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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Evaluation of Wilks’ One-sided Non-parametric Formula Against Analytical Parametric Methods

This paper will be presented to the "Sensitivity and Uncertainty Methods" subgroup of the "Mathematics and Computation Division" of the ANS Winter Symposium. Direct comparisons are made via simulation between the Wilks' non-parametric and Gaussian (parametric) methods which are applied in the SASQUATCH and SINDA-SAMPLE thermal-hydraulic codes, respectively, for the Advanced Test Reactor safety basis. The level of relative conservatism is quantified for the Wilks method and shown to be greater than the parametric method for an equivalent probability and confidence level.

95/95↗
DOE OSTI2016

VCTR: A Versatile Coupled Test Reactor Concept

The United States has currently no domestic fast neutron test and irradiation capabilities necessary to pursue a strong research and development program related to fast neutron reactors. This lack prevents the testing of fast reactor fuels in a prototypic environment, the testing of new safety features and of new coolants. A new irradiation test reactor is a costly investment that will operate for several decades and the initial specifications are crucial to ensure the relevance of such a facility in the long run. This new test reactor should be highly reconfigurable and allow operation as a fast test reactor or a thermal test reactor or a cou-pled fast-thermal test reactor. This report presents the on-going effort at Idaho National Laboratory (INL) to design a versatile coupled test reactor (VCTR—pronounced Victor) that could fulfill the R&D needs requiring high fast neutron fluxes but also high thermal neutron fluxes in configurations that could not be met in current thermal neutron test reactors such as ATR. The objective of this report is to summarize the efforts undertaken at INL from May 2015 to March 2016 aimed at understanding the main options for designing and building a versatile test reactor, with a primary mission to provide thermal and fast neutrons for irradiation of fuels and materials in a manner complementary to the ATR and HFIR reactors. Other missions can also be considered, including the validation of computer codes and the development of specific reactor technologies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

LANL Experience Rolling Zr-Clad LEU-10Mo Foils for AFIP-7

The cleaning, canning, rolling and final trimming of Low Enriched Uranium-10 wt. pct. Molybdenum (LEU-10Mo) foils for ATR (Advanced Test Reactor) fuel plates to be used in the AFIP-7 (ATR Full Size Plate In Center Flux Trap Position) experiments are summarized. Six Zr-clad foils were produced from two LEU-10Mo castings supplied to Los Alamos National Laboratory (LANL) by Y-12 National Security Complex. Details of cleaning and canning procedures are provided. Hot- and cold-rolling results are presented, including rolling schedules, images of foils in-process, metallography and local compositions of regions of interest, and details of final foil dimensions and process yield. This report was compiled from the slides for the presentation of the same name given by Duncan Hammon on May 12, 2011 at the AFIP-7 Lessons Learned meeting in Salt Lake City, UT, with Los Alamos National Laboratory document number LA-UR 11-02898.

Materials Science(36)↗
DOE OSTI2016

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Advanced Test Reactor Overview

Power Point presentation regarding ATR, intended to be presented to university students and/or new INL employees.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Baseline Fracture Toughness and CGR Testing of Alloys X-750 and XM-19 (EPRI Phase I)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials of interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials of interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Baseline Fracture Toughness and SCC Testing of Alloys X750 and XM-19 (Phase I)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) based at INL and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials. The intent of this Phase I research program is to determine baseline properties for the materials of interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.

Jackson, John H.↗
DOE OSTI2020

Comparison of fluence-rate determination techniques for the Advanced Test Reactor-Critical Facility

A technique using small, enriched uranium-aluminum alloy monitors to measure fission rates in fuel plate assembly reactors was developed at the Idaho National Laboratory Radiation Measurements Laboratory in support of Advanced Test Reactor Critical (ATR-C) Facility irradiations. Previous measurements were performed to determine the correlation between the beta-particle activity and the fission rate in the U-Al monitors following irradiation. The American Society for Testing and Materials (ASTM) has established a standard practice E704 for determining neutron fluence rate by radioactivation of depleted uranium by measuring the concentration of fission-products in irradiated uranium monitors. Rather than measuring the beta-particle decay from fission-products, the decay rate of fission-products is determined by gamma-ray spectroscopy. The feasibility of using a method adapted from the ASTM standard practice E704 was investigated for ATR-C irradiations and compared to the present method by analyzing a subset of 21 monitor wires using both methods. The resulting calculated fission rates agreed with an average standard deviation of 7.45%, within the average combined measurement error of 9.05%. Practical considerations such as expected fluence-rates, sample activity, and counting-time were compared between the two methods and the sources of error for each method were investigated. Quality assurance measurements utilizing the modified ASTM standard method are recommended for future test irradiation, but gross-beta-particle counting remains the most practical solution for ATR-C fluence-rate determination.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Progress Report on Disassembly and Post-Irradiation Experiments for UCSB ATR-2 Experiment

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. As stated in previous progress reports, the available embrittlement predictive models, e.g. [1], and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

New Transformer Testing Approach Reduces Reactor Down Time

The ATR-774 transformer yard has been in operation for 57 years. It's old, outdated, and reaching end of life. Plant health has decided to construct a new transformer yard, which will require testing. I've explored the standards library and compiled a list of tests that need to be performed on the transformer yard. This effort led to a significant speed-up in the testing process.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2018

Channel Gap Probe EMPIrE Report

EMPIrE aims to provide data for the "comprehension phase" of the European Low Enriched Uranium (LEU ) U-Mo Fuel Development program for the conversion of European civilian research reactors. The majority of the fuel performance data will be obtained during post-irradiation examinations (PIE) of the fuel. The primary EMPIrE fuel specimen design is based on uranium-molybdenum (U-Mo) alloy particles dispersed in an aluminum matrix clad in the European alloy AG3NE and in the form of mini-plate sized specimens (1" x 4" plates). Secondary EMPIrE specimens include U Mo monolithic fuel clad in European alloy AlFeNi. The experiment is designed in such a way that U-Mo samples achieve three distinct end-of-life burnup groupings. Irradiation has taken place at ATR (Advance Treatment Reactor). The fuel plates are contained in fuel capsules holding four plates each, with 2 plates at 2 levels. Once the capsules with the plates are irradiated, they have to be inspected during the outage in-between cycles. The capsules need to have visual inspection performed and channel gap widths measured prior to reinsertion. These inspections are to ensure that the plates have not deformed significantly. Obstruction of the coolant channels and restriction of the coolant flow are less of a concern in EMPIrE testing as the channels were made intentionally wide. The Channel Gap Probe (CGP) has been developed to measure channel gap widths, which serves as indirect measures of the fuel plate performance and deformation in the course of irradiation which enables the evaluation of the risk from continuing irradiation prior to reinsertion of the plates in reactor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

NSUF Irradiated Materials Library

The Nuclear Science User Facilities has been in the process of establishing an innovative Irradiated Materials Library concept for maximizing the value of previous and on-going materials and nuclear fuels irradiation test campaigns, including utilization of real-world components retrieved from current and decommissioned reactors. When the ATR national scientific user facility was established in 2007 one of the goals of the program was to establish a library of irradiated samples for users to access and conduct research through competitively reviewed proposal process. As part of the initial effort, staff at the user facility identified legacy materials from previous programs that are still being stored in laboratories and hot-cell facilities at the INL. In addition other materials of interest were identified that are being stored outside the INL that the current owners have volunteered to enter into the library. Finally, over the course of the last several years, the ATR NSUF has irradiated more than 3500 specimens as part of NSUF competitively awarded research projects. The Logistics of managing this large inventory of highly radioactive poses unique challenges. This document will describe materials in the library, outline the policy for accessing these materials and put forth a strategy for making new additions to the library as well as establishing guidelines for minimum pedigree needed to be included in the library to limit the amount of material stored indefinitely without identified value.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Measurement of thermal gap conductance using the laser flash method

Thermal gap conductance refers to overall heat conductance through a gas-filled gap as interface thermal resistance between the gas and surface has to be considered. The analysis of new ATR (Advanced Test Reactor) fuel experiments requires experimental data and models for the gap conductance of the cladding/fuel capsule gap, which typically varies from 5 μm to 50 μm and is filled with He at 1-3 atm. Although past research has studied the gap conductance between UO{sub 2} fuel and Zircaloy cladding, new experiments are needed to provide the gap conductance data between HT9 and stainless steel 316, and between advanced fuels and advanced cladding in the Accident Tolerant Fuels program. In this paper, the laser flash method is modified to measure the gap conductance. The gap is formed by welding two stainless steel 316 disks with three spacers. The temperature at the rear surface, after exposure to the laser pulse at the front surface, is modeled with 1D heat transfer through three layers. The analytic solution is then fitted with the measured temperature to determine gap conductance. The sensitivity analysis based on the 1D assumption and 3D numerical simulations were performed to assess the sensitivity and uncertainty of the measurement. It is concluded the heat capacity of the gas in the gap has minimal impact on the measurement, which indicates that the thermal property determined by this method is the thermal conductance, instead of the thermal diffusivity. Convective heat loss through sidewalls under the current experimental design has insignificant influences. Heat transfer through spacers may cause up to 10% overestimation of thermal gap conductance, but the effects can be mostly eliminated by the design of grooves on samples. The result established a pathway for gap conductance determination. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement

The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗