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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Fuel Safety Research Plan for Accident Tolerant Fuels

This document outlines a strategic plan for a safety research program for accident tolerant fuels. Accident tolerant fuels (ATFs) are nuclear fuel materials and designs for existing and future light water reactors. This effort is under the Department of Energy Office of Nuclear Energy’s Nuclear Technology Research and Development Advanced Fuels Campaign. The development of ATF is an industry lead initiative with fuel designs being championed by Westinghouse, Framatome, and General Electric. Industry goals for ATF deployment include implementing the first batch reloads of coated cladding and doped fuel designs into commercial reactors by 2023 and having the first batch reloads of more advanced designs into commercial reactors by 2026. Follow on goals include implementing the first full cores of coated cladding and doped fuel designs into commercial reactors by 2026 and implementing full cores of more advanced designs by 2030. The objectives of the national laboratory research programs within the Advanced Fuels Campaign related to ATF are: 1. Develop and maintain of unique nuclear testing infrastructure, which includes test reactors, hot cell PIE equipment, and unique out of pile separate effects testing facilities. 2. Respond directly to industry requests for support by partnering with the ATF vendors and executing experiments and/or analysis on their behalf. 3. Conduct independent scientific investigations into the performance of the different ATF concepts that are relevant to Industry designs. To achieve these objectives, the safety research program will execute transient irradiation experiments at the TREAT facility at Idaho National Laboratory with the aim of identifying the relevant transient fuel behaviors for each fuel design in order to establish appropriate fuel safety criteria and thresholds. Two test series are described in this plan, (1) an Integral Test Series, and (2) a Focused Effects Test Series. In fulfillment of the first objective for the safety research program, Integral Effects Tests will be developed and demonstrated using standard UO2 Zircaloy fuel designs in both the fresh and previously irradiated case. The Integral Test Series will consist of three campaigns to evaluate transient fuel performance in Reactivity Initiated Accidents, Loss of Coolant Accidents, and in generalized Power Cooling Mismatch scenarios. Development of these test programs also enables the safety research program to fulfill its second objective by making these three test programs available to each industry team. Initially, priority will be given in the integral test series to the coated cladding and doped fuel concepts as their deployment schedule is the most aggressive. To fulfill the third objective of the safety research program, a Focused Effects Test Series is also being developed which will examine specific transient fuel behaviors that are of general interest to all the ATF designs under development. These tests look at transient fuel behavior related to pellet cladding mechanical interaction, coated cladding swelling, rupture, and time at temperature behavior, non oxide fuel melting and mechanical interaction with Silicon Carbide cladding, UO2 and doped UO2 fuel fragmentation and transient fission gas release, and finally fuel coolant interaction. This document provides a general overview of the objectives of the safety research program as well as providing background into how programs like this have been used in the past to develop the criteria necessary for the safe operation of the present fuel designs. An overview of the materials and methods used for safety testing is presented. Finally the Integral and Focused Effects tests series are introduced and described. This document will be maintained as a “living” document with annual updates to reflect evolving requirements and needs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Comprehensive framework for data-driven model form discovery of the closure laws in thermal-hydraulics codes

The two-phase two-fluid model is a basis of many thermal-hydraulics codes used in design, licensing, and safety considerations of nuclear power plants. Thermal-hydraulics codes rely on the closure laws to close the system of conservation equations and describe the interactions between phases. These laws, derived from years of experimental investigations, are semi-empirical correlations that lack generality and have a limited range of applicability. Increase of computational power, availability of new experiments, and development of high-fidelity simulations has increased the number of validation data. The discrepancies between the code predictions and the validation data are a great source of knowledge. Missing physics that are not included in the model but are important for the considered phenomena can be discovered by propagating the information from the experimental results through the model. Furthermore, physics-discovered data-driven model form (P3DM) methodology integrates available integral effect tests and separate effects tests to determine the necessary corrections to the model form of the closure laws. In contrast to existing calibration techniques, the methodology modifies the functional form of the closure laws. Based on the functional form of the correction, the missing physics that were not included in the original model can be discovered. The methodology provides the alternative to the machine learning approach, in which the model is discovered in the form of the intractable black-box relation. In this work, the methodology was applied to the CTF subchannel code to improve the prediction of the two-phase flow phenomena.

42 ENGINEERING↗
DOE OSTI2016

Computer Codes V and V Tests with a Large-Scale Sodium Thermal-Hydraulic Test Facility (STELLA)

The Korea Atomic Energy Research Institute (KAERI) is carrying out the design of a 'Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR)' in cooperation with Argonne National Laboratory (ANL) in the US. In regard to computer codes V and V, a large-scale sodium thermal-hydraulic test program is currently being progressed in parallel with the PGSFR design, which is called STELLA (Sodium Test Loop for Safety Simulation and Assessment). The final objective of this program is to collect a separate and integral effect test database to validate the safety analysis codes and other thermal hydraulic design codes. In the first phase of the STELLA program, the sodium component test loop (hereafter called STELLA-1) was constructed at the KAERI site, and it has been used for performance tests of key sodium heat exchangers in PGSFR. Since this sodium test loop is basically for separate effects tests to verify the performance of key sodium components, the test database obtained from STELLA-1 has been used for the V and V of safety codes as well as heat exchanger design codes. The second phase of the program aiming at an integral effect test is called STELLA-2, which will be used for synthetic review of the key safety issues of PGSFR. The basic and detailed design phases of the STELLA-2 test facility are underway in accordance with the specific design requirements reflecting the whole design features of PGSFR. The facility installation is scheduled to be completed by 2018. Then the main experiments including the start-up tests will begin and the first integral effect test database will be constructed by the end of 2019. Based on the STELLA-2 platform, a simulation of the PGSFR transient will be made to evaluate the plant dynamic behaviors and demonstrate the decay heat removal performance. Multi-dimensional effects coming from a large sodium pool system will be identified as well. The test database obtained from the STELLA-2 experiment will be used for the validation of the safety analysis codes and other thermal hydraulics codes, and this project finally contributes to the specific design phase of PGSFR. This paper presents both the results of heat transfer tests for the key sodium heat exchangers using STELLA-1 and the first results of heat exchanger design codes V and V. The recent progress of the engineering design for the STELLA-2 test section is introduced and its specific design parameters are also presented in this paper with detailed evaluation results of their scaling distortions. A large-scale sodium thermal-hydraulic test program (STELLA) was introduced with the specific activities for separate and integral effects tests. The test database obtained from STELLA-1 was successfully used to validate both the heat exchanger design and safety analysis codes. Specific design features of the integral effects test facility (STELLA- 2) with a scaling ratio of 1/5 in length and 1/125 in volume were also presented. Validation of the STELLA-2 facility design is underway using system codes and CFD tools. Dedicated water-mockup experiments to verify the scaling design of key subsystems are currently being conducted as well. (authors)

42 ENGINEERING↗
DOE OSTI2016

Modelling of LOCA Tests with the BISON Fuel Performance Code

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

BISON code↗
NASA NTRSDate not supplied

A Facility Effect Characterization Test of the BHT6000 Hall Thruster

During the development of the BHT-6000 Hall thruster for flight on NASA’s Power and Propulsion Element, a facility effect test campaign was performed on an engineering unit of the thruster. Testing occurred in Vacuum Facility 5 at NASA Glenn Research Center. The goal of the test campaign was to characterize the behavior of the BHT-6000 Hall thruster as functions of facility effects such that the thruster’s in-flight behavior can be reasonably bounded. In particular, the behavior of the thruster with regard to the background pressure and electrical environment was tested. The background pressure was controlled by injecting supplemental propellant flow into the environment, while the electrical environment was controlled by biasing the facility beam dump. Thrust was found to remain constant over the tested range of pressures while specific impulse was found to increase slightly (1 to 1.5% over a range of up to 20 μTorr, depending on operating point). Plasma plume properties, particularly of ions exiting the side of the thruster, were found to be highly sensitive to pressure and may require pressures of less than a few μTorr for accurate measurements. The thruster was found to operate identically with changing beam dump bias voltage to within measurement uncertainties.

Electric propulsion↗
NASA NTRSDate not supplied

A Facility Effect Characterization Test of the BHT-6000 Hall Thruster

During the development of the BHT-6000 Hall thruster for flight on NASA’s Power and Propulsion Element, a facility effect test campaign was performed on an engineering unit of the thruster. Testing occurred in Vacuum Facility 5 at NASA Glenn Research Center. The goal of the test campaign was to characterize the behavior of the BHT-6000 Hall thruster as functions of facility effects such that the thruster’s in-flight behavior can be reasonably bounded. In particular, the behavior of the thruster with regard to the background pressure and electrical environment was tested. The background pressure was controlled by injecting supplemental propellant flow into the environment, while the electrical environment was controlled by biasing the facility beam dump. Thrust was found to remain constant over the tested range of pressures while specific impulse was found to increase slightly (1 to 1.5% over a range of up to 20 μTorr, depending on operating point). Plasma plume properties, particularly of ions exiting the side of the thruster, were found to be highly sensitive to pressure and may require pressures of less than a few μTorr for accurate measurements. The thruster was found to operate identically with changing beam dump bias voltage to within measurement uncertainties.

Electric propulsion↗
DOE OSTI2017

Counterpart Test on IBLOCA of 13% Cold Leg Break in ATLAS Facility

In the analysis for Loss-of-Coolant Accident (LOCA) as a design-basis accident (DBA) of a pressurized water reactor (PWR), redefinition of the break size based on the risk-informed regulation (RIR) is being extensively investigated due to the potential safety benefits. The pipe integrity and break frequency has been found to depend on the pipe size, so that a complete rupture of a smaller pipe such as a pressurizer surge-line or an emergency core cooling system (ECCS) nozzle is more probable than a double-ended guillotine break (DEGB) of primary coolant circuit piping. The scenario with the break of these pipes can be regarded as intermediate break (IB) LOCA. With an aim of expanding code validation to address safety issues of the IBLOCA and developing an evaluation model (EM) for full spectrum LOCA, the database for the IBLOCA test in an integral effect test (IET) facility is essential. In particular, an experimental data from the counterpart test can contribute to enhance our understanding on the phenomena and resolve the scaling issues in the scaled-down test facilities. The objective of this study is to investigate thermal hydraulic phenomena during the IBLOCA by the counterpart test for the 13% cold leg break in ATLAS facility, which was the A.2 test in the OECDATLAS project. This integral effect test would contribute to evaluate the scaling effect of the ATLAS design and test condition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Subscale maturation of advanced reactor technologies (SMART): A path forward for nuclear thermal propulsion fuel and reactor development

Nuclear Thermal Propulsion (NTP) systems are actively being developed for future crewed missions to Mars. NTP systems excel in missions where both high thrust and high specific impulse are required, but modern NTP systems currently do not have a Technology Readiness Level (TRL) high enough for use in crewed space exploration. TRLs are used to demonstrate the level of rigor with which a component/system has been tested/demonstrated for its intended use. While space systems technology in general must be qualified as a unit, nuclear technology must be first demonstrated to meet qualification level requirements both at the fuel (component) level and the reactor (subsystem) level. Here, in this paper, historic NTP development programs are surveyed to identify a testing and development strategy that can be effectively implemented to allow for NTP reactor development. Based on this strategy, required facilities to enable such activities are identified. Current domestic experimental capabilities to support NTP qualification are limited to separate effects testing of individual components. Separate effects testing is found extensively in historic NTP development efforts but is not sufficient for full fuel and reactor qualification. Combined effects testing allows for an accurate assessment of fuel performance but is not achievable for NTP conditions in existing facilities. Assessment of historic development programs suggests that an intermediate, subscale test facility is necessary to advance NTP TRLs. A solution to meet this need is proposed, namely the Subscale Maturation of Advanced Reactor Technologies (SMART) facility. SMART will mitigate risk to NTP development by enabling performance and reactor physics demonstrations of NTP subsystems. A SMART facility could be built by modifying existing nuclear test facilities, which may potentially enable schedule and cost savings. To pursue reactor qualification beyond the subscale, a new ground test facility will be necessary. This ground test facility should be developed concurrently with SMART to allow for the facility to be operational in time for expedited NTP engine demonstration.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1995

Test-Derived Effective Emittance for Cassini MLI Blankets and Heat Loss Characteristics in the Vicinity of Seams

Thermal performance of Cassini MLI blankets was characterized in thermal vacuum tests. Effective emittance was derived for the standard and high-temperature layups, with or without micrometeoroid protection standoffs, and for a high-performance duo-layup with staggered seams. The all-Kapton layup wa shown to have a 19% higher effective emittance than the hybrid Kapton-Mylar layup, and the duo-layup exhibited a dramatic capability of reducing heat dissipation by 2.6 fold. Several different definitions of effective emittance are discussed, and their relations to distinct hardware-MLI configurations and corresponding modeling techniques clarified. Large temperature differences (up to 70 C) were observed between the center and seam of a blanket. Heat losses in the vicinity of seams are qualified, and a major MLI heat loss mechanism elucidated.

Cassini↗
DOE OSTI2019

Experimental Determination and Modeling of Used Fuel Drying by Vacuum and Gas Circulation for Dry Cask Storage

More than 120 drying tests were conducted using a mock fuel assembly with depleted uranium rods and heater rods to simulate decay heat. These tests followed standard industry practice for vacuum and forced helium drying (FHD). Both single effect tests evaluating a single fuel assembly or cask feature and combined effect tests were conducted. Single effect tests evaluated drying of a specific, known amount of water in a feature. Combined effect tests evaluated specific features and followed the flooding of the chamber, dewatering, and blowdown procedures before beginning the drying procedure. Specific simulated fuel assembly and canister features evaluated included PWR dashpot, BWR water rod, failed fuel rod, spacer disc, and Boral sheet. Ceria pellets were fabricated and used in the failed fuel rod to avoid issues of uranium contamination. A method of monitoring water content in the canister gas and gas stream was developed using optical emission spectroscopy (OES) in addition to monitoring of relative humidity in the chamber and in the vacuum lines.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2020

Medical Proton Test Facilities (MPTFs) Lessons Learned on the Unique Aspects for Single Event Effects (SEE) Testing of Electronics in the 200 MeV Regime

Ever since the closure of the Indiana University Cyclotron Facility (IUCF) in 2014, there has been an increasing use of medical proton therapy facilities (MPTF) for SEE testing with protons in the 200 MeV or greater regime. This talk covers some of the unique features and considerations for utilizing MPTFs including both logistical and technical aspects. This presentation provides an overview of lessons learned for SEE testing at MPTFs.

Medical Proton Test Facilities (MPTFs)↗
DOE OSTI2024

Characteristics of a cost-effective blood test for colorectal cancer screening

Background: Blood-based biomarker tests can potentially change the landscape of colorectal cancer (CRC) screening. We characterize the conditions under which blood test screening would be as effective and cost-effective as annual fecal immunochemical testing or decennial colonoscopy. Methods: We used the 3 Cancer Information and Surveillance Modeling Network–Colon models to compare scenarios of no screening, annual fecal immunochemical testing, decennial colonoscopy, and a blood test meeting Centers for Medicare & Medicaid (CMS) coverage criteria (74% CRC sensitivity and 90% specificity). We varied the sensitivity to detect CRC (74%-92%), advanced adenomas (10%-50%), screening interval (1-3 years), and test cost ($25-$500). Primary outcomes included quality-adjusted life-years (QALY) gained from screening and costs for a US average-risk cohort of individuals aged 45 years. Results: Annual fecal immunochemical testing yielded 125-163 QALY gained per 1000 at a cost of 3811-5384 dollars per person, whereas colonoscopy yielded 132-177 QALY gained at a cost of 5375-7031 dollars per person. A blood test with 92% CRC sensitivity and 50% advanced adenoma sensitivity yielded 117-162 QALY gained if used every 3 years and 133-173 QALY gained if used every year but would not be cost-effective if priced above $$125 per test. If used every 3 years, a $500 blood test only meeting CMS coverage criteria yielded 83-116 QALY gained at a cost of $8559-$9413 per person. Conclusion: Blood tests that only meet CMS coverage requirements should not be recommended to patients who would otherwise undergo screening by colonoscopy or fecal immunochemical testing because of lower benefit. Blood tests need higher advanced adenoma sensitivity (above 40%) and lower costs (below $125) to be cost-effective.

60 APPLIED LIFE SCIENCES↗
NASA NTRS1975

Results of test 0A82 in the NASA/LRC 31 inch CFHT on an 0.010-scale model (32-0) of the space shuttle configuration 3 to determine RCS jet flow field interaction and to investigate RT real gas effects

Tests were conducted in the NASA Langley Research Center 31-inch Continuous Flow Hypersonic Wind Tunnel to determine RCS jet interaction effects on hypersonic aerodynamic characteristics and to investigate RT (gas constant times temperature) scaling effects on the RCS similitude. The model was an 0.010-scale replica of the Space Shuttle Orbiter Configuration 3. Hypersonic aerodynamic data were obtained from tests at Mach 10.3 and dynamic pressures of 200, 150, 125, and 100 psf. The RCS modes of pitch, yaw, and roll at free flight dynamic pressure simulation of 20 psf were investigated.

Thornton, D. E.↗
DOE OSTI2018

NDE Technology Development Program for Non-Visual Volumetric Inspection Technology - Phase I Summary Report

This report provides a summary of the outcomes and findings of the sensor testing and evaluation work performed in FY2017 and first quarter of FY2018 under Phase I of the NDE Technology Development Program. The objective of this summary report is to provide a condensed version of the information provided in the Phase I program plan, Phase I test plan, the two comprehensive Phase I interim test reports issued for Technology Screening and advanced Sensor Effectiveness Testing, and the results from testing and demonstrations conducted Post-Sensor Effectiveness Testing for the purpose of supporting the selection of sensor technologies for maturation under Phase II. The orthogonal research and planning that should accompany technology maturation to support Phase II and Phase III are also included. The full details of the test mock-ups and flaws, sensor technologies, tests, evaluation criteria, and evaluation results are included in Phase I and provided in the interim Phase I test reports on Technology Screening and Sensor Effectiveness Testing.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

Fluid Stratification Separate Effects Analysis, Testing and Benchmarking

High Temperature Gas Reactors (HTGRs) are positioned to disrupt local and global markets via their unique ability to produce carbon-free process heat, high efficiency power generation, and passively safe operational features. However, significant impediments still exist to delay deployment of this particular technology, including a lack of experimental data, verified code application, and lack of consensus with regards to severe accident progression. In particular, air ingress accidents represent a particular challenge to designers and engineers, as they represent low probability, but highly complex, accident scenarios. Including phenomena such as molecular diffusion, free convection, and complex heat and mass transfer paths, experimental and traceable data is essential to maturing the state of the industry. Therefore, this work presents an experimental investigation of the transition to natural convection in HTGR applications using the Stratified Flow Separate Effects Test Facility, housed at Oregon State University. In particular, this work will present data that challenges the assumption that molecular diffusion is a significant factor in this severe accident in the reference facility of the General Atomic 600 MW th Gas Turbine-Modular Helium Reactor (GT-MHR). Rather, experimentally produced data shows a statistically suggestive retardant effect of cross duct orientation, indicating that ingress mechanics may be fundamentally altered via facility geometry. Experimentally, this is achieved using a simplified cross duct that may be positioned in one of two ways so as to provide either horizontal or vertical access to the lower plenum area. Onset of natural convection (ONC) is measured using an oxygen sensor probe, immersed in the helium working fluid, so as to provide direct indication of air presence in the upper plenum. This document will present the doctoral dissertation produced by this work, along with its analysis, experimental data, observations, and conclusions, in addition to select documents that will highlight critical changes during the development process. Those documents will include the quality assurance plan, all subsequent change requests, and the engineering transfer from Harris Thermal in order to establish the most complete quality record possible.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗