPL personal dosimetry systems to the American National Standard Institute (ANSI) test categories

The Laboratory Accreditation Program (LAP) tests the capability and performance of the United States Department of Energy's (USDOE) facilities to accurately measure and quantify the whole-body and extremity radiation equivalent doses to the occupational workers. The dosimetry methods used in personal dosimetry can be Thermoluminescence (TL), Optically Stimulated Luminescence (OSL) and Radiophotoluminescence (RPL), among other techniques available. Currently, the TL (LiF:Mg,Ti) and OSL (Al 2 O 3 :C) dosimetry systems are DOELAP-accredited for occupational dose measurements and regulatory reporting. In this study, the performance of a beryllium oxide (BeO) OSL dosimetry system and of a silver-doped phosphate glass RPL dosimetry system is compared against the performance of an accredited LiF:Mg,Ti TL dosimetry system. The bias and standard deviation in each of the performance test categories are compared between these three systems for exposures to photons, beta, and mixed radiation fields with the criteria established by the American National Standard Institute (ANSI) and Health Physics Society (HPS) N13.11-2022 standard. The practical implementation of the dosimetry program and its equivalency to the occupational personal equivalent dose H p (0.07) and H p (10) measurements were evaluated. The TL, OSL and RPL dosimetry systems passed the ANSI performance test criteria for H p (0.07) and H p (10) occupational dose measurements and met the DOELAP accreditation requirements.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Project 57 Air Monitoring Report: (Jan 1 - Dec 31 2018)

During the late 1950s, the Atomic Energy Commission (AEC) (now the U.S. Department of Energy [DOE]) conducted a series of safety experiments to determine if a nuclear device subjected to a large conventional explosives detonation would result in a nuclear yield. The AEC obtained temporary use of a large portion of western Emigrant Valley from the U.S. Department of Defense (DOD) for one of these experiments: Project 57. Following the Project 57 safety experiment, the AEC fenced the contaminated area and returned control of the surrounding land to the DOD. Emigrant Valley is part of the Nevada Test and Training Range (NTTR). For safety and security reasons, access to the NTTR is controlled using both physical (i.e., fences) and administrative (e.g., signs and postings) controls. Therefore, the public cannot access the Project 57 site and there are no known human receptors that routinely access the site. Project 57 was detonated on April 24, 1957, in Emigrant Valley approximately 13 mi (21 km) northeast of the north end of Yucca Flat. This test was undertaken to develop: (1) a means of estimating immediate distribution and long-term redistribution of plutonium dispersed during a nonnuclear detonation; (2) biomedical evaluation techniques for use in likely plutonium-laden environments; (3) methods of decontaminating ground areas, pavements, and building materials; and (4) alpha survey instruments and field monitoring procedures to promptly estimate contaminant deposition. Although the test did not result in the fission of nuclear materials, it did disseminate plutonium across the ground surface. Various radiological surveys have been performed in the area since Project 57 was conducted. The AEC delineated the area containing radioactive material based on radioactivity surveys performed shortly after the test was conducted. A designated Contamination Area (CA), marked by a fence, was established later. The distribution of americium-241 (Am-241) in the area was determined again in a 1997 flyover, which showed Am-241 ranging from as much as 70,000 counts per second (cps) at ground zero to background (<70 cps). This survey also documented Am-241 radioactivity on the ground surface beyond the east side CA fence at levels of up to 150 cps. In 2007, the DOE expanded the CA by posting “Contamination Area” signs 200 ft to 400 ft (60 m to 120 m) beyond the original fence, which formed a new CA boundary.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Evaluation report on the Thermo Scientific TruDose (neutron,gamma) EPD3

This report summarizes an evaluation study on the Thermo Scientific TruDose n/γ EPD3 by LANL’s RP-SVS group. Two EPD3s were received for assessment (SNs 7500100 and 7500108) along with an EPD3 desktop reader. The evaluation included intercomparisons with Thermo’s N2 EPD which is in wide circulation at LANL but is no longer commercially available. A favourable assessment of the TruDose EPD3 would help support it as a viable option to replace the N2 EPD. The data and observations included in this report could also serve as the basis of a Technical Basis Document for the use of the EPD3 at LANL.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Instrumentation for the In-Core Real-Time Mechanical Testing of Structural Materials (INCREASE) Project

Idaho National Laboratory (INL), in collaboration with the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), the French Atomic and Alternative Energies Commission (CEA), the Joint Research Center (JRC), the Nuclear Research and Consultancy Group (NRG), and the Research Center Rez (CVR), started a Joint Experimental Program (JEEP) project that operates within the Nuclear Energy Agency’s Framework for Irradiation Experiments (FIDES II) program in order to develop capabilities for the in-core real-time mechanical testing of structural materials. This effort will focus on designing a shared capsule capable of housing a variety of in-core mechanical testing instrumentation allowing enhanced experiments for the material science community. The outcome of the project would be high-priority, stress relaxation data for stainless-steel-based materials provided by EPRI and CEA. Stress relaxation is a major phenomenon that contributes to material degradation in nuclear reactor components. Currently, nuclear material stress relaxation is assessed both before and after irradiation, using complex and costly post-irradiation examination (PIE) activities. In-situ data would support the development of precision modeling and simulation of this degradation phenomena and would provide validation and benchmarking for existing models using the PIE data. As part of the U.S. Department of Energy Advanced Sensor and Instrumentation (ASI) program, INL has fabricated and tested out-of-core mechanical test instrumentation. This instrumentation was designed for easy adaptation to the irradiation capsule proposed under this JEEP, and can be deployed to measure real-time stress relaxation under pressurized-water reactor (PWR) conditions. This initial effort will partially serve to replace the testing capabilities lost because of shutting down the Halden Boiling Water Reactor (HBWR). The project would provide these capabilities to the international community via a shared capsule design that is easily adaptable to additional material test reactors. The design features will incorporate expansion to PWR and non-light-water reactor (LWR) environments that will be developed in future work. The capsule and instrumentation will be demonstrated in the Massachusetts Institute of Technology Reactor (MITR) for phase I and the Petten High Flux Reactor (HFR) for phase II irradiations to deliver real-time stress relaxation data on high priority stainless steel structural materials.

36 MATERIALS SCIENCE↗
DOE OSTI2024

INSTRUMENTATION FOR IN-CORE REAL-TIME MECHANICAL TESTING OF STRUCTURAL MATERIALS (INCREASE) PROJECT

Idaho National Laboratory (INL), in collaboration with the Electric Power Research Institute (EPRI), Nuclear Regulatory Commission, French Atomic and Alternative Energies Commission (CEA), Joint Research Centre, Nuclear Research and Consultancy Group, and Research Center Rez, is the operating agent of the In-Core Real-Time Mechanical Testing of Structural Materials (INCREASE) joint experimental program (JEEP) project that operates within the Nuclear Energy Agency (NEA)’s Framework for Irradiation Experiments (FIDES-II) program. The first objective of the INCREASE project is to design a shared capsule that can house a variety of in-core mechanical testing instrumentation for enabling smart irradiation experiments. The second objective is to collect in core stress relaxation data pertaining to high priority stainless steel (SS) materials under light water reactor conditions with target temperature of 340°C +/-20°C and a minimum neutron damage of 2 displacement per atom (dpa). The INCREASE project will be demonstrated in the Massachusetts Institute of Technology Nuclear Research Reactor (MITR).

36 MATERIALS SCIENCE↗
DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using OpenMC

This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.

61 - RADIATION PROTECTION AND DOSIMETRY↗