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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

BISON As-run AGR-3/4 Irradiation Test Predictions

BISON, a nuclear fuel performance application built using the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element library, was used to model the Advanced Gas Reactor (AGR)-3/4 irradiation test using as-run physics and thermal hydraulics data. The AGR-3/4 test consists of the combined third and fourth planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-3/4 test train consists of twelve separate and independently controlled and monitored capsules. Each capsule contains four compacts filled with both uranium oxycarbide (UCO) unaltered “driver” fuel particles and UCO designed-to-fail (DTF) fuel particles. The DTF fraction was specified to be 1×10-2. This report documents the calculations performed to predict the failure probability of tristructural isotropic (TRISO) coated driver fuel particles during the AGR-3/4 experiment on a single compact. This report will demonstrate the capabilities of BISON to model the complex AGR-3/4 irradiation test and identify further development needed to capture the fuel particle failure probability and source term on every compact for further comparison from post-irradiation examination (PIE) data. The calculations include the modeling of the AGR-3/4 irradiation that occurred from December 2011 to April 2014 in the Advanced Test Reactor (ATR) over a total of 10 ATR cycles including seven normal cycles, one low power cycle, one unplanned outage cycle, and one Power Axial Locator Mechanism (PALM) cycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Instrumentation Plan for the OSU High Temperature Test Facility, Revision 4

This instrumentation plan describes the instrumentation requirements and design for the Oregon State University (OSU) High Temperature Test Facility (HTTF). This report outlines HTTF instrumentation design specifications, data acquisition and control system (DACS) design specifications as well as a brief description of the instrumentation to be used in the HTTF. The HTTF is an integral test facility, which has been designed to model the behavior of interest for a Very High Temperature Gas Reactor (VHTR) during the Depressurized Conduction Cooldown (DCC) event. It also has the potential to conduct limited explorations into the progression of the Pressurized Conduction Cooldown (PCC) event and phenomena during normal operations. The facility is scaled 1:4 by height and 1:64 by volume. The maximum core power for the facility is approximately 2.2MW. The core region is of a modular design to allow for both the prismatic block core and the pebble bed core designs to be tested. However, only the prismatic block design is discussed in this report. A correctly scaled integral facility modeling the VHTR that can be used to examine the thermal hydraulic system behavior of the design basis and beyond design basis transients is of significant value. The experimental results obtained in such a program could be used to validate analytical tools and methods being proposed for use with the VHTR. Therefore, it is of importance that the HTTF be instrumented in such a way that the required data is collected to meet its overall goals of phenomena examination and code validation. In addition to the instrumentation used to collect data to achieve the overall goals of the HTTF test program, there is instrumentation in the HTTF, which is primarily used to ensure the safe operation of the test facility. This type of instrumentation shall be termed “Control Instrumentation” and is discussed in this report. This report discusses the DACS, which is a personal computer controlled system that is used to control both the collection and storage of the test matrix and control system data as well as electrical and mechanical operation of the test facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Irradiated AGR 2 Compact 4-3-1 Examination Plan

This plan describes post-irradiation examination activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory on irradiated Compact 4-3-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the Statement of Work for PIE and Safety Testing of Pebble Bed Modular Reactor Limited Fuel at ORNL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Initial Assessment of X-Ray Computer Tomography image analysis for material defect microstructure

The original development work leading to this report was focused on the non destructive three-dimensional (3-D) characterization of nuclear graphite as a means to better understand the nature of the inherent pore structure. The pore structure of graphite and its evolution under various environmental factors such as irradiation, mechanical stress, and oxidation plays an important role in their observed properties and characteristics. If we are to transition from an empirical understanding of graphite behavior to a truly predictive mechanistic understanding the pore structure must be well characterized and understood. As the pore structure within nuclear graphite is highly interconnected and truly 3-D in nature, 3-D characterization techniques are critical. While 3-D characterization has been an excellent tool for graphite pore characterization, it is applicable to a broad number of materials systems over many length scales. Given the wide range of applications and the highly quantitative nature of the tool, it is quite surprising to discover how few materials researchers understand and how valuable of a tool 3-D image processing and analysis can be. Ultimately, this report is intended to encourage broader use of 3 D image processing and analysis in materials science and engineering applications, more specifically nuclear-related materials applications, by providing interested readers with enough familiarity to explore its vast potential in identifying microstructure changes. To encourage this broader use, the report is divided into two main sections. Section 2 provides an overview of some of the key principals and concepts needed to extract a wide variety of quantitative metrics from a 3-D representation of a material microstructure. The discussion includes a brief overview of segmentation methods, connective components, morphological operations, distance transforms, and skeletonization. Section 3 focuses on the application of concepts from Section 2 to relevant materials at Idaho National Laboratory. In this section, image analysis examples featuring nuclear graphite will be discussed in detail. Additionally, example analyses from Transient Reactor Test Facility low-enriched uranium conversion, Advanced Gas Reactor like compacts, and tristructural isotopic particles are shown to give a broader perspective of the applicability to relevant materials of interest.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Investigation of Process Emissions during Thermal Cutting of Clad Metals from the Nuclear Sector - 20183

Typically, reactor pressure vessels as well as steam generators comprising low alloyed steel with a cladding of stainless steel. Recently the permissible values for Cr (VI) in the air of the working areas have been reduced in Germany from 100 μg/m{sup 3} to 1 μg/m{sup 3} air. This results in the danger that even with low-alloy steel with claddings these limit values would be exceeded when using flame cutting technologies. In order to ensure compliance with the permissible limit values, exhaust gas systems with gas cleaning measures are provided. The process emission test shall replace the assumptions required for the design of these systems with measured values. NUKEM Technologies Engineering Services GmbH (NUKEM) has chosen autogenous flame cutting as an example of a thermal cutting system for process emission testing, as there are no proven values in the literature for the Cr (VI) and NiOx emissions produced. The autogenous flame cutting can be used for a variety of cutting tasks within decommissioning projects. In order to simulate the material to be cut, mock ups were generated and cut from information about the material specification for the reactor pressure vessel from existing nuclear power plants including possible claddings. The cutting length needed for analysis in the test will be maximum 100 mm. The airborne emissions from the thermal cutting process will be captured and analyzed as follows: - Dust emission: The generated particles will be collected by a hood equipped with strong suction device and suitable filter elements. The resulting filter is gravimetrically measured before and after the tests. - Particle Size Distribution: The airborne particle size distribution measurement will result in figures for the size and number of particles arising from the thermal cutting process. - Gas Analysis: The gas analysis enables the determination of e.g. CO{sub 2}, CO, O{sub 3}, NO, NO{sub 2}. - Cr (VI) and NiOx Analysis: For the Cr (VI) and NiOx analysis, samples of the filter residue will be taken. As a result Cr (VI) in relation to total Cr as well as the Ni content will be determined by ICP-optically emission spectroscopy and put in relation to the total dust amount. The process emissions test itself as well as the necessary gas analyses and spectroscopic analyses were carried out at the Institute of Materials Science (IW) of Leibniz Universitaet Hannover (LUH). The results of the process emission tests are used to confirm the design of the gas cleaning systems. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program (Rev. 2)

The beyond design basis accident at Fukushima Daiichi in March 2011 and the Three Mile Island accident in 1979 imply that severe nuclear accidents may occur at somewhat higher frequencies than previously predicted, and that the financial liabilities of such accidents can cripple a utility. While current fuel is more than adequate for design basis accidents, the potential for harm from beyond design basis accidents (BDBA) has been a driver for Westinghouse development efforts since 2003. After Fukushima, other fuel vendors also began pursuing new fuel materials that can provide significant increases in the time, enabling reactor operators to respond to unforeseen events before significant releases of fuel materials and fission products occur. In order to be commercially successful, accident tolerant fuel (ATF) products must also provide significant fuel cost savings, operating cost improvements as well as safety improvements. This report documents the engineering, design and testing that occurred from October 1, 2016, through September 30, 2018, by the Westinghouse team. In addition, it includes: 1. The results of modeling and simulation work completed to date that supports technical feasibility of the ATF concept in meeting the goals of the enhanced ATF program including a core level analysis predicting how an existing LWR will perform when loaded with ATF. 2. Regulatory Analysis showing path to LTA/LTR acceptance. 3. Business Analysis showing commercial viability of the ATF concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Springborn, R. H.↗
DOE OSTI2018

The GIF Proliferation Resistance and Physical Protection Methodology Applied to Gen IV System Designs: Some Reflections

The Generation Four International Forum (GIF) formed a proliferation resistance and physical protection (PR&PP) Working Group (PRPPWG) to develop a methodology to evaluate the six GIF reactor technologies. The PRPPWG developed the methodology through a series of development and demonstration case studies, using a hypothetical “Example Sodium Fast Reactor” (ESFR). The ESFR assessment enabled testing of the full methodology on a complete nuclear energy system and many insights were gained from the process. The PRPPWG, in collaboration with representatives of the GIF System Steering Committees/provisional Steering Committees (SSCs/pSSCs) for each of the six GIF reactor technologies, analyzed the PR&PP aspects of the six reactor technologies. The outcome of this activity is documented in a joint PRPPWG and SSCs/pSSCs report made publicly available in 2011, highlighting the analysis performed and its main outcomes. Currently, the PRPPWG and the six GIF SSCs/pSSCs are in the process of repeating the exercise, based on the current status of the six GIF system design concepts, taking into account the designs’ evolution since 2011. In line with Safeguards by Design concept, whilst during the first exercise the level of detail of the design options enabled only a high-level evaluation of the systems’ proliferation resistance, the early design stage allowed designers to introduce modifications for improving the systems’ performance. With the designs’ current level of detail, the opportunity for proposing modifications diminishes, whilst at the same time increasing the opportunity to inform the safeguards R&D community on potential R&D needs to meet safeguardability issues. The paper will first summarize the main concepts behind the PR&PP evaluation methodology. It will then summarize the main outcomes and lessons learned, as in the current white papers on the PR aspects of the six GIF design concepts, and will finally describe the ongoing activity to update them.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Non-destructive Examinations of ATF-2 Baseline Rodlets

This report contains the results of non-destructive examinations of Accident Tolerant Fuels 2 (ATF-2) rodlets irradiated in the Advanced Test Reactor (ATR) Loop 2A. The experiment is part of the U.S. Department of Energy Nuclear Technology Research and Development (NTRD) program’s Advanced Fuels Campaign (AFC). The rodlets were composed of UO2 pellets and Zr-4 cladding. The data have been collected to provide baseline PIE data to support future testing in TREAT and to have performance data against which the data of the ATF concepts can be compared directly. The analyses performed included: visual examinations, axial gamma scanning, neutron radiography and profilometry. All the data collected showed a performance consistent with expectations for this fuel system at low burnup.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Improving to the neutron fluence rate monitor measurement system at the Advanced Test Reactor [Poster]

The existing fluence monitor wire scanning system at the Advanced Test Reactor (ATR) was designed and installed for use in the Engineering Test Reactor (ETR) when it began operation in 1958. The wire scanner was operated in ETR for over 20 years until ATR began operation, when it was moved to the ATR west canal area in 1971 and subsequently moved to the west canal in 2006 where it presently resides. With a continued service life of 65 years the system is well beyond the typical design life of 20 years for these types of systems. The need to update the data acquisition and control system was identified, and the benefits of replacing the existing sodium iodide (NaI) detector with an electronically cooled high-purity germanium (HPGe) detector are discussed. The wirescanner system in the ATR canal is utilized after every reactor cycle by the ATR Radiation Measurements Laboratory (RML) to assess the activation of cobalt and nickel dosimeter wires during the cycle. These wires become activated through exposure to thermal and fast neutrons respectively during the irradiation cycle and are highly radioactive upon shutdown. It is for this reason that the wirescanner is used in the ATR canal rather than transporting the dosimeters to another facility. A scoping study was performed to develop a base-line design to ensure that existing capabilities could be replaced with a new system. The new hardware will enable automated measuring of several flux monitor holders without necessitating the removal of the flux wires. In this way, flux wire measurements will be performed with minimal dose to the technicians and will not be limited by canal operations as is presently the case. The new control and acquisition software will be based on commercially available and supported systems that have a wide user-base to provide long-term stability. An electronically cooled HPGe detector will be used to provide high-resolution gamma-ray measurements, an improvement from the low-resolution sodium-iodide detector that is presently deployed. The electronic cooler eliminates the need for liquid nitrogen to cool the detector head. A new collimator has been designed to house the new detector and allow for sufficient counting rates.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Incorporating A Risk-Informed, Performance-Based Concept into Nuclear Fuel and Materials Development for Advanced Reactors

The NRC’s regulatory concept for advanced reactors under 10 CFR Part 53 promotes the use of an RIPB concept. RIPB is usually referenced as a general, overarching concept. This paper is focused on how to, and if it would be worthy to, incorporate an RIPB concept into nuclear fuel and materials development for advanced reactors. This paper proposes three potential RIPB applications for nuclear fuel and materials development, including RIPB test matrix development, RIPB test matrix reduction, and RIPB design optimization. This paper discusses the projected benefits from incorporating RIPB in reducing development timeline, cost, and regulatory risk.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

CBP for Field Workers – Results and Insights from Three Usability and Interface Design Evaluations

Nearly all activities that involve human interaction with the systems in a nuclear power plant are guided by procedures. Even though the paper-based procedures (PBPs) currently used by industry have a demonstrated history of ensuring safety, improving procedure use could yield significant savings in increased efficiency as well as improved nuclear safety through human performance gains. The nuclear industry is constantly trying to find ways to decrease the human error rate, especially the human errors associated with procedure use. As a step toward the goal of improving procedure use and adherence, researchers in the Light-Water Reactor Sustainability (LWRS) Program, together with the nuclear industry, have been investigating the possibility and feasibility of replacing the current paper-based procedure process with a computer-based procedure (CBP) system. This report describes a field evaluation of new design concepts of a prototype computer-based procedure system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program

Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program The Transient Reactor Test (TREAT) Facility is a 19,000 MW reactor located at the Idaho National Laboratory designed to perform transient testing on nuclear fuel experiments and materials. A neutron radiography facility was installed with the first radiograph being produced in 1967 that utilized the indirect (or transfer) method of neutron radiography to examine irradiated nuclear fuels and experiment test loops. The TREAT neutron radiography facility performed approximately 5,000 radiographs by the spring of 1977. The original radiography facility was upgraded in 1974-1975 to correct previous deficiencies and to improve capabilities. In 1986, an upgraded aperture was installed to increase the resolution of images followed by a major reactor upgrade completed in 1988. Overall, the existing neutron radiography system was in good working order. Originally built in 1958, the TREAT Facility was in operation until it was placed in a shutdown status in 1994. Following the Fukishima disaster and seeing a need for enhanced accident tolerant fuels, the United Stated Department of Energy decided to restart the TREAT facility and resume transient operations. In November 2017, the TREAT reactor was successfully restarted and is currently performing operational testing in preparation for initial experiment irradiations and transient testing. This paper reports on efforts to reactivate the TREAT neutron radiography station. A few minor shielding and system improvement upgrades have been completed along with the installation of a dedicated room for computed radiography equipment. Detailed beam characterization work was not performed following the last reactor upgrade. To characterize the neutron beam, gold foil activation measurements were made to determine an average neutron flux and flux profile. Additional multi-foil activation measurements will be made to calculate the energy spectrum. An open beam image will be taken that will provide the information about variations in the beam profile. A series of system qualification shots will also take place to accurately determine the effective image quality and image acquisition parameters. Full operational status of the TREAT neutron radiography system is scheduled for late summer of 2018.

36 MATERIALS SCIENCE↗
DOE OSTI2016

The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Impact of MURR LEU Conversion on Beryllium Reflector Lifetime

The U.S. National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M 3 ) has developed and is pursuing an integrated approach to address the persistent threat posed by unintentional proliferation of nuclear materials. The NNSA M 3 approach reduces the risk of highly enriched uranium (HEU) and plutonium falling into the hands of non-state actors by minimizing the use of and, when possible, eliminating weapons-usable nuclear material around the world. In this geopolitical context, most research and test reactors, both domestic and international, have completed or started a program of conversion from the use of HEU to low-enriched uranium (LEU) as fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Advanced Test Reactor Core Modeling Update Project: Annual Report for Fiscal Year 2014

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate verification and validation (V&V) protocols according to applicable national standards. The project has now completed its fifth, and last, full year of development activity, and has entered an accelerated deployment phase that is expected to run through 2015.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

GSI 191 Closure Issue for EDF PWR Fleet: The Fuel Assembly Test Results

EDF has re started recirculation studies within the international GSI 191 framework by 2002, when EDF decided to join the NEI sump strainers task force. All the EDF PWR fleet have been upgraded, with major strainer modifications, according to the NEI 04-07 guideline. The so-called chemical effects have been carefully studied by EDF and demonstration have been released that there are no chemical effects on EDF fleet, owing to the nature of our debris. The study of strainer downstream effects started by 2009, with decoupling Downstream Sump Debris Source Term (DST) and by the end of 2016, all the equipment have been successfully qualified. At a previous ANS meeting,it was said that core fuel tests were scheduled. They have started at the beginning of 2014 and they have been closed by mid 2016 for AREVA Fuel type at ERLANGEN facility (Germany). The present presentation is aimed at describing the test facility, the test specifications and the main results. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗