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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 1,153 records · Page 64
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

RELAP-7 Software Verification and Validation Plan

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

KiloPower Project - KRUSTY Experiment Nuclear Design

This PowerPoint presentation covers the following topics: Reference Kilopower configuration; Reference KRUSTY configuration; KRUSTY design sensitivities; KRUSTY reactivity coefficients; KRUSTY criticality safety and control; KRUSTY core activation/dose; and KRUSTY shielding, room activation/dose.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

TREAT Cask Stand Drawings for Unlimited Use

This report contains drawings of the HFEF-15 Cask stand in the TREAT Reactor Building (MFC-720) and the Cask Stand Tower on the south wall of the TREAT High Bay. Public release is sought so these can be sent to DOD-SCO and their contracted companies looking at options for fueling the PELE reactor prior to demonstration in the DOME or elsewhere at INL.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

The primary challenge preventing the deployment of reactor designs that leverage wrought FeCrAl as an advanced light-water reactor (LWR) cladding is irradiation hardening and embrittlement. Wrought FeCrAl alloys experience a loss of post-irradiation ductility and fracture toughness under low-temperature neutron irradiation (< 350°C)resulting from the combined effects of dislocation loop formation and the irradiation-enhanced precipitation of Cr-rich alpha-prime precipitates throughout the microstructure. Over the past decade, significant improvements in our understanding regarding the effect of Cr and Al content have been enabled through neutron irradiations, but even optimized wrought alloys such as C26M remain vulnerable to brittle failure during storage, transportation, and handling following irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Assessment of Space Nuclear Thermal Propulsion Facility and Capability Needs

The development of a Nuclear Thermal Propulsion (NTP) system rests heavily upon being able to fabricate and demonstrate the performance of a high temperature nuclear fuel as well as demonstrating an integrated system prior to launch. A number of studies have been performed in the past which identified the facilities needed and the capabilities available to meet the needs and requirements identified at that time. Since that time, many facilities and capabilities within the Department of Energy have been removed or decommissioned. This paper provides a brief overview of the anticipated facility needs and identifies some promising concepts to be considered which could support the development of a nuclear thermal propulsion system. Detailed trade studies will need to be performed to support the decision making process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Interfacial Area Transport through Vertical-Downward Elbows in Air-water Two-phase Flow

In the study of two-phase flow, models are often developed by the data acquired in a test section with fixed orientation and with no flow restrictions. In view of practical nuclear reactor systems, however, two-phase flow through flow restrictions is important. Flow restrictions and orientations can alter both global and local two-phase flow parameters such as: cause additional pressure loss; alter the flow regime transition; change bubble distributions; affect the bubble velocity and the bubble interaction mechanisms. However, there is insufficient database and predictive models that can account for such effects. In view of above, the objectives of the current study are: (1) to establish database for local and global two-phase flow parameters downstream of a horizontal-to-vertical glass elbow with a non-dimensional radius of curvature, R{sub c}/D, of three, (2) to develop predictive models in the elbow region, and (3) to apply the developed model to the interfacial area transport equation (IATE) and evaluate it in elbow regions. (authors)

42 ENGINEERING↗
DOE OSTI2019

The Annular Core Research Reactor (ACRR) Description and Capabilities

The Annular Core Research Reactor (ACRR) is a unique, one-of-a-kind nuclear reactor facility operated by Sandia National Laboratories for the National Nuclear Security Administration that is considered by many to be a national treasure. There is no other research reactor in the world that has the attributes and capabilities comparable to that of the ACRR. The ACRR was specifically designed to meet the irradiation testing needs of the U.S. nuclear weapons program. The ACRR has four major attributes that make it unique: 1) a large dry central cavity; 2) an epithermal neutron flux; 3) a large pulsing capability; and 4) a fueled-ring external cavity with a larger dry cavity. This report presents the unique capabilities of the ACRR and documents some of the more important metrics associated with its operation. ACKNOWLEDGEMENTS The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

FUEL PERFORMANCE MODELING OF REACTIVITY-INITIATED ACCIDENTS USING THE BISON CODE

The Fukushima Daiichi nuclear accidents in 2011 sparked considerable interest in the U.S. to develop new nuclear fuel with enhanced accident tolerance. Throughout the development of these new fuel concepts they will be extensively modeled using specialized computer codes and experimentally tested for a variety of different postulated accident scenarios. One accident scenario of interest, reactivity-initiated accident, is a nuclear reactor event involving a sudden increase in fission rate that causes a rapid increase in reactor power and temperature of the fuel which can lead to the failure of the fuel rods and a release of radioactive material. The focus of this work will be on the fuel performance modeling of reactivity-initiated accidents using the BISON code being developed at Idaho National Laboratory. The overall goal of this work is to provide the best possible modeling predictions for future experimental tests. Accurate predictive capability modeling using BISON is important for safe operation of these tests and provides a cheaper alternative to the expensive experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

FFTF main heat transport system/dump heat exchanger testing and performance evaluation (initial pub date April 1978)

This report covers testing of a Main Heat Transport System/Dump Heat Exchanger (MHTS/DHX) for the Fast Flux Test Facility. The primary objective of the test program was to verify the design and generate experience in the installation, operation, and control of the DHX for FFTF. This report discusses test results related to erection, preoperational testing, in-service inspections, and steady-state and transient performance. Emphasis is placed on the effect of the test results on successful DHX operation at FFTF. 36 figures, 9 tables.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Baseline Postirradiation Examination of the FUTURIX-FTA Experiments

Postirradiation Examination (PIE) results from the baseline examination of the FUTURIX-FTA irradiation experiment are presented in this report. The FUTURIX-FTA irradiation was designed to study the fuel performance of candidate transmutation fuels. The FUTURIX-FTA experiment was conducted in the true fast neutron spectrum conditions of the Phénix sodium fast test reactor in France. FUTURIX-FTA contained several compositions that were also irradiated as part of the AFC-1 test in cadmium filtered positions in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) that approximate the neutron spectrum of a fast reactor. This experiment was a joint collaboration between the Department of Energy (DOE) in the U.S. and the Commissariat á l'Energie Atomique et aux Energies Alternatives (CEA) in France. A joint agreement between DOE and CEA was established in 2004 to conduct an irradiation experiment of metallic and nitride fuels in the Phénix reactor at CEA. Both the FUTURIX-FTA tests and the AFC-1 tests were conducted for the DOE Advanced Fuels Campaign (AFC) that is part of Fuel Cycle Research & Development (FCRD) program. These programs seek to develop and demonstrate the technologies needed to transmute long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology. The long term goal of this work is to reduce the radiotoxicity of future high level waste repositories. Post-irradiation examination of irradiated fuels experiments provides data related to in reactor fuel performance and input into future fuel design choices. The compositions of the FUTURIX-FTA pins were designed to test if minor actinides could be incorporated into the fuel of a fast reactor without significantly changing fuel performance observed for previously well studied fuels. The fuel performance of metallic fuel in fast reactors has been well documented. Likewise the irradiation performance of nitride fuels in fast reactors has also been documented. However the existing database for nitride fuels is much smaller than the database for metal fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

TREAT neutronics analysis and design support, Part I: Multi-SERTTA

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis of a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.73 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Sensors for measuring thermal conductivity and related methods

A sensor for measuring thermal conductivity includes an insulator, a test material over the insulator, a conductor over the test material, and a gas within an open volume adjacent the test material and the conductor. An electrical source is configured to provide an alternating current through the conductor to heat the test material. Leads are connected to the conductor and configured to connect to a voltmeter. A method of measuring thermal conductivity includes disposing the sensor in a reactor core in which a nuclear fuel undergoes irradiation and radioactive decay. An alternating current is provided from the electrical source through the conductor to heat the test material. A voltage is measured as a function of time at the leads connected to the conductor. A thermal conductivity of the test material is calculated based on the voltage measured as a function of time. Methods of forming a sensor are also disclosed.

Daw, Joshua↗
DOE OSTI2024

DOME Reactor Testing User Guide

DOME User Guide History DOME User Guide Introduction User Engagement Lifecycle for DOME Phase 1 – Front End Engineering and Experiment Design (FEEED) Phase 2 – Detailed Design & Long Lead Procurement Phase 3 – Fabrication, Training, & Delivery Phase 4 – Installation & Testing Phase 5 – Removal & Disposal Next steps

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Uncertainty and Sensitivity Analysis of Reactivity Insertion Accidents using BISON Fuel Performance Code

The development of Accident Tolerant Fuels (ATF) has expanded into a considerable research program in the wake of the events at the Japanese Fukushima Daiichi nuclear power plant in 2011. The ATF goal is to insert a test assembly into a commercial reactor by 2022. To support this, testing of the ATF design under off-normal conditions is needed to assess the 'enhanced accident tolerance' capability of the fuel rod design over standard UO{sub 2}-Zircaloy concepts. Transient testing capabilities are necessary for off-normal testing of events such as loss of coolant accidents (LOCA) and reactivity insertion accidents (RIA). A transient testing program has been developed to support the testing of ATF designs. The Transient Reactor Test (TREAT) facility has been chosen as the site to perform the transient testing of the ATF designs. TREAT is a versatile test facility capable of subjecting fuel specimens to various off-normal conditions. Historic TREAT experiments were performed in a variety of experimental test vehicles including static environment capsules, once-through flowing systems, and also recirculating flowing systems. The planned TREAT experiments will be operated under PWR conditions. Therefore, a static capsule vehicle is under development to handle required temperatures and pressures. The static capsule, termed the Static Environment Rodlet Transient Test Apparatus (SERTTA), will be used for irradiation tests in TREAT to subject ATF fuel rodlets to a range of power pulse conditions and environments. To aid in the design and analysis of future tests in TREAT, the fuel performance code BISON is being used. Fuel performance codes capable of accident modeling have been developed for some time and they have shown an ability to reproduce experimental results with some degree of adequacy, but they still require extensive verification and validation. Future experiments in TREAT will add to a limited database of past transient tests performed in transient reactors to support LWR systems. These tests can be used for future verification and validation of fuel performance codes. Understanding the impact of the uncertainty of influential parameters is critical for not only analyzing the results of the experiment but also for modeling and validation of codes. For example, uncertainties associated with the initial manufactured state of the sample, instrument measurements, material property models, and planned experiment conditions can provide important insight into model performance, experimental data, and play an important role in experimental design. The objective of this work is to perform an uncertainty and sensitivity analysis on a simplified RIA case of a standard UO{sub 2}-Zircaloy sample using expected nominal conditions in the SERTTA capsule in TREAT. The BISON fuel performance code is used in conjunction with DAKOTA (Design Analysis Kit for Optimization and Tera-scale Applications) to perform the uncertainty and sensitivity analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

ECAR-6585 MARVEL Core Support Structure Analysis: Project #33526

Idaho National Laboratory (INL) is developing a small microreactor to produce electrical power utilizing a small nuclear core and stirling engines for power output under the Microreactor Application Research Validation and Evaluation (MARVEL) program. The MARVEL Reactor (Figure 1) will be installed in the Transient Reactor Test (TREAT) Facility. Walsh Engineering Services, PC (WES) has contracted with the INL to detail the final design. There are six main systems in the MARVEL, one of which is the Fuel and Core System (FCS). The FCS contains a subsystem–the Core Support Structures Subsystem (CSS)–that supports and orients each of the 36 nuclear fuel pins in the MARVEL Reactor System. The FCS (apart from the fixed Beryllium Oxide radial reflectors) is shown in light green and is depicted in more detail in Figure 1 and Figure 2. The main subject of this analysis is the lower grid plate, which is the principal item in the CSS which supports the fuel, see Figure 3. Additionally, the upper grid plate thermal expansion of the upper grid plate is presented in Appendix A.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Development and Standup Testing of a System to Observe Fission Gas Release During Transient Heating of Nuclear Fuel Element Segments

As part of an effort to increase the burnup of nuclear fuels used in Light Water Reactors, the mechanisms and timing of fission gas release during high temperature transients must be studied. In this work, a system was built to observe fission gas release of fuel element sections under evacuated conditions, at temperatures up to 800 °C. System capabilities and limitations are overviewed in this report. The system was further tested with a hydrogen-infused titanium foils and gas release was observed in agreement with literature observations. Pressure was monitored with two sensors, and temperature was monitored using a thermocouple and a pyrometer with continuous visual imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗