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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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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↗
DOE OSTI2020

Developing Separate Effects Transient Test Experiments Using an Out-of-Pile Flowing Water Loop

In 2017 the Transient Reactor Test Facility was restarted after being placed in a standby state since 1994. This Test Reactor’s restart has since enabled the progressive development of new nuclear technologies within the U.S. that were previously required to outsource to other countries. While the reactor’s restart was a large feat worthy of recognition, the experimental use of its characteristics has required the further development of in-pile experimental infrastructure sufficient to support programmatic needs. This hardware have taken the form of capsule designs (compact and elongated) as well as loop concepts representing the phenomena of interest for a subset of separate effects tests desire for each respective testing campaign. The Transient Testing Program has a large-integrated effort that aligns with the U.S. Department of Energy’s current needs. This study compliments those programmatic elements by developing, fabricating and demonstrating a full-scale flowing water loop in an out-of-pile environment. The goal of this effort is to develop a pragmatic understanding of the engineering capabilities and limitations associated with geometric form-factors, metering technology, and controls logic under representative thermal hydraulic conditions that would be experienced within the TREAT reactor during an in-pile reactivity initiated accident test. The outcomes of this study result in an evaluation of the conceptual design of a comprehensive flowing water loop including objective figures of merit (FoMs) for comparing unique instrumentation and the basis for their selection during operations. These efforts directly contribute to and are required for the further advancement of transient testing capabilities within the U.S.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Theory and Implementation of Nuclear Safety System Codes - Part II: System Code Closure Relations, Validation, and Limitations

This is Part II of two articles describing the details of thermal-hydraulic sys- tem codes. In this second part of the article series, the system code closure relationships (used to model thermal and mechanical non-equilibrium and the coupling of the phases) for the governing equations are discussed and evaluated. These include several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models be- tween the fields. These models are often developed from experimental data. The experiment conditions should be understood to evaluate the efficacy of the closure models. Code verification and validation, including Separate Effects Tests (SETs) and Integral effects tests (IETs) is also assessed. It can be shown from the assessments that the test cases cover a significant section of the system code capabilities, but some of the more advanced reactor designs will push the limits of validation for the codes. Lastly, the limitations of the codes are discussed by considering next generation power plants, such as Small Modular Reactors (SMRs), analyz- ing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, in- novative fuel types, and many other design features that may not be fully analyzed by current system codes. This second part completes the series on the comparison and evaluation of the selected reactor system codes by discussing the closure relations, val- idation and limitations. These two articles indicate areas where the models can be improved to adequately address issues with new reactor design and development.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Using Falcon to develop an RIA PCMI failure criteria

This paper documents the development of failure criteria for both pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rod failure due to pellet cladding mechanical interaction (PCMI) during a reactivity insertion accident (RIA). This study was initiated to use cladding mechanical properties using data from separate effects tests to provide a basis to translate nuclear safety research reactor (NSRR) test data to commercial light water reactor conditions. The approach employed for the revised licensing criteria combined elements of experimental results and analytical evaluations to develop a fundamental understanding of fuel behavior during RIA events. The transient behavior of intermediate and high burnup fuel rods using well characterized RIA simulation tests was established. Cladding mechanical properties using data from separate effects tests was also developed. The Zircaloy cladding mechanical properties database of separate effects tests was used to develop the cladding integrity models. The cladding integrity models were then used in combination with the application of the Falcon fuel behavior code to translate NSRR test data to commercial light water reactor conditions. Because of the complex manner in which burnup influences fuel rod failure, it was not possible to develop a fuel rod failure threshold directly from the experimental data. An alternative approach was required that combined analytical modeling and experimental data to develop the PCMI-related contribution to the failure threshold. The analytical approach computed the radial average peak fuel enthalpy required to cause cladding failure by PCMI as a function of hydrogen content. This was accomplished by applying a critical strain energy density (CSED)-based cladding integrity model using data from total and uniform elongation tests of irradiated low-tin Zircaloy-4, ZIRLO, and Zircaloy-2 cladding materials. The CSED was developed as a function of cladding hydrogen content because hydrogen induced embrittlement is a primary effect of burnup. It was found that the failure threshold decreased with hydrogen content due to the increase in PCMI loading as a result of gap closure effects and by the decrease in cladding ductility with hydrogen accumulation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Effect of Test Temperature and Strain Rate on Dynamic Mechanical Behavior of Aluminum Alloy 2519A

The dynamic mechanical behavior of aluminum alloy 2519A-T9I6 has been investigated by a split Hopkinson bar at temperature range from 298 to 523 K and at strain rates of from 10{sup 3} to 5 × 10{sup 3} s{sup −1} in this study. The results show that both the strain rate and test temperature have a significant effect on the dynamic behavior and microstructure evolution of aluminum alloy 2519A-T9I6. Under similar strain rate, the strengthening effect of strain rate decreases with the increase in the test temperature. Moreover, the dynamic yield strength at above 423 K drops sharply at the strain rate of about 5000 s{sup −1}, which is attributed to the flow softening caused by decomposition of θ′ precipitates at high strain rate and high temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Framework for Testing the Effectiveness of Bat and Eagle Impact-Reduction Strategies at Wind Energy Projects

The objectives of this framework are to facilitate the study design and execution to test the effectiveness of bat and eagle impact-reduction strategies at wind energy sites. Through scientific field research, the wind industry and its partners can help determine if certain strategies are ready for operational deployment or require further development. This framework should be considered a living document to be improved upon as fatality-reduction technologies advance from the initial concepts to proven readiness (through project- and technology-specific testing) and as scientific field methods improve.

17 WIND ENERGY↗
DOE OSTI2019

SRNL Report on Vendor Solicitation and Selective Ion-Electrode Testing

Effective operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) will require information collected from sampling and/or monitoring to maintain environmental compliance, product quality, and operating efficiency. The U.S. Department of Energy (DOE) – Office of River Protection (ORP) recognizes the potential long-term benefit of developing real-time, in-line monitoring (RTIM) at the WTP to improve the current strategy of collecting tens of thousands of samples and analyzing them. In-line monitoring can decrease processing time associated with waiting for sample analysis, reduce workers’ exposure to radiation and chemical hazards, decrease secondary waste volumes, potentially improve product quality, and provide production support and enhance overall operational control. The Assistant Manager, WTP chartered an Integrated Technical Team (ITT), which includes personnel from the Savannah River National Laboratory (SRNL), Bechtel National, Inc., and DOE-ORP, to evaluate the capability and potential benefits of replacing routine sample collection requirements with real-time, in-line monitoring. The initial focus of the ITT is the Direct Feed Low Activity Waste (DFLAW) Program. This work scope is being performed in multiple phases. Phase 1 is Requirements and Opportunities Analysis, Phase 2 is Applications and Benefits Determination, and Phase 3 is Qualification and Demonstration. Phase 1 included 6 tasks and was completed by SRNL in 2017. The tasks in Phase 1 were: (1) determining the functional requirements for determining analytes and properties of streams, (2) determining the technical basis for process control, (3) determining the sampling points within the Low Activity Waste (LAW) facility and Effluent Management Facility (EMF), (4) data quality and management, (5) process control challenges, and (6) preparing a Preliminary Analysis Plan. DOE-ORP was briefed on the findings from the Phase 1 tasks and requested that Phase 2 work focus on establishing a basis for a materials balance approach that minimized process sampling. The Phase 2 work scope contains tasks to evaluate a materials balance only approach as well as continuing scoping evaluations for instruments that could be used for RTIM. The Phase 2 work scope was completed by both SRNL and Pacific Northwest National Laboratory (PNNL). The tasks performed in Phase 2 were: 1. Overall program administration and technical oversight 2. Evaluation of a Material Balance Approach, a. Evaluation of uncertainty of this approach compared to uncertainty of current methods b. Evaluation of automating the material balance within the framework of existing control software, c. Development of the methodology to obtain and evaluate data during cold commissioning to support a material balance approach, and d. Development of a framework to utilize during presentation of the revised approach to stakeholders and regulators. 3. Screening Evaluations of Instruments to Supplement the Material Balance Approach, a. AP-105 melter condensate evaluations, and b. Screening testing with simulants. 4. Evaluation of the current DFLAW feed qualification program to identify analytes that could be eliminated, a. Inorganic analytes, b. Radionuclides, and c. Organic analytes. This report describes the results of Task 3b, Screening testing with simulants. The Savannah River National Laboratory requested demonstrations of commercially available analytical instruments by the vendors or laboratory demonstration. The techniques tested included: Raman spectroscopy, Infrared spectroscopy (IR), laser induced breakdown spectroscopy (LIBS), X-Ray Fluorescence Spectroscopy (XRF), and ion-selective electrochemical probes. Additionally, the instrumentation would need to be durable enough to withstand the environment of the plant, preferably with little to no maintenance. The insertion point of the instrument would be dependent upon the instrumentation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Postirradiation examination from separate effects irradiation testing of uranium nitride kernels and coated particles

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly accumulate burnup during separate effects irradiation testing. In general, the burnup and fuel temperatures measured postirradiation were consistent with the design calculations. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU and was achieved after only 68 effective full-power days of irradiation. The dilatometric evaluation of passive silicon carbide thermometry indicated that the fuel was irradiated at temperatures ranging from 410 to 460 °C. Because the irradiation temperatures and burnup were low, the UN kernels showed minimal fission gas release that was within the range of the expected recoil (athermal) release. While it is possible to measure fuel swelling using x-ray computed tomography, the observed swelling was too small to quantify in this case. Extensive microstructural characterization of the irradiated fuel was performed in this study, and no significant irradiation induced changes were observed.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Experimental study on the guillotine break accident of the main steam line with Atlas

A main steam line break (MSLB) accident is defined as a pipe break in the main steam system of a nuclear power plant. The MSLB accident is characterized as cool-down event due to the increased steam flow rate, which causes excessive energy removal from the steam generators and eventually reactor coolant system. This results in a decrease in reactor coolant temperature. The cool-down causes an increase in core reactivity due to the negative moderator and Doppler reactivity coefficients. The thermalhydraulic integral effect test was performed to simulate a double-ended guillotine break accident of the main steam line of a pressurized water reactor. With an aim of simulating the MSLB accident as realistically as possible, a pertinent scaling approach was taken, especially focusing on a break flow. The main objectives of this test were not only to provide physical insight into the system response of a nuclear power plant during the MSLB accident but also to produce an integral effect test data to validate the safety analysis codes. In the present test, asymmetric thermal-hydraulic behavior was clearly observed between the loops. In the cold leg-1A and -1B where the affected steam generator is connected, the flow rates were higher and the fluid temperatures were lower than those in the cold leg-2A and -2B. This could be attributed to the excessive heat removal in the cold leg-1A and -1B through the affected steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes. Furthermore, this data can be utilized to identify any code deficiency for a MSLB simulation, especially for DVI-adapted plants. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Experimental validation on the operational and cooling performance of the passive safety system with Atlas

After the Fukushima accident, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions. In Korea, a passive auxiliary feedwater system (PAFS) was developed and adopted to a GEN-III+ nuclear power plant, APR+ to completely replace the conventional auxiliary feedwater system. The integral effect tests were performed to validate the cooling and operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. The representative design basis accidents (DBAs) such as a feedwater line break, a steam line break, and a steam generator tube rupture transients were experimentally simulated in this study. In addition, an integral effect test was performed for simulating a station blackout transient as a limiting beyond DBA. From the integral effect test results, it could be concluded that the APR+ has the capability of coping with the hypothetical DBA and bDBA by adopting PAFS and proper set-points of its operation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary of appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Testing the effectiveness of an acoustic deterrent for gray whales along the Oregon coast

This study was conducted to determine whether a low-powered sound source could be effective at deterring gray whales from areas that may prove harmful to them. With increased interest in the development of marine renewal energy along the Oregon coast the concern that such development may pose a collision or entanglement risk for gray whales. A successful acoustic deterrent could act as a mitigation tool to prevent harm to whales from such risks. In this study, an acoustic device was moored on the seafloor in the pathway of migrating gray whales off Yaquina Head on the central Oregon coast. Shore-based observers tracked whales with a theodolite (surveyor’s tool) to accurately locate whales as they passed the headland. Individual locations of different whales/whale groups as well as tracklines of the same whale/whale groups were obtained and compared between times with the acoustic device was transmitting and when it was off. Observations were conducted on 51 d between January 1 and April 15, 2012. A total of 143 individual whale locations were collected for a total of 243 whales, as well as 57 tracklines for a total of 142 whales. Inclement weather and equipment problems resulted in very small sample sizes, especially during experimental periods, when the device was transmitting. Because of this, the results of this study were inconclusive. We feel that another season of field testing is warranted to successfully test the effectiveness of the deterrent, but recommend increasing the zone of influence to 3 km to ensure the collection of adequate sample sizes. Steps have been taken to acquire the necessary federal research permit modification to authorize the increased zone of influence and to modify the acoustic device for the increased power. With these changes we are confident we will be able to determine whether the deterrent is effective at deflecting gray whales. A successful deterrent device may serve as a valuable mitigation tool to protect gray whales, and other baleen whales, in the event that marine energy development poses a collision or entanglement risk.

16 TIDAL AND WAVE POWER↗
DOE OSTI2016

TESTING THE EFFECTS OF EXPANSION ON SOLAR WIND TURBULENCE

We present a multi-spacecraft approach to test the predictions of recent studies on the effect of solar wind expansion on the radial spectral, variance, and local 3D anisotropies of the turbulence. We found that on small scales (5000–10,000 km) the power levels of the B-trace structure functions do not depend on the sampling direction with respect to the radial suggesting that on this scale the effect of expansion is small possibly due to fast turbulent timescales. On larger scales (110–135 R{sub E}), the fluctuations of the radial magnetic field component are reduced by ∼20% compared to the transverse (perpendicular to radial) ones, which could be due to expansion confining the fluctuations into the plane perpendicular to radial. For the local 3D spectral anisotropy, the B-trace structure functions showed dependence on the sampling direction with respect to radial. The anisotropy in the perpendicular plane is reduced when the increments are taken perpendicular with respect to radial, which could be an effect of expansion.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2012

Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

MARS-KS code validation activity through the atlas domestic standard problem

The 2 nd Domestic Standard Problem (DSP-02) exercise using the ATLAS integral effect test data was executed to transfer the integral effect test data to domestic nuclear industries and to contribute to improving the safety analysis methodology for PWRs. A small break loss of coolant accident of a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. Ten calculation results using MARS-KS code were collected, major prediction results were described qualitatively and code prediction accuracy was assessed quantitatively using the FFTBM. In addition, special code assessment activities were carried out to find out the area where the model improvement is required in the MARS-KS code. The lessons from this DSP-02 and recommendations to code developers are described in this paper. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Scaling methodologies and similarity analysis for thermal hydraulics test facility development for water-cooled small modular reactor

Small modular reactors (SMRs) represent a promising option for providing clean and sustainable energy due to their potential for enhanced safety, reduced capital costs, and increased siting flexibility. However, new reactor systems require the development and operation of representative scaled-down test facilities to support the verification and validation of system computer codes and models. Here this study reviews the research on scaling methodologies and similarity principles pivotal in developing non-nuclear integral effects test and separate effects test facilities for water-cooled SMRs. The study focuses on a review of the scaling methods, similarity approaches, and possible challenges posed by the unique and compact design features of integral-pressurized water reactor-type SMRs, and their representative test facilities. This study also reviews previous research related to scaling and similarity methodologies and provides insights into design considerations for achieving prototypic conditions in test facilities. The findings and recommendations emphasize the broader impact of appropriate scaling and similarity principles to ensure meaningful and transferable results from non-nuclear test facilities to accelerate the safe and efficient deployment of next-generation water-cooled SMRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗