SEARCH · Search NASA

Results for “Advanced Reactor”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 649 records · Page 36

All sources NASA NTRS DOE OSTI

DOE OSTI2019

Assessment of Pre-Irradiation SiC CMC Joint Performance in Representative Cladding Geometries

Silicon carbide fiber reinforced, silicon carbide matrix composites (SiC-SiC) offer strength at high temperatures, corrosion resistance, and stability during irradiation, and are being developed by General Atomics as cladding for accident tolerant fuel (ATF) applications and advanced reactor concepts. Advanced joining techniques capable of withstanding harsh reactor environments are key to enable GA’s SiGAtm SiC-SiC composite. The objective of this work is to obtain joint-specific material properties for assemblies in representative planar and cladding geometries that will allow for the generation of a material properties database and more accurate simulation of SiC joint behavior across arange of temperatures and irradiation conditions. To support this, three different joint formulations are being assessed for mechanical and thermal performance pre and post irradiation in the High Flux Isotope Reactor (HIFR) at ORNL. GA has previously identified its hybrid SiC (HSiC) joint as the most promising and best suited joining approach for cladding applications [1]. Additionally, oxide and transeutectic phase (TEP) joints are being investigated to make the material property database more comprehensive as these joints have previously shown strength resilience under irradiation in planar geometries [2]. The thermal and mechanical data obtained in this work will address vital knowledge gaps, enabling more accurate modeling of joints in SiC-based components for accelerated material qualification.The scope of work presented here focuses on the mechanical and thermal performance of the joint material prior to irradiation in HFIR. Endplug push-out (EPPO) testing to determine joint strength in tube geometries, and shear strength testing for planar geometries have provided a benchmark for pre-irradiation mechanical performance. These results show that all three selected joint methodologies result in cladding joints capable of withstanding the expected joint stresses in a LWR reactor environment [3]. This has been accompanied by He leak testing of the joint regions on tube specimens to assess the suitability of each joint methodology for cladding applications. Here, only the HSiC formulation consistently met critical leak rate requirements suggesting that other joining strategies require additional development. Finally, pre-irradiation thermal testing has been completed to establish a baseline for comparison after irradiation. Subsequent irradiation of these joints types in HFIR will provide further detail on joint material properties and their appropriateness for nuclear applications.

Gonderman, Sean↗

DOE OSTI2022

Achieving American Leadership in the Nuclear Energy Supply Chain Factsheet

The United States is committed to achieving a 50 to 52 percent reduction from 2005 levels in economy-wide net greenhouse gas pollution by 2030, creating a carbon pollution-free power sector by 2035, and achieving net zero emissions economy-wide by no later than 2050. Nuclear energy is essential to meeting those goals. That includes the existing fleet and advanced reactors that are moving forward for deployment and under development. Nuclear power plants produce 20 percent of the total electricity supply in the United States today and are the largest source of carbon-free energy. However, we have an opportunity to expand significantly nuclear energy’s percentage of electricity generated in the United States. The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) report - developed in response to President Biden’s Executive Order “America’s Supply Chains” signed in 2021 - describes the nuclear supply chain and investigates challenges for continued operation of today’s 93 reactors and construction of advanced reactors. The supply chain is critical for successfully enabling the continued operation of the existing domestic fleet of light-water reactors as well as supporting deployment of advanced nuclear technologies.

Source record↗

DOE OSTI2023

DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview

Slides discussing the DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview to include focus on LEU UCO TRISO fuel in cylindrical compacts, with data for fuel qualification in support of reactor licensing and a path to establish a domestic commercial TRISO fuel fabrication capability. Timelines from 2004 through 2029. Major program highlights include: Developed fuel performance models based on past observed behavior, Study to understand differences between US and German (“good”) fuel, Developed and improved fuel fabrication and characterization methods, Fabricated lab-scale fuel (AGR-1) with very high quality; excellent in-pile and post-irradiation high-temperature fuel performance, Scale-up coating fabrication (AGR-2) with excellent in-pile performance, AGR-3/4 fuel fabrication and irradiation experiment to assess FP transport behavior, Scale-up matrix fabrication (AGR-5/6/7); & fuel performance evaluation in progress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Advanced Test Reactor Experiment Safety Analysis

This presentation contains general information about the Advanced Test Reactor experiment safety analysis program. Accidents and dose consequence are discussed in general terms. Pictures of ATR facilities and experiments are included. In cases where experiments are shown, identification numbers have been removed. Department of Energy limits on dose consequence are provided. These limits are found in many non-CUI documents and a derived from publicly available sources. A summary of the presentation content is as follows: -Define Experiment Safety Analysis -Discuss what we are protecting and the applicable limits -Show how safety analysis fits into the experiment process -Explore the scope of experiment activities and associated accidents -Present how we categorize accidents and establish accident frequencies -Identify how controls are implemented and process improvements

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Radioisotope production at the advanced test reactor: process and lessons learned

The mission of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) focuses on creating irradiation facilities for nuclear materials and fuels research. While radioisotope production is not a core part of the ATR mission, it is important for providing a U.S. domestic source of critical radioisotopes such as cobalt-60 (Co-60). Here, this paper gives an overview for radioisotopes currently being produced in ATR as well as other potential radioisotopes that can be produced in ATR. It also provides guidance on radioisotope production that can be applied to other test reactors.

ATR↗

DOE OSTI2015

After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2018

NEW MATRIX COEFFICIENTS FOR THE LOBE POWER CALCULATION AND INDICATION SYSTEM IN THE ADVANCED TEST REACTOR

Theoretical coefficients are tuned for use in the Advanced Test Reactor (ATR) Lobe Power Calculation and Indication System (LPCIS). The LPCIS indicates localized fission powers in the ATR, used primarily by operators to provide irradiation of various experiments at requested power levels. The LPCIS solves an overdetermined matrix, such that the best-fit solution can be obtained from at least five of up to eleven independent indications selected for input. Hourly data logs are reviewed for three and one-half years (from November 2013 through June 2017) of ATR operation to compare localized powers indicated by LPCIS with powers indicated by an independent indication, the Water Power Calculator (WPC). WPC indications are used for the calibration of LPCIS because of assumed high reliability and theoretical simplicity. WPC is also used for automatic core protective functions, which LPCIS is not. As the WPC provides power only for four quadrants and the LPCIS indicates power in five lobes, a calculated metric is required to compare the two indications. The metric chosen for the current work has been in use for decades at ATR as a metric of LPCIS health. The proposed new coefficients better fit theory of physical phenomena involved and also provide a 40% improvement in the calculated fit.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2018

Regulatory Gap Analysis of Select NUREG-0800 Chapters for Applicability to Molten Salt Reactors

The past licensing approach for advanced reactors was based on adapting the current LWR-based NRC process that exists today to the advanced reactor concept. This is the approach used by advanced technologies such as the Pebble Bed Reactor licensing application, the Clinch River Breeder Reactor licensing application, and more recently the NuScale licensing application. Current Standard Review Plan (SRP) guidance for large LWRs is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: Light Water Reactor (LWR) Edition. However, the SRP is cumbersome to apply to liquid-fueled molten salt reactor (MSR) technologies. Therefore, a thoughtful licensing approach for MSRs is necessary.This report documents for DOE-NE an additional regulatory gap analysis of select chapters from the LWR SRP to further inform a generic nonpower MSR test program. Although the review process outlined in NUREG-0800 is LWR-specific, the plant functions and processes to be reviewed will be similar for the MSR. Therefore, a well-planned nonpower MSR test program can provide the performance data needed to better facilitate the eventual commercial-sized MSR review. The LWR SRP chapters dealing with the reactor, reactor coolant, engineered safety features, auxiliary systems, and waste management systems were reviewed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

The Fast Modular Reactor (FMR) Pre-Application Regulatory Engagement Plan

The Fast Modular Reactor (FMR) is a 50 MWe Gascooled Fast Reactor (GFR) being developed by General Atomics Electromagnetic Systems (GA-EMS) under U.S. Department of Energy’s (DOE’s) Advanced Reactor Demonstration Program (ARDP), specifically Advanced Reactor Concepts 2020 (ARC-20). The 3-year conceptual design of the FMR is being conducted with verifications of key metrics in fuel, safety, and operational performance. GAEMS is pursuing design, licensing, and commercialization of the proposed reactor, with demonstration by 2030 and deployment by the mid-2030s. The U.S. Nuclear Regulatory Commission (NRC) recommends the pre-application regulatory engagement to provide for early identification of regulatory requirements for advanced reactors and to provide all interested parties with a timely, independent assessment of the safety and security characteristics of advanced reactor designs. As such, GAEMS is developing a pre-application regulatory engagement plan (REP) of the FMR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Extraction of Neptunium, Plutonium, Americium, Zirconium, and Technetium by Di-(2-Ethylhexyl)- Iso -Butyramide (DEH i BA) at High Metal Loadings

Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. Further, by increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. In conclusion, the extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Analysis of Using a Hybrid ZIRCEX Process to Generate High-Assay Low-Enriched Uranium (HALEU) - 20158

Idaho National Lab (INL) is investigating the feasibility of providing High-Assay Low Enriched Uranium (HALEU) by recovering Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) and downblending the HEU with natural, low-enriched, or depleted uranium (NU, LEU, or DU). That investigation identified that significant quantities of HALEU could be produced with a Hybrid ZIRCEX process to further advanced reactor research while reducing INL's environmental liability. Many new advanced reactor designs are being developed with improved safety, efficiency, and economics. Most require nuclear fuel with U-235 enrichment between 5% and 20%, which is defined as HALEU. Although technically feasible, there is no current domestic capability to make HALEU in the US. While it is anticipated that industry will provide HALEU through commercial enrichment once advanced reactors mature, until a market is established, an interim source of HALEU is needed to enable research and demonstration. An interim source of HALEU could be provided by recovering the HEU in some DOE-managed SNF and downblending it to between 5% and 20% using the Hybrid ZIRCEX Process. ZIRCEX is a dry head-end process to remove cladding (zirconium or aluminum) from SNF. Drying of the SNF is needed prior to introduction to ZIRCEX. After cladding removal, uranium and fission products in the bed material are oxidized and elutriated. In a Hybrid ZIRCEX process, the HEU process stream is sent to a very compact, modular solvent extraction process for uranium purification. The uranyl nitrate from solvent extraction is denitrated and calcined to produce a HALEU product ready for fuel fabrication. The long-lived fission products from solvent extraction are immobilized in glass using a small in-can melt. By removing the cladding before dissolution, the volume of glass waste is reduced by a factor of up to 300 times (when compared to not removing the cladding from the fuel). Other radioactive wastes from the process could be disposed of as Class A or Class B low-level waste. In addition to producing HALEU for advanced reactors, consumption of SNF would reduce DoE's environmental liability. Incorporating low-enriched spent fuel (versus exclusive use of LEU, DU, or NU) into the downblending scheme could reduce the SNF inventory even further. Several downblending and enrichment scenarios using INL fuels were investigated in this study. The results of this analysis show that the Hybrid ZIRCEX process could make a significant contribution to needed HALEU feedstock and reduce the environmental liability of managing SNF at INL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

The second and third NGNP advanced gas reactor fuel irradiation experiments

The United States Dept. of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is currently scheduled to irradiate a total of five low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The irradiations are being accomplished to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas cooled reactors. The experiments will each consist of at least six separate capsules, and will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The effluent sweep gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) started irradiation in June 2010, and the third and fourth experiments have been combined into a single larger irradiation (AGR-3/4) that is currently being assembled. The design and status of the second through fourth experiments as well as the irradiation results of the second experiment to date are discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

A review of neutronics and thermal hydraulics–based screening methods applied to accelerated nuclear fuel qualification

This paper reviews the state-of-the-art engineering approach for using thermal hydraulic (TH) and neutronics modeling and simulation (M&S) tools to perform rapid screening studies of novel nuclear fuel concepts within the context of accelerated fuel qualification. Global research efforts have introduced nuclear fuel and material concepts that mark a significant departure from traditional reactor materials. The number of new technologies being considered for development for light water reactors and advanced reactor types has created the need for an accelerated fuel qualification procedure. A key component of this procedure is the rapid identification of the most promising fuel concepts using computational screening studies. Advanced TH and neutronic M&S tools should be leveraged to efficiently determine whether the reactor performance and safety characteristics of a given concept warrant additional studies or whether the concept requires modification or elimination. This paper reviews best practices for performing these TH and neutronics screening studies at various stages during a fuel concept's progression through the qualification procedure. The motivation behind standardizing this approach is to minimize time and resources spent on qualification activities for fuel concepts that could be quickly refined or eliminated from consideration based on their reactor physics and TH characteristics. Adoption of this screening procedure—which focuses primarily on nuclear fuels but may be applicable to other reactor materials—will also help accelerate new material qualification by generating boundary conditions crucial to fuel performance evaluations and highlighting needed areas of separate effects experimentation. This article reviews the motivation behind the introduction of novel nuclear fuel concepts, provides incentive for utilizing TH- and neutronics-based screening studies, describes the screening approach and methodology, and includes discussion on how to interpret screening results to provide recommendations for the continued development of a given concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

At-power subcritical multiplication in the Advanced Test Reactor during nuclear requalification testing

Power division information during nuclear requalification of the Advanced Test Reactor (ATR) is of considerable interest as an importance function for observed changes to core reactivity. The degree to which a given physical subdivision of a critical reactor acts as a neutron source for other lobes is not analytically characterized for general application. When ATR operates at power, individual power-producing lobes rely on each other as neutron sources in order to maintain constant power, which in general requires either exactly critical multiplication within a reactor or an external neutron source. Here, this work shows that fuel element and lobe powers in ATR can be related with subcritical multiplication theory. Subcritical multiplication factors are computed with a physically validated analytical method based on actual at-power operation, quantifying for each lobe its dependence on other lobes as an external neutron source. This explanation is significant for ATR due to the desire to irradiate a large variety of experiments simultaneously, each having its impact on the core neutron population. For any physical subdivision of any other critical reactor, it is likewise true that the subdivision undergoes only subcritical multiplication.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2022

The Regulatory Treatment of Low Frequency External Events as Part of a Risk-Informed, Performance-Based Approach

To assist the developing advanced reactor industry in future licensing efforts, the U.S. Department of Energy Advanced Reactor Demonstration Program Regulatory Development area initiated a project at Argonne National Laboratory to examine the regulatory treatment of external hazards as part of a risk-informed performance-based (RIPB) licensing framework. A RIPB licensing framework for advanced reactors built on establishing an affirmative safety case offers the benefits of increased flexibility regarding key design and licensing decisions based on a detailed assessment and understanding of plant risk. Historically, reactor licensing addressed events of very low frequency primarily through the application of design margin and defense-in-depth philosophy. In contrast, RIPB approaches attempt to evaluate these scenarios at a level of detail commensurate with their risk, which often necessitates an explicit treatment of their frequency and associated consequence. While the detailed analysis of low frequency events provides insights that can help justify alternative treatments to past conservatism, the findings are dependent on the quality and confidence associated with the analyses. The assessment of external hazards presents a unique challenge, as their potential frequency of occurrence, especially of large magnitude events, is inherently uncertain given the long return periods in question. This project aims to identify the benefits and challenges of such approaches for advanced reactor vendors and aid in the development of consistent and appropriate analysis methodologies. The paper summarizes project findings and explores the application of various approaches for different external hazards. In addition, the current work also evaluates the application of the quantitative health objectives as a limit on external event risk, as they are a potential regulatory requirement under the current draft 10 CFR Part 53, which is a new technology-neutral reactor licensing pathway in the U.S.

Grabaskas, David↗

DOE OSTI2024

Meta-Analysis of Advanced Nuclear Reactor Cost Estimations

Supporting Data can be downloaded at: https://gain.inl.gov/content/uploads/4/2024/06/INL-RPT-24-77048-R1.xlsx Nuclear energy is a critical cornerstone of the current United States clean energy supply and may play a larger role in the future in support of a transition to a net-zero economy. The current fleet of nuclear reactors predominantly consists of large light-water reactors (LWRs), while many of the reactor designs under consideration are smaller and/or different technologies. Because these new designs have not yet been built, there is a high degree of uncertainty associated with their cost. This complicates energy-planning efforts because cost projections are not always standardized, consistent, and centralized in an easily accessible location. To help support energy planning in the US, this report provides advanced nuclear cost ranges using a transparent methodology along with other relevant information that can be used to help support decision making and energy planning. The purpose of this work was to conduct a methodical process for cost evaluation using only public information that was vetted with the end-goal to provide reference cost projections for nuclear energy. To provide a solid basis for these values, the approach and assumptions are explicitly laid out throughout the report allowing any user of the data to challenge or reconsider them. Because future US nuclear-reactor costs are still unknown due to little recent observed data, the report opted to compile a comprehensive list of bottom-up estimates and evaluate averages/trends within the data to identify reference ranges. This was deemed preferable to opining on the robustness or validity of one cost estimation versus another. To that end, the work evaluated thousands of lines of cost subaccounts from several bottom-up cost estimates. A wide variety of different reactor types captured in the data are of various sizes and technologies. Some of these reactors will be representative of advanced reactors under development while others will not. Thus, the results here are dependent on the data that are available and the accuracy of the estimates that are used. Each bottom-up estimate was reviewed to determine whether it was complete. Incomplete data sets were corrected to ensure an adequate basis of cross-comparison. The report is not without limitations and should be interpreted as an initial step to develop cost ranges for nuclear technology. Ultimately, future work can build upon the methodology with refined cost estimates to reduce uncertainty. US-based overnight capital cost (OCC) estimates were compiled from extensive data sets into ranges for both large and small reactor sizes for 2030. To project the cost declines over time, learning rates were sampled from literature sources. No SMRs were previously built; hence, learning rates based on bottom-up approaches (e.g., by quantifying the impact stemming from fabrication of different components, modular work, site construction, commissioning) were prioritized. For larger reactors, actual learning rates from deployments were used to project future costs (adjusted to account for standardization or lack thereof between designs). Other costs included are fixed and variable operations and maintenance costs. The final variables were capacity factors and ramp rates to support energy planning.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-en-riched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and pre-dictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Advanced Test Reactor Long-Term Asset Management Accomplishments 2016 to 2022

Advanced Test Reactor Long-Term Asset Management Accomplishments report contains photographs and information on completed projects or completed phase of a project for 2022. This book will be presented for the year-end report in Washington, D.C. as well as delivered to our clients at the Navy Nuclear Laboratory.

99 GENERAL AND MISCELLANEOUS↗