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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

sCO 2 Brayton Cycle: Roadmap to sCO 2 Power Cycles NE Commercial Applications

The mission of the Energy Conversion (EC) area of the Advanced Reactor Technology (ART) program is to commercialize the sCO 2 Brayton cycle for Advance Reactors and for the Supercritical Transformational Electric Production (STEP) program. The near-term objective of the EC team efforts is to support the development of a commercially scalable Recompression Closed Brayton Cycle (RCBC) to be constructed for the first STEP demonstration system with the lowest risk possible. This document details the status of technology, policy and market considerations, documentation of gaps and needs, and outlines the steps necessary for the successful development and deployment of commercial sCO 2 Brayton Power Systems along the path to nuclear reactor applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

National Reactor Innovation Center NRC Early Site Permit Roadmap

The National Reactor Innovation Center (NRIC) at Idaho National Laboratory (INL) in Idaho Falls, Idaho, was authorized under the Nuclear Energy Innovation Capabilities Act (Public Law 115–248) to provide private sector technology developers with resources and infrastructure for testing, demonstration, and performance assessment to accelerate deployment of new advanced reactor technology concepts. NRIC’s mission is to enable clean, affordable, reliable energy by supporting U.S. government investments in nuclear energy research, development, demonstration, and commercialization of new nuclear energy systems. It intends to provide existing facilities and other undeveloped and previously developed sites at INL to advanced reactor vendors for prototype technology testing and deployment. Some of these demonstration projects may be subject to DOE authorization, while other projects may require approval by the U.S. Nuclear Regulatory Commission (NRC) prior to construction and operation. NRIC and INL are considering whether to pursue one or more ESPs to help facilitate new reactor deployment at the INL Site. An ESP would allow early investment in licensing infrastructure in advance of a reactor design being ready to deploy. Through the ESPs, NRIC would help reduce the cost and time required to deploy demonstration reactors at the INL Site. This also will allow INL to bring its nuclear and siting experience to the ESP development process allowing for better and more efficient improvements and utilization of the Site. A preapproved ESP at INL will also remove the uncertainty of NRC siting approvals from the demonstration project proponents, who may not be well positioned to complete the analyses in a timely manner. The purpose of this NRIC ESP roadmap is to provide rigor and understanding needed for INL and DOE decision-makers to make data-informed decisions on pursuing NRC ESPs in support of future advanced reactor demonstrations at INL. The roadmap provides a structured method for addressing potential challenges and identifying options for resolution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

National Reactor Innovation Center NRC Early Site Permit Roadmap

The National Reactor Innovation Center (NRIC) at Idaho National Laboratory (INL) in Idaho Falls, Idaho, was authorized under the Nuclear Energy Innovation Capabilities Act (Public Law 115–248) to provide private sector technology developers with resources and infrastructure for testing, demonstration, and performance assessment to accelerate deployment of new advanced reactor technology concepts. NRIC’s mission is to enable clean, affordable, reliable energy by supporting U.S. government investments in nuclear energy research, development, demonstration, and commercialization of new nuclear energy systems. It intends to provide existing facilities and other undeveloped and previously developed sites at INL to advanced reactor vendors for prototype technology testing and deployment. Some of these demonstration projects may be subject to DOE authorization, while other projects may require approval by the U.S. Nuclear Regulatory Commission (NRC) prior to construction and operation. NRIC and INL are considering whether to pursue one or more ESPs to help facilitate new reactor deployment at the INL Site. An ESP would allow early investment in licensing infrastructure in advance of a reactor design being ready to deploy. Through the ESPs, NRIC would help reduce the cost and time required to deploy demonstration reactors at the INL Site. This also will allow INL to bring its nuclear and siting experience to the ESP development process allowing for better and more efficient improvements and utilization of the Site. A preapproved ESP at INL will also remove the uncertainty of NRC siting approvals from the demonstration project proponents, who may not be well positioned to complete the analyses in a timely manner. The purpose of this NRIC ESP roadmap is to provide rigor and understanding needed for INL and DOE decision-makers to make data-informed decisions on pursuing NRC ESPs in support of future advanced reactor demonstrations at INL. The roadmap provides a structured method for addressing potential challenges and identifying options for resolution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

AGC-3 Irradiation Creep Strain Data Analysis

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete irradiated properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

42 ENGINEERING↗
DOE OSTI2020

AGR-3/4 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the third and fourth test irradiations, Advanced Gas Reactor-3/4 (AGR-3/4), began at INL in April 2015. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing technical input, preparations for PIE testing and analysis, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

AGR-3/4 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the third and fourth test irradiations, Advanced Gas Reactor-3/4 (AGR-3/4), began at INL in April 2015. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing technical input, preparations for PIE testing and analysis, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

AGR-3/4 PIE at Oak Ridge National Laboratory

The Advanced Reactor Technologies (ART) Technology Development Office (TDO) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the third and fourth test irradiations, Advanced Gas Reactor-3/4 (AGR-3/4), will begin at Idaho National Laboratory (INL) in approximately April 2015. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing technical input, developing equipment and techniques for small-area deconsolidation of compacts, and initiating a system design for hot cell operations (expected to begin in about April 2015), and contributing expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by the Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing stack load data. Missing stack loads were estimated from measured ram pressures using regression equations developed from the existing data from cycle 157D. Estimated stack loads during this period are considered to be an accurate representation of actual load applied to the stacks. These loads deviate slightly from the planned loads. This deviation does not prevent the data from being Qualified for use, but must be taken into account when analyzing the effect of load on creep. Stack displacement increased consistently throughout the first two cycles with total displacement ranging from 0.4 to 0.8 in. During ATR outages, a set of pneumatic rams raised the stacks of graphite creep specimens to ensure the specimens were not stuck within the test train. This stack raising was performed twice. All stacks were raised successfully each time. The load and displacement data are Qualified for use by the ART TDO Program.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Radial Deconsolidation of Leach-Burn-Leach of AGR-3/4 Compacts 3-3, 12-1, and 12-3

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2016). A series of fuel irradiation experiments is being performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Establishing Regulatory Jurisdictional Boundaries at Collocated Advanced-Reactor Facilities

This white paper discusses establishing and applying nuclear facility jurisdictional boundaries at advanced-nuclear-reactor facilities. It was written for industry review and evaluation with possible consideration by the U.S. Nuclear Regulatory Commission (NRC) for subsequent regulatory action. The paper proposes a regulatory basis for establishing jurisdictional boundaries at operational advanced (i.e., non-light water) reactor installations at sites proximate to and sharing systems with non-NRC regulated facilities (e.g., users of process heat, fossil-plant retrofits, microgrid electrical power, desalination, etc.). Advanced-reactor technologies can be applied to support many industrial applications to replace the burning of fossil fuels as well as produce steam for electricity. The principal application of the current light-water reactor fleet is electricity generation. These varied industrial applications may involve an advanced-reactor design in combination with different site-specific energy-conversion systems. Some of these process-heat applications will require process-heat delivery systems to lie partially outside the advanced-reactor operator’s facility. Energy conversion systems are conventional, non-nuclear equipment and buildings. Given these varied applications, there should be a clear understanding between the advanced-reactor applicant and the NRC regarding a nominal demarcation between those systems that reside within the nuclear facility under the regulatory jurisdiction of the NRC (i.e., within the scope of a 10 Code of Federal Regulations (CFR) 50 operating license or Part 52 design certification and a combined license) and those that fall outside the scope of the NRC (e.g., an industrial facility). Additionally, it is important to have a clear understanding regarding the plant scope that should be addressed in an advanced-reactor facility Part 52, design certification application and the part of the plant scope that could be addressed as part of a site specific combined license application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Influence of Sodium on the Tensile Properties of Optimized Grade 92 Steel

This report provides an update on the evaluation of sodium compatibility of advanced alloy, optimized Grade 92, and the reference materials. The report is a deliverable (Level 3) in FY15 (M3AT-15AN1602072), under the Work Package AR-15AN160207, “Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies program. This work package supports the advanced structural materials development by providing corrosion, microstructure, and tensile property data from the standpoint of sodium compatibility of advanced structural alloys. During this fiscal year, sodium exposure experiments were performed at 600°C in forced convection sodium loops at Argonne National Laboratory. The oxygen content of sodium was controlled by the cold-trapping method to achieve ≈1 wppm oxygen level. Specimens were removed from the sodium loop after a pre-determined exposure time for post-sodium-exposure-examination, including weight and thickness measurements, cross-sectional examinations, microhardness measurements, microstructural characterization, and tensile tests at the sodium exposure temperature. Data of sodium-exposed specimens were compared with thermal aging data obtained at 650°C to separate the effects of sodium exposures and thermal aging. This report adds to the results on corrosion performance, microstructure and tensile properties of ferritic-martensitic exposed to sodium at 600°C. Specimens of G92-0, Opt. G92-4, Opt. G92-4T, and G91-H30176 for times up to 8,200 h, and Opt. G92-8T for 1,653 h were exposed to sodium at 600°C.

36 MATERIALS SCIENCE↗
DOE OSTI2023

AGR-1, AGR-2, AGR-3/4, and AGR-5/6/7 DimensionalChange Analysis

A series of fuel irradiation experiments have been planned in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to support the licensing and operation of the Advanced Reactor Technologies high temperature gas-cooled reactor. The advanced gas reactor (AGR) experiments are comprised of multiple independent capsules containing multiple cylindrical fuel compacts, placed inside of a graphite cylinder in ATR. The purpose of the AGR experiments is to provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for accident testing, and support the development of fuel performance and fission product transport models. The advanced graphite creep (AGC) experiments provide irradiation creep data for design and licensing. To date, six irradiation campaigns have been completed: AGR-1 (December, 2006 – November, 2009); AGR-2 (June, 2010 – October, 2013); AGR-3/4 (December, 2011 – April, 2014); AGC-1 (September, 2009 – January, 2011); AGC-2 (April, 2011 – May, 2012); and AGC-3 (November, 2012 – April, 2014).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DOE ART-GCR Program Overview

To provide Advanced Reactor Technologies Gas-Cooled Reactor Program: Overview of ART-GCR and Graphite program objectives, status, and activities. Also Identify research areas and outcomes that will benefit stakeholders and clients (HTGR designers, suppliers, regulators, DOE-NE, etc.), and to Identify remaining R&D gaps and future needs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preliminary results of AGC-4 irradiation in the advanced test reactor and design of AGC-5 - 18469

The U.S. Department of Energy Advanced Reactor Technologies Program will irradiate up to six nuclear graphite creep experiments in Idaho National Laboratory Advanced Test Reactor (ATR). The graphite experiments are being irradiated over an approximate 8-year period to support development of a graphite irradiation performance database on new nuclear grade graphites now available for use in high-temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the very high-temperature gas reactor and other future gas reactors. Each of the experiments consist of a single capsule that contains six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load and the other half of specimens not subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of the diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with online temperature and compressive load monitoring and control. Also, samples are taken of sweep gas effluent to measure any oxidation or off-gassing that may occur during initial startup of the experiment. The first experiment Advanced Graphite Capsule (AGCJ-I started irradiation in September 2009 and completed in January 2011; the second experiment AGC-2 started irradiation in April 2011 and completed in May 2012; the third experiment AGC-3 started irradiation in late November 2012 and completed in April 2014; and AGC-4 is currently being irradiated in ATR. This paper will briefly discuss preliminary irradiation results from the AGC-4 experiment and design of AGC-3. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Baseline Characterization Database Verification Report – IG-110 Billet 08-9-052-7

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 Advanced Reactor Technologies (ART) program. This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Nuclear Data Management and Analysis System (NDMAS), and from that database will help populate the handbook of property data available to member nations of the Generation IV International Forum (GIF). The transfer of this data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for IG-110 billet 08-9-052-7 and facilitates the release of the associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Reliability and Integrity Management Program Implementation Approach

Nuclear energy is the most reliable and environmentally sustainable energy source available today. In the United States, nuclear-generated power accounts for approximately 20% of total electricity and over 55% of clean energy. New advanced reactors have enormous potential to help further decarbonize the energy market, enhance grid resiliency, create new jobs, and build a stronger economy. More than 50 new reactors are being developed in the United States, and the federal government realizes an urgent need to deploy nuclear technologies to meet the country’s energy, environmental, and national security goals. As such, the U.S. Department of Energy launched multiple programs to support advanced reactor deployment. The research described in this paper explores implementation strategies for the Reliability and Integrity Management Program that directly supports the U.S. Department of Energy goal to enable the near-term deployment of the advanced reactor technologies. The project is conducted under the Regulatory Development Program for advanced reactors sponsored by the U.S. Department of Energy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Investigation and Demonstration of Reliability Target Allocation to Support Reliability and Integrity Management Program

Nuclear energy is the most reliable and environmentally sustainable energy source available today. In the United States (U.S.), nuclear-generated power accounts for approximately 20% of total electricity and over 55% of clean energy. New advanced reactors have enormous potential to help further decarbonize the energy market, enhance grid resiliency, create new jobs, and build a stronger economy. More than 50 new reactors are being developed in the U.S., and the federal government realizes an urgent need to deploy nuclear technologies to meet the country’s energy, environmental, and national security goals. As such, the U.S. Department of Energy (DOE) launched multiple programs to support advanced reactor deployment. The research described in this report explores implementation strategies for the Reliability and Integrity Management Program that directly supports the DOE goal to enable the near-term deployment of the advanced reactor technologies. The project is conducted under the Regulatory Development Program for advanced reactors sponsored by the DOE.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗