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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 OSTI2019

Shortening the Design and Certification Cycle for Additively Manufactured Materials by Improved Mesoscale Simulations and Validation Experiments: Fiscal Year 2019 Status Report.

This report outlines the fiscal year (FY) 2019 status of an ongoing multi-year effort to develop a general, microstructurally-aware, continuum-level model for representing the dynamic response of material with complex microstructures. This work has focused on accurately representing the response of both conventionally wrought processed and additively manufactured (AM) 304L stainless steel (SS) as a test case. Additive manufacturing, or 3D printing, is an emerging technology capable of enabling shortened design and certification cycles for stockpile components through rapid prototyping. However, there is not an understanding of how the complex and unique microstructures of AM materials affect their mechanical response at high strain rates. To achieve our project goal, an upscaling technique was developed to bridge the gap between the microstructural and continuum scales to represent AM microstructures on a Finite Element (FE) mesh. This process involves the simulations of the additive process using the Sandia developed kinetic Monte Carlo (KMC) code SPPARKS. These SPPARKS microstructures are characterized using clustering algorithms from machine learning and used to populate the quadrature points of a FE mesh. Additionally, a spall kinetic model (SKM) was developed to more accurately represent the dynamic failure of AM materials. Validation experiments were performed using both pulsed power machines and projectile launchers. These experiments have provided equation of state (EOS) and flow strength measurements of both wrought and AM 304L SS to above Mbar pressures. In some experiments, multi-point interferometry was used to quantify the variation is observed material response of the AM 304L SS. Analysis of these experiments is ongoing, but preliminary comparisons of our upscaling technique and SKM to experimental data were performed as a validation exercise. Moving forward, this project will advance and further validate our computational framework, using advanced theory and additional high-fidelity experiments.

36 MATERIALS SCIENCE↗
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 OSTI2020

Risk Informed, Performance-Based, Technology-Inclusive Regulatory Infrastructure: Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities

This report summarizes a risk informed, performance-based and technology-inclusive approach to determine source terms for dose related assessments at advanced nuclear facilities. This approach uses a graded process which allows both the non-mechanistic source terms calculation methods, which adopt conservative approaches and assumptions based on known physical and chemical principles, and more importantly the mechanistic source term calculation methods, which consider design-specific scenarios and use best estimate models with uncertainty quantification for a range of licensing basis events (LBEs), to be used for the design and licensing of advanced nuclear technologies. The source terms developed with this graded approach, and radionuclide inventories elsewhere in the facility that are determined during source term analysis, can be used to address licensing issues to support the application processes of 10 CFR Part 50 for a construction permit and operating license or 10 CFR Part 52 for a Combined Operating License (COL), Standard Design Certification, Early Site Permit, Standard Design Approval or Manufacturing License. They can also be used for other purposes including equipment environmental qualification, control room habitability analyses, and assessments of severe accident risks in environmental impact statements. There are many advanced reactor concepts being developed including high temperature gas-cooled reactor (HGTR), sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), micro-reactor, etc. The graded approach presented in this report for source terms determination is, to the extent possible, generic to any of these reactor designs and to future reactor designs. This report provides information on the review of the regulatory foundation for use of conservative bounding source terms as well as event-specific mechanistic source terms for advanced nuclear reactor designs.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Unique safety features and licensing requirements of the NuScale small modular reactor

Small modular reactors (SMR) offer a novel approach to the construction and operation of nuclear power plants. The NuScale VOYGR™ plant uses a simplified SMR design that is based on proven light-water reactor technology with substantial improvements in nuclear safety. It consists of a 250 MWt reactor core housed with other primary system components in an integral reactor pressure vessel surrounded by a steel containment vessel, all of which is immersed in a large pool of water that also serves as the ultimate heat sink. At the core of the NuScale safety case are three primary safety systems: the decay heat removal system, the emergency core cooling system, and the containment. The ability of the NuScale Power Module (NPM) passive safety systems to remove core decay heat for an unlimited duration is demonstrated through analysis of a beyond-design-basis extended loss of AC power with no replenishment of water to the ultimate heat sink or operator actions. The NuScale methodology to evaluate an indefinite loss of AC power uses the proprietary NRELAP5 systems analysis computer code. Analysis results show that the reactor coolant system liquid level above the core is maintained and that containment pressure remains below the vessel design pressure. Once full passive air cooling is established, containment pressure and temperature will decrease over time with decreasing core decay heat. NuScale received standard design approval in September 2020 and design certification in January 2023 for its 50 MWe NPM configured as a 12 module plant. NuScale is currently seeking standard design approval to increase its core power to 250 MWt, nominally 77 MWe per module, in a 6-module plant (VOYGR™-6) configuration. The high-level safety of NuScale’s SMR technology is foundational to a new standard of nuclear power plant resilience.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Migration to a fully-integrated control room - 349

The nuclear industry is faced with a pressing need to reduce operating costs in order to remain competitive with other forms of electric generation. While current industry initiatives are focused on incremental gains by streamlining work processes and reducing worker burdens, very few, if any, involve significant reductions in the base workload of a nuclear plant. Migration to a fully-integrated control room affords just such an opportunity. This report describes an NPP digital instrumentation and control (I and C) system design, along with its corresponding human systems interface (HSI) design, which achieves high plant availability, while selectively applying digital equipment redundancy and integrating traditional I and C system partitions, to minimize the digital equipment needed, and thereby minimize initial installation costs and ongoing operations and maintenance (O and M) costs. An additional key component of digital equipment minimization is optimization of the digital interface to conventional plant sensors and controlled plant devices (e.g., pumps, valves), which requires no disturbance to plant cables and results in eliminating thousands of electro-mechanical relay control components. The design is rooted in the highly integrated digital technology applied to the most recent advanced light water reactors under United States (U.S.) Nuclear Regulatory Commission (NRC) design certification review and approval. While this design employs technology from new plants, that technology has been adapted to the modernization of current operating NPPs. Hence, the highly integrated digital I and C and HSI design is referred to as the compact control room digital modernization (CDM). Along with minimal digital equipment, the corresponding HSI design employs compact operator workstations with visual display units (VDU) and soft controls for both the main control room (MCR) and remote shutdown room (RSR), to immediately improve human performance and ultimately facilitate operating staff reduction. Some advanced light water reactors have demonstrated this staff reduction through full scope simulator testing of normal and abnormal plant conditions, with U.S. licensed operating crews from currently-operating U.S. plants. Digital equipment minimization, compact operator workstations, and optimized plant component interfaces that require no disturbance to field cables, also facilitate implementation of the CDM over multiple phases, most of which can be conducted during normal refueling shutdown windows. This phased implementation method includes a novel transition of the I and C, MCR and RSR with an operations centered focus. This paper presents the CDM design, along with qualitative cost/benefit comparisons to currently operating plants with conventional analog I and C and HSI technology; this generation of I and C/HSI is referred to as a conventional control room (CCR). A CCR also includes NPPs that have been upgraded with digital systems, but retain the current partitions between plant I and C systems (i.e., no I and C integration) and retain the current plant system to control board configuration of current NPP control rooms. Even though these comparisons are qualitative, a compelling case is presented for the CDM. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

2021-2026 Weapons Research Services Strategic Plan

The Weapons Research Services (WRS) Division provides Los Alamos National Laboratory with services, resources, and staff that are critical to fulfilling its national security mission. WRS serves the entire LANL Weapons Program and is housed within the Weapons Physics Directorate (ALDX), which is at the core of the design, certification, and assessment of the nation’s nuclear weapons. More specifically, WRS is a research and development organization that provides unique tools designed to meet the specific needs of the Lab’s weapons mission rather than administrative or Information & Technology (IT) tools used broadly by the Lab. WRS partners with other LANL organizations, including Global Security to conduct mission research and I&T organizations to meet basic IT infrastructure needs.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2018

Progress of full-scale MCNP6{sup TM} model development of NuScale small modular reactor - 25394

This paper presents the recent progress of development of full-scale MCNP6{sup TM} small modular reactor within NuScale Power LLC (NuScale) using MCNP6{sup TM} version 1.0. This model currently is being used to determine of vessel neutron fluence, nuclear damage, and radiological consequence analysis. The results of the best estimate fluence analysis to support Design Certification Application (DCA) and technical details has been included within License Technical Report (LTR) which had been submitted to US NRC for review in 2017. Currently the detailed analysis of using this full-scale NuScale SMR model is focusing on operational radiological external dose exposure, nuclear damage, nuclear heating and neutron activation of SMR components to support optimized design and RAI (request additional information) response activities. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Method advancements for fuel assessments during seismic and LOCA events

The evaluation of fuel assembly structural responses of light-water reactors to seismic and LOCA excitations has received increased attention in the last five years. Real seismic events such as those that affected the Fukushima Daiichi and North Anna plants in 2011 have caused the industry to more closely examine our preparedness for seismic events. Furthermore, the review of design certifications for new plants has raised questions about the methods used to evaluate these events and is leading to revised guidance from regulators (e.g. US NRC Information Notice 2012-09). As a result, the industry is seeing fuel structural design margins based on current methods begin to erode for seismic and LOCA events. This paper will review points for consideration as the industry responds to these changes. To begin, it is worthwhile for the industry to understand the history and the critical thought behind the current methods such that improved criteria and methods can be formulated. This review identifies three known areas of weakness in the current framework for this methodology. First, a broader understanding of spacer grid behavior is needed. Spacer grid behavior in response to impact loads is of paramount importance, but the industry needs to re-frame our characterization of this behavior in order to focus on the actual critical safety function of this component. Secondly, despite guidance from the Standard Review Plan, the effects of irradiation, on spacer grid characteristics, in particular, must be considered. Thirdly, the area of fluid-structure-interaction may be the greatest frontier to conquer in this methodology and may provide our best possibility to recover eroded margins. However, fluid-structure-interactions must be cautiously evaluated and applied with conservatism. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Integrating quantitative defense-in-depth metrics into new reactor designs

Risk-informed, performance-based (RIPB) methods have progressed to the point where high-level guidance can be used to augment traditional, deterministic, nuclear safety design practices in areas important to nuclear reactor safety. This paper describes an approach for augmenting the traditional defense-in-depth (DID) qualitative approach with quantitative risk information from a plant-specific probabilistic risk assessment (PRA) in a way that is structured, can be applied on a consistent basis, and allows for clear acceptance criteria. Adding performance-based targets that should be achieved is expected to result in safer and more economical plant designs. Evaluations of DID can be conducted throughout the design process as well as in support of design certification and operating license applications to identify where defense protections could be enhanced or relaxed. Consistent with the United States Nuclear Regulatory Commission's policy statement encouraging greater use of PRA to improve safety decision making and regulatory efficiency, this scenario-based DID method can be used to evaluate changes and overall plant design as part of the normal design control process. Although the RIPB method presented in this paper was developed for application to advanced passive light water reactor designs, the metrics could be tailored to other reactor designs. This risk-informed approach to DID helps to ensure that public and worker risk insights are integrated into the design process holistically.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗