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At least 919 records · Page 51
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Proliferation Resistance and Physical Protection Crosscutting Topics

This report is a companion document to a series of six white papers, prepared jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) and the six System Steering Committees (SSCs) and provisional System Steering Committees (pSSCs). This publication is an update to a similar series published in 2011 presenting crosscutting Proliferation Resistance & Physical Protection (PR&PP) characteristics for the six systems selected by the Generation IV International Forum (GIF) for further research and development, namely: the Lead-cooled Fast Reactor (LFR), the Sodium-cooled fast Reactor (SFR), the Very high temperature reactor (VHTR), the gas-cooled fast reactor (GFR), the Molten salt reactor (MSR) and the Supercritical water–cooled reactor (SCWR).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Safeguards Considerations for Coated Particle Fuel Fabrication Facilities

This study aims to identify where existing fuel fabrication safeguards are limited in application to coated particle fuel fabrication facilities; perform a review of techniques that may be appropriate to address these limitations; and identify where further study is required. The purpose is to ensure the necessary technology and safeguards developments are established to facilitate safeguards-by-design for coated particle fuel fabrication facilities, analogous to the research and development (R&D) efforts in the advanced reactor space, including those requiring coated particle fuels and progress to higher technology readiness levels. Finally, this study will also help meet high priority objectives set forth by the IAEA in establishing safeguards R&D precedence and measurement/procedural standards at emerging fuel fabrication facilities. To achieve these objectives, the report is formatted to (1) distill relevant background information on fuel fabrication of oxide and coated particle fuels; (2) provide an overview of the fuel fabrication techniques for a reference oxide fuel fabrication facility and two selected TRISO fuel fabrication facilities; (3) develop facility models of safeguards approaches for the reference facility and the two TRISO facilities; and (4) identify any observed technology needs or gaps from this analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Cost Benefit Analyses through Integrated Online Monitoring and Diagnostics (Final Report)

The objective of this research is to improve the economic competitiveness of advanced reactors through the optimization of cost and plant performance, which can be achieved by coupling intelligent online monitoring with asset management decision-making. As advanced reactors are early in the development life-cycle, online monitoring systems and associated sensor networks can be incorporated directly into the design without constraints related to retrofitting and system upgrades

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

IER-484 CED4a Report: AFRRI ER1 Dosimetry Characterization

This report presents an analytical function as the reference neutron and photon dose (normalized to an ion chamber integral) for future nuclear accident dosimeter (NAD) exercise at the Armed Forces Radiobiology Research Institute (AFRRI). Various dosimeters and equipment were positioned within exposure room 1 (ER1) and irradiated using a 1.1 MW Training, Research, Isotope, General Atomics (TRIGA) Mark-F nuclear reactor. An ion chamber located in ER1 was used as a normalization factor in Equation 2. This allows for a quick calculation of neutron and photon dose based on the ion chamber data. Currently, AFRRI is in preparation to host a NAD exercise this summer.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Building a better neutrino trap

First detected by Los Alamos researchers Frederick Reines and Clyde Cowan in a nuclear reactor in 1956, tiny particles dubbed neutrinos are so abundant they constantly pass through human bodies by the trillions. But despite decades of study, the neutrino's bizarre nature is still little understood. Lab scientists are trying to rectify that.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

MARVEL Guard Vessel System FEA and ASME Analysis

The INL is seeking to build the first microreactor in the United States, and the first advanced reactor in over 50 years. With the Microreactor Applications Research, Validation & EvaLuation (MARVEL) Project, this microreactor should be operational within a significantly faster timescale than has ever been seen in the nuclear industry. It uses two passive-convection loops, primarily Sodium Potassium eutectic (NaK), and Ga-In-Sn eutectic as an intermediate fluid. Due to the high reactivity of NaK, the primary coolant loop requires a secondary confinement boundary to prevent uncovering of the core in the event of a NaK leak. This secondary boundary is mostly accomplished through the addition of the Guard Vessel, which is a bucket-like metal containment, which is welded onto the primary boundary to maintain an inert gas environment and pressure

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Residual Stress Measurements of Cr-coated Cladding

Accident-tolerant fuel concepts for light-water reactor applications have been developed and tested in diverse research programs around the world. Industry teams have developed coated cladding concepts using different coating application methods, such as physical vapor deposition and coldspraying. Recently, variations in thickness and application method were investigated, and their effects on the microstructure and mechanical properties were reported. The results suggest an increase in performance in an unirradiated state, however, it remains unclear what phenomenon drove this performance. One explanation to this performance is associated with the coating process imparting a residual stress. Residual stress analysis was performed on coatings produced with various process parameters to understand how a thin layer of Cr impacts the mechanical properties after coating application and after simulated operating conditions. The gathered information is intended to address safety concerns or support margin identification related to accident-tolerant fuels and to support subsequent modeling and simulation efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Nuclear Technology Selection and Preliminary Integration at St. Charles and Carrollton Chemical Plants

Nuclear energy is identified as a valuable candidate for decarbonizing industrial facilities as a non-emitting baseload energy source capable of dispatching energy as needed for an integrated facility. As industrial facilities are unique, analyzing what nuclear technology is the best candidate should be evaluated on a case-by-case basis. For two identified facilities of the Dow Chemical Company, an ideal mix of nuclear energy resources is calculated, and integration strategies are put forward. This research was funded through a Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher, specifically seeking near-term engineering solutions for the technical issue of introducing nuclear generated energy into the Dow chemical facilities. The research is split into four main phases: background on reactors and chemical facilities, preliminary engineering analysis determining how reactor technologies may be integrated with specific facilities, site characterization and discrete hazards analysis, and finally an optimization of reactor-facility selection through the use of Framework for Optimization of ResourCes and Economics (FORCE) tools.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Issue Summary of INL Phase IV Transient Results for IAEA CRP on HTGR UAM Benchmark

This report details the Parallel and Highly Innovative Simulation for Idaho National Laboratory (INL) Code System (PHISICS)/Reactor Excursions and Leak Analysis Program (RELAP5)-3D results obtained for the transient core exercises defined for Phase IV of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on high-temperature gas cooled reactor (HTGR) uncertainty analysis in modeling (UAM). The Phase III models and results are linked to the earlier Standardized Computer Analyses for Licensing Evaluation (SCALE)/Sampler/New ESC-based Weighting Transport (NEWT) data generated for the lattice physics (lattice) stage Phase I of the CRP. The focus of this report is the Uncertainty/Sensitivity Assessment (U/SA) of the prismatic modular high-temperature gas cooled reactor (MHTGR)-350 design, and specifically for Exercises IV-1 and IV-2 of the benchmark: the Control Rod Withdrawal (CRW) and Pressurised Loss of Cooling (PLOFC) events. The statistical U/SA methodology is implemented and demonstrated using the RAVEN code, based on perturbed cross-section libraries obtained from the SCALE/Sampler sequence. Uncertainties in nuclear data (cross-sections and the average number of neutrons produced per fission, 235U[¯v ]) lead to standard deviations (uncertainties of one s) of approximately 0.5% in the core eigenvalues of the MHTGR-350 and core models. For the coupled neutronics/thermal fluid model, local power density uncertainties up to 3.6% were observed in the colder regions of the core, while the local maximum fuel temperature uncertainties reached 1.5% for the models that included thermal fluid uncertainties. The addition of thermal fluid uncertainties dominated the impacts of nuclear data uncertainties in all cases. The main contributors to uncertainties in the power density and fuel temperatures during the transients were uncertainties in the reactor operating conditions (total power, inlet mass flow rate and inlet gas temperature). Variations in the bypass flows did not have significant impact on any of the output variables. For the nuclear data uncertainties it was found that the 235U(¯v ) / 235U(¯v ) covariance produced the largest sensitivities in terms of its impact on the eigenvalue and peak reactor power. It was also observed that the impact of any nuclear data uncertainties on the maximum fuel temperature was much less significant that the impact on eigenvalue and power. Another important finding was that although the use of eight or more energy groups is recommended for best-estimate HTGR simulation, two-group models produced acceptable uncertainty and sensitivity results for most FOMs. Since the statistical U/SA methodology is computationally expensive, and most transient solver requirements will scale directly with the number of energy groups, two energy groups could be used by HTGR developers during the early stages of design when larger uncertainty margins can be tolerated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

MFC and Fuel Evaluation for KSU

Idaho National Laboratory's Materials and Fuels Complex is well equipped for research, development, and demonstration of nuclear fuel designs and reactor components. The capabilities for fuels RD&D at MFC are briefly presented and the observation of fuel behavior that merit RD&D at MFC are introduced, with emphasis on metal fast reactor fuel.

Materials and Fuels Complex↗
DOE OSTI2023

SPC-70783 Rev 0 Seismic Detector and Switch Survey and Verification Specification

A. Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core and Stirling engines under the Microreactor Application Research Validation and Evaluation (MARVEL) program. The MARVEL microreactor Reactor Protection System (RPS) will have a seismic trip function. The selected part for this design is ETNA2 manufactured by Kinemetrics, Inc. B. Acceptance of the items delivered under this scope of work can be performed through a direct evaluation of the supplier’s controls and a factory acceptance test.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2021

Nuclear Material Control and Accountancy Approach for Pebble Fueled Reactors using a Novel Pebble-Type Identification and Classification Technology

In FY21, Argonne National Laboratory (ANL) with researchers at Texas A&M University (TAMU) designed and engineered a prototypical device for accounting types of irradiated pebbles (for example with different 235U enrichments or pure graphite) in Pebble-Fueled Reactors (PFRs). Through engagements with reactor designers, a need arose to assist in identifying and categorizing types of pebbles as a complementary nuclear material control method that would synergize with designers’ use of fuel burnup measurements for material accountancy needs. As part of an overall nuclear material control approach, a concept of pebble batch accounting was investigated using extrinsic non-radiological features to identify intrinsic characteristics. This concept of batch accounting led to the ability of identifying types of pebbles based on characteristics such as initial 235U enrichment of pebble batches or dates of introduction into the reactor core. Identification was achieved by embedding the 5-mm thick graphite periphery of pebbles with 2-mm diameter inert Yttria-Stabilized Zirconia (YSZ) microspheres to achieve an averaged volumetric density (i.e., common spacing between microspheres) unique to that type of pebble. With an ultrasound imaging system in proximity with each pebble, the YSZ microspheres in the pebble proved visible and their spacing became the unique feature upon which pebble type categorization could occur. At the culmination of FY21, ANL intended on delivering and installing the prototype at TAMU for initial testing but, due to the on-going pandemic, this was postponed until FY22.

Gariazzo, Claudio↗
DOE OSTI2024

Post-irradiation Examination of Eurofer-97 Steel Irradiated to 20 dpa at 200–400°C in HFIR under the EUROfusion (ORNL-KIT) Collaboration Program

The Oak Ridge National Laboratory/Karlsruhe Institute of Technology (ORNL/KIT) collaboration focuses on research involving irradiation experiments and post-irradiation examinations (PIE) of isotopically modified Eurofer-97 steels for fusion reactor applications. This collaboration capitalizes on ORNL’s expertise in radiation effects on materials and its capabilities in irradiation and PIE. The research aims to qualify Fe-9Cr-based Eurofer-97 steel under simulated fusion conditions, which include high doses and high transmutation rates. A unique isotopic modification technique is utilized in the research, wherein high-transmutation isotopes such as Fe-54 and Ni-58 are alloyed into the Eurofer steel to align with the expected helium production rates in fusion reactor conditions. This report presents the results of post-irradiation mechanical testing activities for the EUROFER-97/2 specimens after irradiation to ~20 dpa at various irradiation temperatures. The irradiation doses and temperatures for the ES capsules ranged from 18.4–20.4 dpa and 202–256°C, respectively. The mechanical property datasets, obtained through baseline testing and PIE of the tensile and fracture specimens, include microhardness data from 93 irradiated and non-irradiated tensile specimens and 28 irradiated bend bar specimens, uniaxial tensile property data from 45 irradiated and non-irradiated tensile specimens, and fracture toughness data from the bend bar specimens. Further data analyses provide statistical information on the microhardness values and tensile properties, as well as the reference ductile-brittle transition temperature (T 0 ) data.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Internship Presentation: Integrating Safety and Cybersecurity: Security-by-Design with SOWT Analysis for Reactor Testing

This study covers leveraging reactor testing facilities that are primarily designed with a focus on safety to enhance cybersecurity testing. By incorporating reactor security-by-design with reactor safety-by-design principles and adopting defense-in-depth strategies that emphasize both safety and security, the research evaluates applicable cyber tools, models, and solutions. This includes simulating specific cyber-attack scenarios using reactor simulators and performing SWOT (Strengths, Weaknesses, Opportunities, Threats) analysis to improve the cybersecurity of reactor systems.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

Preparations for a European R&D roadmap for an inertial fusion demo reactor

A European consortium of 15 laboratories across nine nations have worked together under the EUROFusion Enabling Research grants for the past decade with three principle objectives. These are: (a) investigating obstacles to ignition on megaJoule-class laser facilities; (b) investigating novel alternative approaches to ignition, including basic studies for fast ignition (both electron and ion-driven), auxiliary heating, shock ignition, etc.; and (c) developing technologies that will be required in the future for a fusion reactor. This paperr presents a brief overview of these activities, along with new calculations relates the concept of auxiliary heating of inertial fusion targets, and provides possible future directions of research and development for the updated European Roadmap that is due at the end of 2020. This article is part of a discussion meeting issue ‘Prospects for high gain inertial fusion energy (part 2)’.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Characterization of C-Reactor (105-C) Moderator Tanks 204 and 205

Area Completion Project (ACP)/Savannah River Nuclear Solutions (SRNS) has requested Savannah River National Laboratory (SRNL) to perform chemical and radiological analyses on moderator heavy water samples from C-Reactor (105-C) in Moderator Tanks 204 and 205. The heavy water samples have been characterized using SRNL analytical methods. Information generated from the characterization of these heavy water samples provide a basis for determining the residual contamination remaining in the tanks. These analyses are needed in order to update the C-Reactor Contaminant migration groundwater model and report, to help determine a disposition pathway for the moderator water, and to ship samples off-site to the Southwest Research Institute (SWRI). This report presents characterization results for the March 2024 C-Reactor Moderator Tank 204 and 205 heavy water samples. SRNL results are critical to allow ACP to complete this Performance Based Incentive project on time. Based on the characterization data provided in this report, results of these samples meet customer expectations and satisfy the Technical Assistance Request (TAR).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Geo- and reactor antineutrino sensitivity at THEIA

We present the sensitivity of the THEIA experiment to low-energy geo- and reactor antineutrinos. For this study, we consider one of the possible proposed designs, a 17.8-ktonne fiducial volume THEIA-25 detector filled with water-based liquid scintillator placed at Sanford Underground Research Facility (SURF). We demonstrate THEIA’s sensitivity to measure the geo- and reactor antineutrinos via inverse-beta decay interactions after one year of data taking with $11.9 × 10^{32}$ free target protons. Considering all uncertainties on input throughout the whole analysis chain, the expected number of geo- and reactor antineutrinos is $220^{+30}_{–24}$ (stats+syst) and $168^{+26}_{–24}$ (stat+sys), respectively, after one year of data taking. The corresponding expected fit precision of a sole experiment is evaluated at 8.7% and 10.1%, respectively. We also demonstrate the sensitivity towards fitting individual Th and U contributions, with best fit values of $N_{\text{Th}} = 40^{+26}_{–22}$ (stat+sys) and $N_\text{U}=180^{+30}_{–24}$ (stat+sys). Finally, from the fit results of individual Th and U contributions, we evaluate the mantle signal to be $S_{\text{mantle}} = 9.3±[5.2,5.4]$ NIU (stat+sys). This was obtained assuming a full-range positive correlation $(ρ_c ϵ[0,1])$ between Th and U, and the projected uncertainties on the crust contributions of 8.3% (Th) and 7.0% (U).

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗