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All sources NASA NTRS DOE OSTI

DOE OSTI2017

Development of Membrane Distillation Technology Utilizing Waste Heat for Treatment of High Salinity Wastewaters

Membrane distillation (MD) can serve as a cost effective method to treat high salinity wastewaters, especially if waste heat is utilized for its operation. The goal of this work is to study the feasibility of employing MD technology for the treatment of high salinity wastewaters which are generated as a result of natural gas production and active reservoir management for CO 2 sequestration. This study also aimed at quantifying the availability of waste heat which could be used as the primary source of thermal energy to drive the MD process. Bench scale studies were carried out with several commercially available membranes using both synthetic and actual high salinity wastewaters. The membranes tested with real wastewater samples exhibited excellent rejection of dissolved ions, including naturally occurring radioactive material (NORM). Long term bench scale experiments with real produced water from the Marcellus shale play revealed that iron oxide could precipitate on the feed side of direct contact MD (DCMD) module but had a negligible effect on the performance of the system. Based on the bench scale experiments, an ASPEN Plus simulation, which incorporated a mathematical model with fundamental heat and mass transfer equations, was developed and used to evaluate the process energy requirements. Despite low operating temperature and pressure for MD, it had a large thermal energy requirement for operation, which could make this technology expensive compared to reverse osmosis (RO) and forward osmosis (FO). However, it is important to note that RO and FO have limited applicability for high salinity shale gas wastewater. Moreover, MD could be integrated with renewable energy sources such as wind or solar energy or industrial waste heat sources to offset the thermal energy requirements and enhance the economics of wastewater treatment. Natural gas compressor stations (NG CS) have been proposed in this study as the source of waste heat based on the estimate of quantity and quality of available waste heat in the exhaust streams of gas turbine and internal combustion compressor engines. Using the actual installed capacity and location of compressor stations in the U.S., this analysis revealed that an average of 610 TJ/day of thermal energy is available at existing NG CS and that this source of waste heat is of high quality with temperatures above 645 K. A systems-level integration of MD with waste heat from NG CS revealed that approximately 56% of available waste heat at NG CS in Pennsylvania is sufficient to concentrate (from 10% to 30% salinity) all wastewater associated with shale gas extraction. Techno-economic assessment (TEA) accounting for capital as well as operating and maintenance costs with and without waste heat integration was conducted for a hypothetical 0.5 million gallons per day (MGD) DCMD plant using experimental and modeling results from this study. TEA revealed that the total cost of shale gas wastewater treatment using MD is $5.7/m 3 feed for the base case scenario in which the thermal energy requirements are met by external steam, however, the total cost can be reduced to $0.74/m 3 feed when integrated with free waste heat from NG CS. Sensitivity analysis revealed that thermal energy cost and feed TDS are the most important factors in total cost of high salinity wastewater treatment using DCMD.

03 NATURAL GAS↗

DOE OSTI2017

Turbo FRMAC 2018

The Turbo FRMAC (TF) software program is the software implementation of the science and methodologies utilized in the Federal Radiological Monitoring and Assessment Center (FRMAC). The software automates the calculations described in volumes 1 of "The Federal Manual for Assessing Environmental Data during a Radiological Emergency" (2015 version). In the event of the intentional or accidental release of radioactive material, the software is used to guide and govern the response of the Federal, State, Local, and Tribal governments. The manual, upon which the software is based, is unclassified and freely available on the Internet.

Fulton, John↗

DOE OSTI2015

Analysis of dynamic radiation level changes using surface networks [Book Chapter]

The Fukushima nuclear accident reminds us of severe risk of radioactive substances. Citizen scientists voluntarily collect and share radiation data using geo-tagged sensors for radiation preparedness. However, radiation levels are affected by a number of factors including for example weather conditions, naturally occurring radioactive materials (NORM), and large marble structures. It is therefore difficult to determine whether a higher radiation level comes from a normal variation in background or not. This research analyzes the radiation changes using surface networks that can be used to characterize complex surfaces. A new algorithm has been developed to identify salient peaks and their hills by merging insignificant peaks recur-sively. Salient peaks and their hills are converted into graphs. The structural similarities of graphs were compared over time. The radiation measurements in the city of Koriyama, Japan were analyzed as a case study. The results demonstrated that structural analysis of dynamic radiation levels revealed stable changes, while numeric analysis of radiation levels presented statistically significant differences. This method is able to detect radiation level changes irrespective of background variations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2018

Nondestructive Pressure Measurement Technique for Irradiated Nuclear Fuel Rods

Commercial nuclear fuel rods are pre-pressurized with helium before irradiation. During irradiation, the rod internal pressure (RIP) increases due to cladding dimensional changes and pellet fission gas release. RIP is a limiting parameter for rod operation and is a barrier to achieving higher fuel burnup. Currently, the only method for RIP measurement is rod puncture which is a destructive method. Because of the challenges and cost associated with handling irradiated fuel, the existing database on RIP is extremely limited. This paper presents progress towards developing a novel acoustic-based nondestructive examination method to accurately measure the internal gas pressure and gas composition in pressurized containers without the need for puncturing the pressure boundary, with an emphasis on applications to spent nuclear fuel (SNF) rods and radioactive material dry storage canisters.

Montgomery, Rose↗

DOE OSTI2019

Radiation Induced Redox Chemistry of Californium-249

A complete understanding of californium chemistry necessitates knowledge of its radiation induced redox behavior, owing to its inherent nuclear instability propagating self-radiolysis. Only a handful of studies have investigated californium radiation chemistry, due to lack of availability and difficulty associated with handling highly radioactive material. To date, reaction rate coefficients have only been experimentally determined for the reduction of Cf(III) by the hydrated electron (e¯aq, k > 3 × 109 M–1 s–1), and subsequent decay of the corresponding transient Cf(II) (k = (7 ± 1) × 105 s–1)[1]. However, there are a number of other important transient radiolysis products radiolytically generated in solutions pertinent to Cf manipulations, e.g. the hydrogen atom (H•, Eo = 2.31 V), hydroxyl radical (•OH, Eo = –2.73 V), and nitrate radical (NO3•, Eo = –2.3 – –2.6 V). These species are more than capable of influencing the redox behavior of Cf, and have been shown to do so with a number of actinides, e.g. neptunium and americium. Here we present a time-resolved picosecond pulsed electron irradiation study for californium-249.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

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 OSTI2018

EPA and Idaho National Laboratory create first of its scale

Advancing the science and engineering of decontaminating water distribution systems and safely disposing of high-volumes of contaminated water are high priorities for the EPA. To help address these science gaps, EPA homeland security researchers have developed the first-of-its-scale Water Security Test Bed (WSTB). Water is an important component of everyday life. We rely on safe, clean water for our health and the health of the environment. Protecting this vital resource is critical to living safely and securely. The U.S. Environmental Protection Agency is responsible for working with water utilities to protect water systems from contamination and to clean up systems that become contaminated. Purposeful or inadvertent distribution system contamination can result in large amounts of infrastructure and water that must be cleaned. Water distribution systems contamination can be caused by, for example, natural disasters such as Superstorm Sandy or by individuals hoping to cause harm. The WSTB, constructed at the Department of Energy’s (DOE) Idaho National Laboratory, replicates a section of a typical municipal drinking water system including a piping system with roughly 445 feet of pipe and two fire hydrants laid out in an “L” shape using 40-year-old, eight-inch cement mortar lined, ductile iron pipes, excavated after twenty years of use for water conveyance. Researchers built the WSTB above ground for easy access during experiments, and to facilitate fast leak detection. Over the next several years, EPA and partner researchers will conduct experiments using simulants of various biological, chemical, and radioactive materials that simulate high toxic versions of these agents. At this “full” sized system, researchers will demonstrate approaches to contamination detection, sensor and model testing, infrastructure decontamination, water treatment and cyber testing developed at lab and pilot scale. Researchers are also able to verify if laboratory results that were effective in controlled conditions are as effective in full-scale testing. Having the ability to conduct bench scale studies in a real world setting can help address specific research gaps, and help water researchers understand conditions in which real world drinking water systems operate.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2020

Analysis of RF high power sources for 1MW – range, 10 MeV CW industrial accelerator.

Cobalt-60 is the dominant source of ionizing radiation for sterilizing products in the $200B/year medical device industry. Security and accessibility concerns around this radioactive material - sourced from countries including Russia and China - present national security and economic risks. There is a continued national urgency to replace such high-risk sources with low-risk alternatives such as electron or x-ray beams driven by accelerators. While e-beam sources have been in use for many years, they still represent a small portion of the market. X-ray sources are becoming economical alternatives. Meeting the processing capacity required to replace Co-60 requires a step-change in the capability, efficiency, and overall operating cost of electron accelerators to make adoption feasible. The radiofrequency (RF) power that drives accelerating cavities represents significant capital and operating costs for these accelerators. For accelerator-based systems to be a viable alternative for these applications, RF systems must be economical to manufacture, deploy, and operate.

43 PARTICLE ACCELERATORS↗

DOE OSTI2021

Intelligent Consequence Control by Aerial Reconnoiter Using Unmanned Systems

In a nuclear accident, or in the aftermath of an improvised nuclear device or radiological dispersal device detonation, radioactive materials may pose a severe health threat to individuals near the site of the incident. The Intelligent Consequence Control by Aerial Reconnoiter Using Unmanned Systems (ICARUS) does the dull, dirty, and dangerous (and, sometimes, deep) work of mapping and characterizing valuable information to enable a safe and appropriate response. ICARUS provides timely answers to the questions, “What is it?” “Where is it?”, and “How far does it extend?” Using commercial, off-the-shelf proven components, the integrated chemical, biological, radiological, nuclear, and explosives detection architecture for our drones provides a solution for collecting radiological, chemical, and optical information on developing complex situations, returning actionable data for determining if it is safe for further response efforts. ICARUS can determine and transmit the location, identity, and intensity of radionuclide contamination to a remote base station. It can deliver on the unexpected.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2021

A Brief Overview of Radiochemistry: Californium-252: Unique and Versatile Man-made Radioisotope

This chapter summarizes radiochemistry and its applications with a focus on areas where computational science can be beneficial. This is presented for a target audience of computational scientists and chemists, and the general public who are interested in radiochemistry topics that may benefit from the use of computational methods. Past, present, and future applications of radiochemistry are presented along with a brief examination of the field’s workforce. Discussions of the current state of several radiochemistry specialty areas are also included. These discussions generally exemplify topics that are readily amenable to computational tools. Topics discussed include medical uses for radioactive materials and related computational needs; the current and future trends in the separations of the f-block elements; production and uses of Californium-252; and the need for exascale computing to solve current problems in the chemistry of Americium and Plutonium.

Ezold, Julie↗

DOE OSTI2021

A Brief Overview of Radiochemistry: f-Element Separations: The State of the Art and Future Directions

Kimberlin, Ashleigh↗

DOE OSTI2022

Sodium Extraction from Full-Length Fermi-1 Blanket Assembly

Enrico Fermi Nuclear Generating Station is a nuclear power plant on the shore of Lake Erie. On October 5, 1966, Fermi 1, a prototype fast breeder reactor, suffered a partial fuel meltdown (no radioactive material was released). As a result, 34 metric tons of sodium-bonded blanket material from the decommissioned Fermi-1 reactor were shipped to and are currently stored at the Idaho Nation Lab (INL). This material is currently awaiting removal from the state of Idaho by 2035. In its current state, the Fermi-1 material is not suitable for disposal due to the reactive characteristic of its bond sodium. This work demonstrates removal of bond sodium from an entire full-length unirradiated Fermi-1 outer radial blanket assembly using a gravity assisted melt-drain-evaporate (MEDE) treatment process. Quantitative and qualitative analyses were performed to characterize the extent of sodium removal from the blanket material following MEDE treatment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

316L Stainless Steel Wire Arc Additive Manufacturing and Characterization for Potential SNF Canister Production

In this paper, 316L stainless steel (316L SS) straight walls, curved walls, and vessels were printed by wire arc additive manufacturing (WAAM) using the 316L SS welding wire, to demonstrate the feasibility of spent nuclear fuel (SNF) canisters using this advanced manufacturing technique. Helium leak test with leak tight criteria of 1 × 10-7 ref-cc/s were performed on printed vessels with various wall thickness, and testing pressures were from 0.345 MPa (50 psi) to 1.724 MPa (250 psi). Printed wall microstructures were characterized using optical microscopy (OM) and scanning electron microscopy (SEM). Tensile specimens were machined out from printed walls along the length, height, and thickness directions, respectively, and tested at different temperature conditions, room temperature, 80 °C, 150 °C and 250 °C, respectively. Results showed t hat, all five WAAM-printed canisters with different thickness passed the helium leaktight criteria of radioactive materials containers by the ANSI N14.5; Dispersed micrometer-level size inclusions/voids were observed at printed wall cross sections, dendritic microstructures were found with grain size variation through the welding torch shifting direction, and very small amount of δ-ferrite and σ-phase were also observed in those cross sections; The printed walls were ductile, all tensile specimens broke with large plastic deformation/elongation and without brittle failure, and printed specimens yield strengths and tensile strengths under room temperature and 80 °C conditions are higher than minimum values of 316L SS for pressure vessels and for general applications at room temperature required by the ASTM standard, respectively.

Tang, Wei↗

DOE OSTI2023

Towards an Era of Low-Temperature Integral Critical Experiments: Surrogate Testing of Low-Temperature TEX Configurations

United States and IAEA regulations for the transportation of radioactive materials require that fissile materials must be subcritical under expected ambient temperatures ranging from -40°C to 38°C. While historical experiments have been conducted at room and elevated temperatures, no criticality integral benchmarks exist for temperatures below room temperature. The Low-Temperature Thermal Epithermal eXperiments (LT-TEX) aim to provide benchmark evaluations down to -40°C. The LT-TEX design incorporates stacked HEU plates interspersed with varying thicknesses of polyethylene disks to provide varying levels of moderation, with five configurations at 20°C having already been benchmarked. A novel vacuum/cryogenic chamber was fabricated to contain the LT-TEX configurations in a thermally isolated environment and cool it to -40°C. Prior to testing with fissile materials, testing with surrogate fuel plates was performed to characterize the thermal properties of the system, ensure no condensation forms on the stack, and to test that the uranium plates would not warp, or fracture during the thermal cycling. Surrogate testing has characterized the cryogenic chamber and verified its viability to perform the benchmark experiments, allowing for the next steps in the LT-TEX experimental campaign. The results of the surrogate testing are discussed and compared to heat transfer calculations performed as part of the experiment design process, to establish confidence in the thermal behavior of the system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Evaluation and improvement on external-hazard proof of JSFR fuel handling system

Responding to the the Fukushima Dai-ichi nuclear power plant (1F-NPP) accident, the earthquake and the tsunami proof of the fuel handling system (FHS) in Japan sodium-cooled fast reactor (JSFR) is studied. In the earthquake proof estimation, the margin of seismic resistance against the earthquake of the 1F-envelop condition and the sloshing behavior in the EVST is estimated. In terms of the tsunami proof, the scenario to lead fuel subassemblies into the stable cooling state and the potential of the cooling system is introduced in case of loss of the emergency power supply. As a result, it is clear that JSFR FHS originally could already be prepared to have the potential to prevent the release of radioactive material. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

APR1400 In-Core Instrumentation Penetration Failure by Zirconia Melt

Light water reactors have many penetrations at the reactor vessel lower head which are called ICI (in-core instrumentation tubes), ICM-GT (in-core monitoring guide tubes), CRGT (control rod guide tubes) and drain tubes. For Korean PWRs (pressurized water reactors), there are dozens of bottom-up type ICI penetration tubes (45 for OPR1000 and 61 for APR1400). For the typical BWRs (boiling water reactors), hundreds of penetrations are positioned at the reactor vessel lower head due to the structural complexity at the reactor upper head. These penetrations are regarded as the most vulnerable structures during a severe accident as they could be failed by high temperature of melt (or debris) and its decay heat. That is, tube rupture and/or tube ejection could occur because there exist an empty hole inside the penetration tube which can be a melt flow path, and they are attached to the inside of the reactor vessel by a partial weld with a gap between the tubes and reactor vessel. Therefore, the research on the penetration failure is of importance, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. This paper focuses on the experimental investigation on the tube ejection failure for the APR1400 ICI penetration. The tube ejection experiments were performed as it was exposed to high temperature of zirconia melt and high pressure. The tube ejection failure experiments were performed for the APR1400 in-core instrumentation penetration nozzle. Zirconia melt was used as a simulant of molten corium and interacted with the penetration specimen. The specimen was pressurized up to 2.6 bar during the interaction with melt to induce the tube ejection. The penetration weld was heated up to its melting temperature, and the penetration tube and weld above the reactor vessel surface were eroded severely by the melt. However, even though it was estimated that the weld failure occurred and the reactor vessel with stainless steel cladding was ablated severely by the melt, the tube ejection did not take place, and consequently the integrity for the APR1400 ICI penetration was confirmed in the present experimental conditions. The goal of this research is to improve the penetrations failure models in MAAP5 codes based on the experimental data. To achieve this goal, it is necessary to increase the heat flux and pressure acting on the penetration tube to find the conditions for the tube ejection failure. Moreover, the penetration failure experiments will be performed by delivering melt into the penetration specimen to simulate the melt relocation process in a real severe accident situation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

What is involved in radiation protection at accelerator facilities?

Radiation protection for accelerator facilities involves a range of actions to operate the facility within safe boundaries (often called an accelerator safety envelope) and it starts from the very early stage of facility development. Each facility will establish fundamental safety principles, requirements, and measures to control the radiation exposure to people and the release of radioactive material to the environment.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2023

A Comparison of As-loaded Dose Calculations to Measured Dose Rates

Demonstrating that the radiation dose fields surrounding spent nuclear fuel (SNF) storage and transportation systems meet the applicable limits set forth in 10 CFR 72 for storage and 10 CFR 71 for transportation is essential for the safe handling of radioactive material. The Used Nuclear Fuel-Storage, Transportation, and Disposal Analysis Resource and Data System (UNF-ST&DARDS) [2] is used to provide realistic estimates of SNF-related safety margins. The UNF-ST&DARDS dose rate analysis approach differs from that used in typical licensing approaches, which use design-basis assemblies with bounding source term characteristics to demonstrate the packaging design complies with the regulations. These bounding licensing approaches can increase the time needed to qualify fuel for loading into dry storage and the time a loaded system must wait prior to transportation. UNF-ST&DARDS dose rate assessments allow quantification of realistic, uncredited safety margins associated with actual fuel loading compared with the regulatory limits. While realistic estimation of the dose field surrounding SNF systems may allow for additional flexibility in operations, it is essential to understand how these predictions compare to measured doses. The U.S. Department of Energy Office of Integrated Waste Management and the Prairie Island Indian Community conducted a transportation dose assessment to estimate the site-specific incident-free radiation doses from shipping SNF by rail from the Prairie Island Nuclear Generating Plant (PINGP) through the Prairie Island Indian Community Reservation and Trust Land [5,6]. For that effort, the dose rates were obtained for 50 TN-40 and TN-40HT systems in storage configurations. This work compares the predicted dose rates from UNF-ST&DARDS as-loaded calculations with dose rates measured from 50 SNF storage systems at PINGP. The remainder of this paper discusses the data obtained for the evaluation, the modeling methods, and the results of the calculations.

spent nuclear fuel (SNF), UNF-ST&DARDS, Validation↗