The nuclear rocket program - Its status and plans.
Nuclear rocket program status and plans, presenting NERVA reactor and engine system tests
Corrington, L. C.↗
SEARCH · Engineering Papers
Results for “Nuclear Reactor Testing”
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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Static tests of 4-inch outside-diameter frangible tube energy absorbers for nuclear aircraft
Static tests of aluminum and steel frangible tube energy absorbers for nuclear aircraft reactors
Gumto, K.↗
Gas-core rocket reactors - A new look
Open cycle gas core nuclear rocket engine, considering reactor critical experiments and cold and hot flow tests
Ragsdale, R. G.↗
Investigation of an r-f radiant energy source with uranium hexafluoride and tungsten hexafluoride injection.
An experimental investigation was conducted using the UARL 1.2-megw r-f induction heater to simulate thermal environment and fuel region characteristics expected in nuclear light bulb reactors. Argon was injected at the ends of the test chamber to provide radial-inflow vortex flow pattern. In some tests, simulated fuel consisting of a mixture of argon gas with either tungsten hexafluoride gas, uranium hexafluoride gas, or tungsten particles was injected into the discharge region from probes located at the centers of the end walls. The encouraging initial results obtained are discussed.
Roman, W. C.↗
Evolution of Microstructure and Surface Characteristics of FeCrAl alloys when Subjected to Flow Boiling Testing
FeCrAl alloys are candidate materials for manufacturing accident-tolerant fuel (ATF) cladding intended for light water reactors to increase fuel reliability and safety during design-basis and beyond-design-basis accident scenarios. The evolution of the materials' surface characteristics, microstructure, and mechanical properties when exposed to the Critical Heat Flux (CHF) in flow boiling testing is crucial for safety analysis while providing insights into their thermal-hydraulic performance in nuclear reactors. After CHF, the surface chemistry of two FeCrAl alloys, APMT and C26M, was studied to understand their evolution at the early stage of high-temperature excursions in short time periods. The results indicated a thin layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the layer, as indicated by X-ray photoelectron spectroscopy (XPS) and depth profiling. The cross-sections prepared by focused ion beam (FIB) revealed the growth of an oxide layer, in the range of 90-180 nm thick, on the alloys' surfaces. The evolution of the materials' surface chemistry also led to a noticeable post CHF excursion increase in their wettability, with a slight increase in roughness. Further, the investigation of the materials' mechanical properties indicated a modest increase in hardness by 10-15% as well as an increase in their yield strength, as evidenced by the microindentation and ring compression tests conducted before and after CHF testing. Scanning electron Microscopy (SEM) and X-ray diffraction (XRD) were used to investigate microstructural features of the materials and their changes after CHF treatment.
36 MATERIALS SCIENCE↗
Ship Motion Testing Needs Evaluation
The American Bureau of Shipping performed an evaluation of national maritime nuclear testing capabilities and determined there currently is an inability to test maritime nuclear power plants in a ship motion environment. A Ship Motion Test Facility (SMTF) is proposed to fill this gap. To inform design of this facility, this report documents an evaluation performed by a panel of technical experts that identifies key physical phenomena associated with maritime nuclear power plant operation that are important to safety and reliability and have a low state of knowledge. This evaluation concluded that all of the major test needs to understand a nuclear plant's behavior in a ship motion environment are for thermal-hydraulics related phenomena. Based on this finding, the evaluation panel concluded that the SMTF would not be required to use nuclear fuel to provide a heat source for the fluids. This would significantly decrease the cost, timeline, and complexity for this test facility to enable the advancement of maritime nuclear technology.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Analysis of controls for integrated energy storage system in energy arbitrage configuration with concrete thermal energy storage
Novel highly detailed NuScale reactor and turbo-generator models have been built in Modelica to analyze potential controls for an integrated energy system (IES) configuration combining a nuclear reactor and concrete thermal energy storage (CTES) operating in energy arbitrage. A shakedown test demonstrates the viability of the control schemes to operate across expected operating conditions while maintaining safety limits and buffers. A five-day data set from Idaho is used to show a load-following application of this system that increases the average reactor power level by 20% over the data set and matching demand at nearly all points. Here, by integrating these models, feedback mechanisms across the primary and secondary sides of the nuclear reactor are integrated, and realistic control evaluation is done. All models are consistent with the Hybrid repository, a suite of Modelica models, maintained at Idaho National Laboratory.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Baseline Radiological Conditions for the Transient Reactor Test (TREAT) Restart
The Transient Reactor Test (TREAT) facility, located on the Idaho National Laboratory (INL), northwest of the Materials and Fuels Complex (MFC), has been in programmatic shutdown for more than twenty-three years. Construction of TREAT by the Teller Construction Co., Portland, Oregon, was started in February, 1958, and completed in early November, 1958. The reactor first achieved criticality on February 23, 1959. Operations continued until April 28, 1994, when the reactor was shut down due to the lack of a programmatic mission. Recently, the U.S. Department of Energy (DOE) identified a need to resume transient testing of new reactor fuels. The Resumption of Transient Testing Program (RTTP) has readied the TREAT facility to resume transient testing operations. The first reactor criticality since the shut down in 1994 was achieved on November 14, 2017. This report details the radiological measurements taken in the TREAT Reactor Building and the surrounding environment prior to and just after its restart in order to characterize the radiological conditions that existed prior to and during the initial restart operations. The authors would like to acknowledge the support from RTTP Management and the TREAT Operations staff for allowing the extensive radiological surveys, as well as the health physics technicians from the INL Radiological Control organization who performed many of the measurements referenced in this report.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Film condensation with high heat fluxes and scaled experiments using pure steam for reactor containment cooling
Condensation tests were performed using a newly developed test facility for scaling the passive containment cooling system (PCCS) to a small modular reactor (SMR). The PCCS of the SMR plays a pivotal role in ensuring greater safety, reliability, and compactness than what is afforded by traditional reactors. Therefore, a well-designed PCCS is essential to SMRs. However, previous studies and test data were unsuitable for scaling, due to high variation in the test geometry and operating conditions. This study intends to close this research gap by using a novel designed scaled test facility consisting of vertical condensing test sections featuring 1-, 2-, and 4-inch-diameter condensing tubes with annular water cooling, and by applying superheated and saturated steam with different steam mass flow ranges of 5–25 g/s. Further, the primary test data, including axial temperatures, mass flow rates, and pressures, were used in conjunction with a standard data reduction method to estimate critical parameters such as heat fluxes, heat transfer coefficients, and condensation rates. These scaled test data would support improving empirical correlations and validating condensation models to identify scaling distortion for SMR PCCSs.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Status of the OECD/NEA Watts Bar unit 1 benchmark and calculations of local reactor core power data of the Zero Power physics tests
The paper aims to provide an update on the status of the Organization for Economic Cooperation and Development (OECD) / the Nuclear Energy Agency (NEA) Tennessee Valley Authority (TVA) Watts Bar 1 (WB1) Multi-Physics Multi-Cycle depletion benchmark and the results associated with Exercise 1. The benchmark relies on the set of benchmark progression problems developed by the Department of Energy (DOE) Consortium for the Advanced Simulation of Light Water Reactors (CASL) for the Virtual Environment for Reactor (VERA), which are based on real plant design and operational data. The OECD/NEA TVA WB1 benchmark is designed for validation of both traditional and novel high-fidelity multi-physics codes to analyze Pressurized Water Reactors (PWR) depletion cycles. The activities are conducted under the Expert Group on Reactor Systems Multi-Physics (EGMUP) at NEA/OECD. In this work, we analyze Exercise 1 of the benchmark: stand-alone Three-Dimensional (3D) neutronics at Start-up Zero Power Physics Test (ZPPT) at Hot Zero Power Conditions (HZP) and the resulting power maps. The code used to model the exercise is the continuous energy Monte Carlo code Serpent 2.1.31 along with the nuclear data library ENDF/B-VII.1. Serpent results are compared with the high-fidelity deterministic code MPACT, which is part of VERA. The paper presents three selected results from power calculations required output. The results compared are the normalized axially integrated radial core power maps, the normalized axial averaged core power shapes, and the normalized core hottest and coldest assemblies radial power maps. The Root Mean Square Deviation (RMSD) between Serpent and MPACT is 0.51 % for the axially integrated radial core power map, 1.41 % for the axial averaged core power shape, 2.42 % for the hottest assembly axial power shape, 0.88 % for the coldest assembly axial integrated hottest assembly radial pin power map, and 0.17 % for the axially integrated coldest assembly radial pin power map. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
SPC-2918 Rev 0 MARVEL Test Loop Flowmeters
This specification contains the requirements for the design, manufacture, testing, and delivery of magnetic flowmeters to be installed on the MARVEL Primary Coolant Apparatus Test (PCAT) system. The PCAT system is a heated test loop which will be used to simulate the coolant system of a proposed sodium or NaK cooled microreactor. The system has four identical cooling loops. Installation of flowmeters on all four loops is anticipated at this time, however, the number of flowmeters may be reduced by the Project. Note that the number of flowmeters described herein is not a commitment to a particular number of procured units. The number of flowmeter units procured shall be identified on the purchase order. The MARVEL reactor is still under design, so the PCAT flowmeter design may not be directly transferrable to the reactor. It is intended that the design and development effort for the PCAT flowmeters be leveraged to the extent practical for the reactor flowmeter design.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores
A team from the national laboratories conducted a study that assessed advanced reactor technology options and the irradiation needs of a broad group of stakeholders. This identified a gap in the fast neutron irradiation capability - the need for a test reactor that can deliver high neutron fluxes of fast neutrons. The study compared fast and thermal neutron irradiation technology options based on technology maturity and the ability to meet strategic objectives, along with pathways to deployment and tradeoffs in mission. It also provided a recommendation on specific technology options to address the current gap in fast neutron irradiation capabilities. In this context, a detailed scope of work was prepared that pertains to a trade study focusing on the core of a fast neutron spectrum test reactor. This report presents the main results obtained by a team of Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) scientists. It aims at developing an overall understanding of how the various areas important to the design and successful operation of such a facility interact with one another. Developing a sense of the relevant sensitivities at a systems level creates an understanding of where and how margins may be available to improve performances (e.g., flux level, capacity factor, and thermal hydraulic limits). Emphasis is given to the reactor (core design options and safety) and its fuel cycle (fuel supply and fabrication, as well as spent fuel management).
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Updated Results of Ultrasonic Transducer Irradiation Test
Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10{sup 21} n/cm{sup 2}. A multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET-ASI) program also provided initial support for this effort. This irradiation, which started in February 2014, is an instrumented lead test and real-time transducer performance data are collected along with temperature and neutron and gamma flux data. The irradiation is ongoing and will continue to approximately mid-2015. To date, very encouraging results have been attained as several transducers continue to operate under irradiation. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Uncertainty and Sensitivity Analysis of TREAT Transient Test no. 15
The U.S. Department of Energy (DOE) is preparing for the resumption of transient testing using the Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory (INL). In order to support the analysis of TREAT operation and experiment planning, the MAMMOTH reactor physics team has recently begun development of uncertainty and sensitivity analysis for the TREAT core simulated via MAMMOTH. MAMMOTH is a high-fidelity three-dimensional transient multi-physics code developed at INL under the DOE-NE Advanced Modeling and Simulation program (NEAMS). This research is part of the development presenting some initial uncertainty and sensitivity analysis of TREAT Transient Test no. 15. In this summary, we employ the Sobol sampling strategy implemented in RAVEN to efficiently perform the sensitivity and uncertainty analysis. The idea is using Sobol functional decomposition with generalized polynomial chaos (gPC) expansions to estimate the responses of interest. The gPC coefficients are then computed via the sparse grid spectral projection method. In addition, this strategy includes evaluation of the second or higher order sensitivity indices, allowing the study of interactions between several input variables. The principal focus of this analysis is to quantify the uncertainties in the total power and maximum fuel temperatures with respect to uncertainties in the initial transient conditions, such as rated power, fuel temperature, graphite fuel density and control rod boron concentration. In addition, both the first order and second order time-dependent sensitivity indices are computed and analyzed. The results show that the interactions between initial control rod boron concentration and the rated power will contribute at most 1% to the total uncertainty of the power peak.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Scaling Analysis of Vortex Shedding Frequency between In line Plates
The Global Threat Reduction initiative (GTRI) was established by the National Nuclear Security Administration (NNSA) in 2004. One of the highlights of the program is to convert research and test reactors at home and abroad from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuels. By switching directly from HEU to LEU, an LEU reactor produces fewer neutrons than an HEU reactor and in some cases; a direct switch from HEU to LEU results in the reactor being unable to achieve criticality. Under certain circumstances, criticality is achieved and the original neutron fluxes are restored with LEU fuel by adding more fuel elements to the core. Not all reactors, exemplified by the Advanced Test Reactor (ATR) at the Idaho National Lab (INL), have the luxury of simply adding more fuel to the core. In such cases, entirely new designs must be created and tested. The ATR utilizes plate type fuel. During the design and testing of the ATR fuel, vortex shedding was identified as potential source for experiment and fuel failure. For fuel conversion, elements must be tested in pile prior to the qualification of the new fuel design. For some test elements, the primary test plate is followed by a secondary handling plate. It is known that objects downstream influence the vortex shedding. While vortex shedding is an important and well-studied phenomenon, not only in the context of nuclear engineering, but in fields as diverse as aeronautical engineering and biology, there is a significant gap in the understanding of vortex shedding in regards to two bluff bodies in line direct line with one another. The study provides a scaling analysis of the frequency of vortex shedding for plates in line. The scaling analysis is then compared to numerical simulation of two in line plates. The study provided a scaling analysis and suggested parameter for determining the expected trends in the Strouhal number for a plate with some object downstream. The analysis made the consideration of differing length scales as the cause for a reduction in vortex shedding frequency as well as looked at the potential impact of boundary layer growth on vortex shedding for a flat plate. The comparison between the expected scaling and numerical data was less than ideal, but there are several potential causes. One of the biggest issues with the comparison is the uncertainty associated with the defect velocity. The frequency was well established in several cases, but the velocity was not always well established resulting in a larger uncertainty. In the cases where the velocity was near zero (i.e. b/d = 0.5), the uncertainty dominated since the corrected Strouhal number is divided by the defect velocity. The Reynolds numbers were not incredibly large, ranging from 350-1600. The low Reynolds numbers potentially undermine the principle assumption made in the scaling analysis, i.e. Reynolds number is negligibly small. Additionally, the exact definition of the 'reduced velocity' was never explicitly defined, merely inferred from the assumption that the x length scale and y length scale were in fact different. Another potential problem is the defect velocities in the lower Reynolds number cases were not much smaller than the free stream velocity. Further work needs to be done at higher Reynolds numbers and with more data points to ensure the assumptions made are valid as well as confirm the accuracy of the scaled relationship. (authors)
42 ENGINEERING↗
Investigation of similarity metrics for simulation based scaling analysis
Scaling analysis is a technique for characterizing degree of similarity between processes to occur in a real prototypic system and those to occur in its reduced scale (experimental) analog. In the context of nuclear reactor thermal-hydraulics and safety, scaling analysis is instrumental in assessing the applicability of data obtained from a scaled test facility for calibration of models and validation of simulation codes. Quantification of scale-related distortions is the key element in scaling analysis. In this work, we examine an approach for scaling analysis using computer simulation codes. Similarity analysis using information-theoretic approach is performed by comparing performance of experiment and application to variation of system conditions. Three different information theoretic measures namely mutual information, Kullback-Leibler divergence and overlap coefficient are selected for their potential use as metric for similarity analysis. We test the metrics by comparing them against physics-based scaling ratios, which are well known in fundamental thermal-hydraulic processes. In this exploratory analysis, a case of pressure drop of a steady-state flow through a channel is employed to test the proposed approach. It was found that in the absence of model uncertainties, mutual information serve as an effective measure of the scale effects present in the experiment. However, model uncertainty can significantly impact the quality of similarity analysis. Our analysis indicates that model uncertainties can have significant impact on the information theoretic metrics used for scaling analysis. Furthermore, if the model uncertainties exceeds beyond a critical value, the scale effect present in the experiment cannot be adequately evaluated. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Development of Sodium Fire Analysis Code Capabilities for Versatile Test Reactor
Not Available
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals
Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗