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At least 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Rapid Analytical Methodology for Chloride Molten Salt Reactor Safeguards and Process Monitoring

Technologies that enable near real-time isotopic analysis of advanced molten salt reactor (MSR) fuels are critically needed to safeguard these reactors, increase their operational efficiency, and enable their widespread deployment with confidence. We present a systematic approach to developing near real-time dissolution and chemical isolation of U, Pu, and major fission products from highly radioactive chloride molten salt samples. Chemical yields greater than 95% were observed for both uranium and the lanthanides. Interference reduction enabled the detection and quantification of key diagnostic isotopes (including 112Ag, 147Nd, and 153Sm) that were previously undetectable in the original sample. The results from this initial scoping study lay the foundation for the development of future automated systems that can enable cost-efficient, near-real time chemical separation and analysis of extremely highly radioactive molten salt samples.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI1994

Nuclear Safety [Vol. 35, No. 2, July-December 1994]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 179 Consideration of Postaccident Consequences in the Determination of Safety Objectives for Future Nuclear Power Plants in France, D. Queniart, A. Sugier, and J. Lochard; ACCIDENT ANALYSIS: 187 Nuclear Safety Research: The Phebus FP Severe Accident Experimental Program, P. von der Hardt, A. V. Jones, C. Lecomte, and A. Tattegrain; 205 Containment Performance Analysis of the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory, S. H. Kim, R. P. Taleyarkhan, and V. Georgevich; 213 Assessment of Fission Product Deposits in the Reactor Coolant System: The DEVAP Program, G. Le Marois and M. Megnin; 222 Erratum to “A Review of the Available Information on the Triggering Stage of a Steam Explosion," Vol. 35, No. 1, CONTROL AND INSTRUMENTATION: 223 Effects of Normal Aging on Calibration and Response Time of Nuclear Plant Resistance Temperature Detectors and Pressure Sensors, H. M. Hashemian; DESIGN FEATURES: 235 Defense in Depth Against the Hydrogen Risk—A European Research Program, F. Fineschi; ENVIRONMENTAL EFFECTS: 246 Technical Note: A Preliminary Analysis of the Risks to Hong Kong Resulting from Potential Accidents of Daya Bay Nuclear Power Plant, Z. Shi and X. Wei; OPERATING EXPERIENCES: 253 Reactor Shutdown Experience, Compiled by J. W. Cletcher; SPECIAL SECTION ON TMI-2 VESSEL INVESTIGATION PROJECT: 256 Three Mile Island—New Findings 15 Years After the Accident A. M. Rubin and E. Beckjord; 269 Relocation of Molten Material to the TMI-2 Lower Head, J. R. Wolf, D. W. Akers, and L. A. Neimark; 280 Insight Into the TMI-2 Core Material Relocation Through Examination of Instrument Tube Nozzles, L. A. Neimark; 288 Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head, D. W. Akers and B. K. Schuetz; 301 Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head, D. R. Diercks and G. E. Korth; 313 Margin-to-Failure Calculations for the TMI-2 Vessel, J. Rempe, L. Stickler, S. Chavez, G. Thinnes, R. Witt, and M. Corradini; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 328 1993 Accident Sequence Precursor (ASP) Program Results, L. N. Vanden Heuvel, J. W. Cletcher, D. A. Copinger, J. W. Minarick, B. W. Dolan, and P. D. O’Reilley; RECENT DEVELOPMENTS: 339 Reports, Standards, and Safety Guides, D. S. Queener; 345 Proposed Rule Changes as of June 30, 1994; ANNOUNCEMENTS: 234 Thirty-First Annual Meeting of the National Council on Radiation Protection and Measurements; 356 1995 International Incineration Conference; 358 Fifth International Controls and Instrumentation Conference; 359 ANS International Topical Meeting on Safety of Operating Reactors; 359 Fifth International Conference on Nuclear Criticality Safety; 360 International Conference on Probabilistic Safety Assessment Methodology and Applications; 351 The Authors; 357 Reviewers of Nuclear Safety, Vol. 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Small-Scale Recycling of Irradiated Uranium and Transuranic Elements (Np, Pu) Using 3D-Printed Centrifugal Contactors

Niowave’s technology employs super-conducting linear accelerators to induce fission on low-enriched uranium targets followed by subsequent radiochemical processing to purify 99 Mo as well as a number of other fission products. Argonne’s Radiochemistry group (CFCT), in partnership with Niowave and UNLV, will be managing the TCF project to develop a closed-cycle loop for Niowave’s uranium targets. The roles of the DOE national laboratory partner will be to 1) develop a basic chemical understanding of the separations and purifications required to meet industry standards, 2) initiate the additive manufacturing (AM) of the centrifugal contactors, 3) develop a process flowsheet using Argonne computer codes.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Measurements of Extinct Radionuclides in Filter Paper Samples

Radiochemical diagnostics rely on measurements of short-lived fission product isotopes. Once months to years have passed since isotope formation, the most useful short-lived signatures have decayed away and become extinct. Recent work at our lab has developed new methods to reconstruct the concentrations of extinct radionuclides using mass spectrometry to measure the stable daughter elements formed from radioactive decay. However, the applicability of these techniques to different types of samples has not been fully investigated.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

AGR-5/6/7 Operations Update for Technical Coordination Team (Apr 2020)

This is a powerpoint presentation designed to update the Technical Coordination Team (TCT) associated with the AGR irradiation experiment program on operations of the AGR-5/6/7 experiment over the past six months (since the last TCT meeting). The presentation primarily deals with operational challenges to failed fuel particles in capsule 1. Specifically, how to keep fission products from capsule 1 from entering the other four capsules in the experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

LANL's Contribution to the Design and Preparation of the ANL Bubble Experiment #2

In collaboration with Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL) is assisting in the design and development of portions of the second ANL Bubble Experiment to be performed in late 2020 at ANL. The ANL Bubble Experiment, as called in this report, is a series of direct electron irradiations of a uranyl sulfate solution to produce radiolysis-induced gas bubbles of hydrogen and oxygen. The gas bubbles formed in the solution enhance mixing and heat transfer. The study of the dynamics, shape, and size of radiolysis-induced gas bubbles is of great importance to understand and characterize the thermal and fluid behavior of the solution, especially for the solution based, neutron-induced fission production technique for Mo-99. During the first experiment performed in 2014, ANL’s 35 MeV electron linear accelerator provided average powers of 6, 12, and 15 kW using a rastered beam to homogeneously heat the 15 x 15 x 80-cm uranyl-sulfate solution. Gas bubble size, shape, velocity, solution temperature, and hydrogen and oxygen concentrations were recorded during the irradiations. The details of the experimental setup and results for the first experiment are described in the ANL reports, Design and Construction of Experiment for Direct Electron Irradiation of a Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies and Experimental Results for Direct Electron Irradiation of a Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies. The purpose of this report is to describe the issues and lessons learned associated with the first experiment and elaborate on designs to improve and obtain more accurate experimental results for the upcoming experiment.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Research Needs for Uranium-Zirconium- Based Metallic Fuels

The purpose of this report is to summarize the needs and present recommendations related to the future direction for U-Zr-based metallic fuel research (including binary U-Zr and ternary U-Pu-Zr alloys). These needs and recommendations were determined by subject matter experts from various institutions during a two-day workshop held at the University of Florida in November, 2019. During open-floor discussions, the highest priority gaps in our understanding of U-Zr-based fuels were down-selected, and near- and long-term needs that directly impact the implementation of these metallic fuels were identified. The identified near-term needs include investigation of the following phenomena: i) swelling and fission gas release, ii) fuel-cladding chemical interaction, iii) phase evolution/constituent redistribution, and iv) thermal properties of the fuels. The long-term needs are: i) investigation of fuel creep and plasticity and ii) fission product (lanthanide) transport. In addition, there was general agreement that all institutions and subject matter experts would benefit from an open-source metallic fuels database with thermophysical property and microstructural data, along with fuel operation/irradiation history, which should be regularly updated with vetted information from new experimental and computational investigations and used to advance metallic fuels research and development. Finally, we recommend that metallic fuel research should be ongoing and that fuel qualification and fuel optimization should be equally prioritized; research combining experiments with modeling and simulation has the largest potential impact.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Molten Salt Reactor Engineering Study for Off-Gas Management

In previous work published by the current team, the waste processing and waste form options were summarized for molten salt reactors (MSR) (Riley et al., 2018b; Riley et al., 2019). The primary types of waste from an MSR are summarized in Figure S1 and include (1) off-gas streams, (2) salt waste streams, (3) separated salt streams, (4) metal waste streams, (5) carbon waste streams, (6) decommissioning and decontaminating (D&D) waste streams, and (7) operating waste streams, or those that are generated from maintenance procedures. The primary focus of this report will be on management of the off-gas stream as it represents the pressure boundary for fuel-salt MSRs and, thus, is required for fission product confinement during reactor operation and reduction of the source term during a reactor accident.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

MSR Fuel Salt Qualification Methodology

This report documents the continuing activities at Oak Ridge National Laboratory (ORNL) in support of the US Nuclear Regulatory Commission (NRC) efforts to develop an efficient, appropriate process for liquid fuel salt system qualification. Fuel qualification for salt-cooled molten salt reactors (MSRs) is part of the ongoing coated particle fuel qualification effort. Fuel qualification is a process which provides high confidence that physical and chemical behavior of fuel is sufficiently understood so that it can be adequately modeled for both normal and accident conditions, reflecting the role of the fuel design in the overall safety of the facility. Uncertainties are defined so that calculated fission product releases include the appropriate margins to ensure conservative calculation of radiological dose consequences. The initial phase of ORNL’s activities is documented in ORNL/LTR-2018/1045 Molten Salt Reactor Fuel Qualification Considerations and Challenges.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Automated Internal Energy Calibration by OnTheFly for AGR-5/6/7

The software program, OnTheFly, was developed at Idaho National Laboratory (INL) to keep high-purity germanium (HPGe) detectors energy calibrated during very long experiments. Over time, spectra produced from a HPGe detector will slowly stretch or contract, causing the energy calibration to change. If the energy calibration changes too much, it will cause energy lines to be misidentified. The higher the energy line the more affected they are by this change. OnTheFly tracks and adjusts the slowly changing energy calibrations so the energy lines will not be misidentified. OnTheFly was used with the Fission Product Monitoring System (FPMS) during the AGR-5/6/7 irradiation experiment performed at INL’s Advanced Test Reactor (ATR). Five HPGe detectors were used to monitor the five capsules in AGR-5/6/7. There were also two extra HPGe detectors that were maintained as spares. Before each ATR cycle of the experiment, the HPGe detectors were energy calibrated with a thorium radioactive source. During each ATR cycle, each detector would acquire data for 8 hours and then reset and acquire data for another 8 hours. After each 8-hour run, the spectra would be saved and analyzed autonomously. The results were then read by OnTheFly and used to produce a new energy calibration for each detector. The new energy calibrations were compared to the current energy calibrations and, if certain criteria were met, the new energy calibration would replace the current energy calibration. The new energy calibration would be used for future spectra. OnTheFly was run at least every couple of days but could be run as much as after each 8-hour run. OnTheFly kept the spectra energy calibrated during the ATR cycles over the 2 years and 5 months that the AGR-5/6/7 experiment ran.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Laboratory Investigation of Gas Transport Through Variably Saturated Rock [Slides]

Project Overview: Noble gas fission products, such as xenon (Xe), are monitored to detect nuclear tests. For well-contained subsurface tests, transport to the surface will impact detected isotopic ratios, complicating event identification. Role of water saturation in subsurface transport is not well understood. Historic tests have taken place near or below the water table. To accurately predict breakthrough times and fractionation, gas transport must be characterized through intact rock with variable saturation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Distribution of U, Pu and Np in a Mo-99 Recovery Process using Sachtopore NP (titania based) Absorbent Material

A laboratory scale mockup of a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (U-238), plutonium (94% Pu-239), neptunium (Np-237), and non-radioisotopes of common fission products (including Mo). The goal of this work was to determine the amounts of U, Pu, and Np in the liquid waste streams created by the wash streams for the production process. The key question was how the initial acid wash volume will be handled, as this solution contains residual raffinate which amounted to ~ 3% ofthe mass ofthe feed solution in this experiment. If the initial 1.9 bed volumes (BV) of acid wash were recycled as in this experiment then the combined wash solutions can be expected to contain 0.4 g U/L, 2.7 pg Pu/L and < 14 pg Np/L. If, on the other extreme, the entire acid wash solution was discarded as waste, then the combined wash solutions can be expected to contain 11 g U/L, 3.8 pg Pu/L and ~ 80 pg Np/L. The Sachtopore NP absorbent (for Mo-99 recovery) retains a minuscule fraction ofthe U (2.3 mg U per g titania or 0.03% ofthe U in the feed). The uncertainties are too large to determine how close the absorbent is to saturation with U. Less mass but a higher fraction ofthe total Np was retained by the absorbent (0.7 pg Np per g titania or ~2% of the Np in the feed). As has been reported elsewhere, a large fraction ofthe Pu fed to the titania bed was retained by the absorbent (92% of Pu in the feed or 20 pg per g titania).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Y-89 Cross Sections for MCNP

The ENDF/B-V 89 Y evaluation is found as MAT-9202 on Tape 542, the so-called Fission Product Tape. Evaluation contained on Tape 542 may be complete, but may not contain as much detail as evaluations found on the general-purpose files.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Evaluation of the Generation-2 Dechlorination Apparatus for Treating Salt Wastes

This document satisfies the requirements for two milestones, i.e., M2FT-21PN030103021 and M3FT-21PN030103026, which are described in detail below. Also, details for the work package at the Pacific Northwest National Laboratory (PNNL) are provided below. The goal of this milestone was to demonstrate the generation-2 dechlorination apparatus (G2DA) for removing Cl from chloride-based electrochemical salt simulants. The two salts chosen for this initial study were KCl to simulate a component from the eutectic salt (LiCl is harder to detect with analytical instruments) and CsCl to simulate a fission product chloride that might be present in these types of waste streams. The overall scope of this work included receiving, installing, and testing the G2DA. This new system was designed to address several shortcomings identified with the generation-1 apparatus (G1DA), which are described in detail within this document. The G2DA was designed to address all of these shortcomings. The studies presented in here show that the G2DA worked as intended and will help provide scaled-up batches for additional testing at Argonne National Laboratory.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Status Update on the High Precision Isotopic Measurements on High Burnup LWR Fuel in 2020

The US Department of Energy (DOE) Office of Nuclear Energy (NE) is currently investigating the feasibility of directly disposing dual-purpose (storage and transportation) canisters (DPCs) in a spent nuclear fuel (SNF) repository. Criticality during the repository performance period (10,000 years or more) is one of the major concerns related to direct disposal of DPCs, specifically as the system undergoes degradation in the repository environment and timeframe. Oak Ridge National Laboratory (ORNL) is developing an as-loaded criticality analysis methodology using full (actinides + fission products) burnup credit that exploits the inherent criticality margin associated with actual canister-specific loading configuration. Burnup credit criticality analysis requires validation of the depletion/decay codes used to generate the burned isotopic inventory of an assembly by comparing the code-predicted inventory with experimentally determined isotopic data. Currently, isotopic measurement data for boiling water reactors (BWR) SNF are limited, and additional measurements will be highly beneficial for BWR burnup credit analysis, which is essential to demonstrate disposability of BWR DPCs. Moreover, new pressurized water reactor (PWR) samples of isotopic measurements will expand the PWR sample population and consequently will greatly improve sample statistics. In turn, this will reduce uncertainty in the computational determination of the isotopic composition of commercial SNF and eliminate additional penalties currently used for lack of data. As part of the isotopic analysis task, eight diverse, high burnup (HBU) samples from several PWR rods are being dissolved and isotopically analyzed to provide high-quality measurement data for PWR SNF to reducing the uncertainties associated with PWR isotopic depletion validation. The isotopics of interest focus on nuclides important to burnup credit, shielding, and decay heat in PWR SNF. Additionally, eight high burnup BWR fuel samples from the Limerick nuclear power plant have been identified for measurement to support reducing uncertainties associated with BWR isotopic depletion validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A Review of Nuclear Forensics: 2016-2020

The study of nuclear forensics harkens back to the Manhattan Project-era, when scientists first started to analyze the debris from the 1945 Trinity test. Political turmoil stemming from the Cold War and the rehabilitation of Germany following WWII has led to new challenges in international security involving nuclear proliferation. Nuclear materials have, on occasion, been lost, misplaced, or stolen from former Soviet countries, and illicit materials have been interdicted all over the world. The National Technical Nuclear Forensics Center (NTNFC) was established in 2006, and has been at the forefront of drive to advance nuclear forensic capabilities in the United States. The ultimate goal of nuclear forensics is to examine nuclear and other radioactive materials using analytical techniques to determine origin and history of the material, particularly in the context of law enforcement investigations. Nuclear forensics can be divided into two parts: predetonation and post-detonation. Pre-detonation forensics, as the name implies, is the investigation of a nuclear material or weapon that has not been detonated or involved in an explosion, whereas postdetonation forensics is the study of activation or fission products in debris or the environment following the use of a nuclear or radiological dispersal device (RDD). Both parts require a number of analytical chemical and radiochemical techniques to determine identification of the material. Many advancements in analytical techniques, including rapidity, sample size, and forensic signatures have been made in recent years. The analytical methods that can be used in a nuclear forensic investigation, such as mass spectrometry and gamma spectroscopy, have been described in detail in previous reviews, including Straub et.al, and will not be explained here. This review will discuss recent publications (from 2016 to present) describing advancements of techniques such as radiochronometry, morphology, development of novel reference materials, and inter-laboratory collaborations for both pre- and post-detonation nuclear forensics.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Flowsheet Evaluation of Dissolving Used Nuclear Fuel in PUREX Solvent

This study explores the potential benefit to the used nuclear fuel actinide recovery flowsheet of a process simplification and establish the experimental work needed to develop the simplified flowsheet. The simplification is based on replacing acid dissolution of used nuclear fuel with dissolution in the tri-butyl phosphate solvent used in the industrially mature Plutonium Uranium Reduction Extraction flowsheet. Though characterized by considerable uncertainty, simplified flowsheets appear feasible and potentially offer significant reductions in process complexity, nitrate inventory, secondary liquid effluent generation and plant footprint. Initial development of the technology should focus on key fission product and actinide dissolution as a function of process conditions and understanding the fundamental chemistry of tri-butyl phosphate solvent chemistry. Development of ancillary technologies for minor actinide separations and fuel pretreatment for tritium management is also recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Efficient Failure Probability Calculations and Modeling Interface Debonding in TRISO Particles with BISON

This document constitutes completion of the NEAMS milestone, which is titled Demonstrate efficiency improvements in TRISO failure probability calculations and the effect of debonding on fission product transport in TRISO fuel particles in Bison . In this report we present the development of the following: (1) the anisotropic elasticity model of pyrolytic carbon and extended material models using the local coordinate system for aspherical particle geometry; (2) a capability of modeling interface debonding in TRISO particles; and (3) an efficient high-fidelity approach to calculate the failure probability of TRISO particles. Benchmark problems and AGR-2 tests are used to verify the models and demonstrate these new capabilities.

36 MATERIALS SCIENCE↗