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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)

An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air-ingress mitigation methods are proposed in this study. Among them, the following two mitigation ideas are extensively investigated using computational fluid dynamic codes (CFD): (1) helium injection in the lower plenum, and (2) reactor enclosure opened at the bottom. The main idea of the helium injection method is to replace air in the core and the lower plenum upper part by buoyancy force. This method reduces graphite oxidation damage in the severe locations of the reactor inside. To validate this method, CFD simulations are addressed here. A simple 2-D CFD model is developed based on the GT-MHR 600MWt design. The simulation results showed that the helium replace the air flow into the core and significantly reduce the air concentration in the core and bottom reflector potentially protecting oxidation damage. According to the simulation results, even small helium flow was sufficient to remove air in the core, mitigating the air-ingress successfully. The idea of the reactor enclosure with an opening at the bottom changes overall air-ingress mechanism from natural convection to molecular diffusion. This method can be applied to the current system by some design modification of the reactor cavity. To validate this concept, this study also uses CFD simulations based on the simplified 2-D geometry. The simulation results showed that the enclosure open at the bottom can successfully mitigate air-ingress into the reactor even after on-set natural circulation occurs.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2011

Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be deployed commercially and have only been demonstrated in testing at a laboratory scale.

Moses, David Lewis↗
DOE OSTI2012

SHARP/PRONGHORN Interoperability: Mesh Generation

Progress toward collaboration between the SHARP and MOOSE computational frameworks has been demonstrated through sharing of mesh generation and ensuring mesh compatibility of both tools with MeshKit. MeshKit was used to build a three-dimensional, full-core very high temperature reactor (VHTR) reactor geometry with 120-degree symmetry, which was used to solve a neutron diffusion critical eigenvalue problem in PRONGHORN. PRONGHORN is an application of MOOSE that is capable of solving coupled neutron diffusion, heat conduction, and homogenized flow problems. The results were compared to a solution found on a 120-degree, reflected, three-dimensional VHTR mesh geometry generated by PRONGHORN. The ability to exchange compatible mesh geometries between the two codes is instrumental for future collaboration and interoperability. The results were found to be in good agreement between the two meshes, thus demonstrating the compatibility of the SHARP and MOOSE frameworks. This outcome makes future collaboration possible.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

GIF Very High Temperature Reactor: Proliferation Resistance and Physical Protection White Paper

This report is part of a series of six white papers, prepared jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) and the six System Steering Committees (SSCs) and provisional System Steering Committees (pSSCs). This publication is an update to a similar series published in 2011 presenting the status of Proliferation Resistance & Physical Protection (PR&PP) characteristics for each of the six systems selected by the Generation IV International Forum (GIF) for further research and development, namely: the Sodium-cooled fast Reactor (SFR), the Very high temperature reactor (VHTR), the gas-cooled fast reactor (GFR), the Molten salt reactor (MSR) and the Supercritical water–cooled reactor (SCWR). This white paper represents the status of Proliferation Resistance and Physical Protection (PR&PP) characteristics for the Very-High-Temperature Reactor (VHTR) reference designs selected by the Generation IV International Forum (GIF) VHTR System Steering Committee (SSC). The intent is to generate preliminary information about the PR&PP features of the VHTR reactor technology and to provide insights for optimizing their PR&PP performance for the benefit of VHTR system designers. It updates the VHTR analysis published in the 2011 report “Proliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy Systems”, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) and the System Steering Committees and provisional System Steering Committees of the Generation IV International Forum, taking into account the evolution of both the systems, the GIF R&D activities, and an increased understanding of the PR&PP features. The white paper, prepared jointly by the GIF PRPPWG and the GIF VHTR SSC, follows the high-level paradigm of the GIF PR&PP Evaluation Methodology to investigate the key points of PR&PP features extracted from the reference designs of VHTRs under consideration in various countries. A major update from the 2011 report is an explicit distinction between prismatic block-type VHTRs and pebble-bed VHTRs. The white paper also provides an overview of the TRISO fuel and fuel cycle. For PR, the document analyses and discusses the proliferation resistance aspects in terms of robustness against State-based threats associated with diversion of materials, misuse of facilities, breakout scenarios, and production in clandestine facilities. Similarly, for PP, the document discusses the robustness against theft of material and sabotage by non-State actors. The document follows a common template adopted by all the white papers in the updated series.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Fabrication and Design Aspects of High-Temperature Compact Diffusion Bonded Heat Exchangers

The Very High Temperature Reactor (VHTR) using gas-cooled reactor technology is anticipated to be the reactor type for the Next Generation Nuclear Plant (NGNP). In this reactor concept with an indirect power cycle system, a high-temperature and high integrity Intermediate Heat Exchanger (IHX) with high effectiveness is required to efficiently transfer the core thermal output to a secondary fluid for electricity generation, hydrogen production, and/or industrial process heat applications. At present, there is no proven IHX concept for VHTRs. The current Technology Readiness Level (TRL) status issued by NGNP to all components associated with the IHX for reduced nominal reactor outlet temperatures of 750–800 degrees C is 3 on a 1–10 scale, with 10 indicating omplete technological maturity. Among the various potential IHX concepts available, diffusion bonded heat exchangers (henceforth called printed circuit heat exchangers, or PCHEs) appear promising for NGNP applications. The design and fabrication of this key component of NGNP with Alloy 617, a candidate high-temperature structural material for NGNP applications, are the primary focus of this paper. In the current study, diffusion bonding of Alloy 617 has been demonstrated, although the optimum diffusion bonding process parameters to engineer a quasi interface-free joint are yet to be determined. The PCHE fabrication related processes, i.e., photochemical etching and diffusion bonding are discussed for Alloy 617 plates. In addition, the authors’ experiences with these non-conventional machining and joining techniques are discussed. Two PCHEs are fabricated using Alloy 617 plates and are being experimentally investigated for their thermal-hydraulic performance in a High-Temperature Helium Facility (HTHF). The HTHF is primarily of Alloy 800H construction and is designed to facilitate experiments at temperatures and pressures up to 800 degrees C and 3 MPa, respectively. Furthermore, some preliminary microstructural and mechanical property characterization studies of representative diffusion bonded Alloy 617 specimens are presented. The characterization studies are restricted and less severe from an NGNP perspective but provide sufficient confidence to ensure safe operation of the heat exchangers in the HTHF. The test results are used to determine the design operating conditions for the PCHEs fabricated.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

Understanding Fundamental Material Degradation Processes in High Temperature Aggressive Chemomechanical Environments

The objective of this project is to develop a fundamental understanding of the mechanisms that limit materials durability for very high-temperature applications. Current design limitations are based on material strength and corrosion resistance. This project will characterize the interactions of high-temperature creep, fatigue, and environmental attack in structural metallic alloys of interest for the very high-temperature gas-cooled reactor (VHTR) or Next–Generation Nuclear Plant (NGNP) and for the associated thermo-chemical processing systems for hydrogen generation. Each of these degradation processes presents a major materials design challenge on its own, but in combination, they can act synergistically to rapidly degrade materials and limit component lives. This research and development effort will provide experimental results to characterize creep-fatigue-environment interactions and develop predictive models to define operation limits for high-temperature structural material applications. Researchers will study individually and in combination creep-fatigue-environmental attack processes in Alloys 617, 230, and 800H, as well as in an advanced Ni-Cr oxide dispersion strengthened steel (ODS) system. For comparison, the study will also examine basic degradation processes in nichrome (Ni-20Cr), which is a basis for most high-temperature structural materials, as well as many of the superalloys. These materials are selected to represent primary candidate alloys, one advanced developmental alloy that may have superior high-temperature durability, and one model system on which basic performance and modeling efforts can be based. The research program is presented in four parts, which all complement each other. The first three are primarily experimental in nature, and the last will tie the work together in a coordinated modeling effort. The sections are (1) dynamic creep-fatigue-environment process, (2) subcritical crack processes, (3) dynamic corrosion – crack initiation processes, and (4) modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Baseline Characterization Database Verification Report - NBG-18 Billet 635-14

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Next Generation Nuclear Plant (NGNP)/Very High Temperature Reactor (VHTR)/Advanced Reactor Technology (ART). This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite regarding individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Data Management and Analysis System (NDMAS), and from that database will help populate handbook of property data available to member nations of the Generation IV International Forum (GIF).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Challenges in the Development of Advanced Reactors

Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

42 ENGINEERING↗