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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

lmprovements and Path Forward for Regulatory Acceptance of SAS4A/SASSYS-1

In the U.S., a key component of the commercialization of advanced reactors is completion of a license application, which must ultimately be approved by the Nuclear Regulatory Commission (NRC). The NRC’s approval of the license application is contingent on, among other things, satisfactory demonstration of the design basis and response to transient and accident scenarios using accepted codes and methods. This effort seeks to improve the regulatory acceptability of the SAS4A/SASSYS-1 advanced reactor design and safety analysis system software by identifying and addressing gaps in the SAS4A/SASSYS-1 documentation basis that support software qualification and dedication. The software’s unique capabilities to assess inherent fast spectrum feedback effects and passive safety features, which facilitate assessment of key safety metrics, position it as a key licensing tool in the field of sodium-cooled fast reactors (SFR).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Risk Informed Design and Analysis Criteria for Nuclear Structures

Target performance can be achieved by defining design basis ground motion from results of a probabilistic seismic hazards assessment, and introducing known levels of conservatism in the design above the DBE. ASCE 4, 43, DOE-STD-1020 defined the DBE at 4x10-4 and introduce only slight levels of conservatism in response. ASCE 4, 43, DOE-STD-1020 assume code capacities shoot for about 98% NEP. There is a need to have a uniform target (98% NEP) for code developers (ACI, AISC, etc.) to aim for. In considering strengthening options, one must also consider cost/risk reduction achieved.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Steam Oxidation, Burst and Critical Heat Flux Testing of FeCrAl Cladding in the Severe Accident Test Station

Since August 2011, Oak Ridge National Laboratory (ORNL) has been evaluating the steam oxidation resistance of a wide range of materials to identify a potential accident tolerant fuel (ATF) cladding [1-14]. The primary goal of an ATF cladding is to enhance safety margins in light water reactors (LWR) during beyond design basis accident scenarios (such as the nuclear accident at Fukushima Daiichi in March 2011) without requiring changes to the reactor design or fuel enrichment, or significantly affecting reactor economics during normal operation.

36 MATERIALS SCIENCE↗
NASA NTRS2018

Vehicle Design Data Format and Process for a Complete TARIS and OLTARIS Radiation Analysis for Designers and Engineers

Protecting astronauts from space radiation is a complex task when it comes to modeling and simulation. This document shows what information is needed from a spacecraft designer using CAD (Computer-Assisted Design) at each phase of the design to enable the engineers to evaluate the design phase against space radiation limits to determine the suitability of the design for space flight. The current personal exposure limits are listed in NASA STD-3001. A proxy to determine the REID (Radiation Exposure Induced Death) in NASA STD-3001 is the whole body effective dose equivalent (E or effective dose). For short-term tissue effects, organ-averaged gray equivalent (G (sub T)) is used. The TARIS (Tool for the Assessment of Radiation In Space - for LaRC (Langley Research Center) engineers) and OLTARIS (On-Line TARIS) - for designers) systems are used to generate these response functions. The E can use ICRP60 or NASA Q-values. A possible space radiation design basis environment for short-term tissue effects is described and used in all analyses. A single space vehicle was designed with three astronaut configurations and two of those configurations were used in a storm shelter thickness perturbation analysis. Conversion of the data from the CAD model to input necessary for TARIS and OLTARIS is also discussed in detail with relevant examples.

Singleterry, Robert C.↗
DOE OSTI2015

Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate operation of systems and components important to safety as required in GDC 20. This paper provides an overview of the design process employed to develop a pre-conceptual FHR instrumentation architecture intended to lower plant capital and operational costs by minimizing reliance on expensive, safety related, safety-significant instrumentation through the use of inherent passive features of FHRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Gasification Combined Heat and Power from Coal Fines

The goal of this final project report is to comprehensively summarize the work for public domain conducted on project DE-FE 0031520. In accordance with the Statement of Project Objectives (SOPO), the University of Kentucky Center for Applied Energy Research (UKy-CAER) (Recipient) has completed a Front End Engineering Design (FEED) study 5 MWe polygeneration unit using the Coal Fields Regional Industrial Park in Hazard, Kentucky (KY) as a model location and using coal recovered from the adjacent Pads Branch Impoundment as feed. The scope of work is a thorough study that can be used as a template for the application of the proposed technology throughout Eastern KY and other coal communities nationwide. The Technology Design Basis was developed based on three elements: basic site characteristics; fuel feedstock; and environmental health and safety requirements. The Basic Engineering Design Elements include the preliminary process design (the process area descriptions; the Block Flow Diagram (BFD), Process Flow Diagrams (PFDs) and Piping and Instrumentation Diagrams (P&IDs), process simulation outputs and Heat and Mass Balance (H&MB) stream tables, the major process equipment and instrumentation lists which includes specifications and materials of construction); the cost estimate (capital cost estimate, operational cost estimate, and comparison with state-of-the-art commercial scale similar technologies, options to lower cost, construction cost and schedule estimate); and other supporting elements (guidelines for execution and management of the project), logistical summary, and balance of plant summary, the Environmental Health and Safety (EH&S) assessment; and a Social and Economic Impact (S&EI) assessment. All project milestones, deliverables and success criteria were met. The proposed 5 MWe polygeneration facility being fed from impoundment recovered coal has been shown to be technically feasible. Annual revenues are estimated at $\$ $2.4M from electricity, steam and hydrocarbon production; annual operating expenses are $\$ $5.2M ($\$ $2.9 M in operating expenses, and $\$ $2.4M in indirect expenses). The gasifier facility, as configured during the FEED, will lose money when operating under commercial operating conditions and terms. However, eight options improving the economics can be considered during the detailed design phase to make the facility profitable by taking advantage of the positive differential of revenue to variable operating costs, and reducing the influence of fixed operating costs. Coupled with the findings from the S&EI assessment estimating a total economic impact of $\$ $374M for the proposed polygeneration facility at >$\$ $1.5B, if three additional such projects are implemented, clearly positive economics are demonstrated. No EH&S issues are identified that would prohibit proceeding with the project, although permitting efforts are estimated to take 3-5 years.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2017

Transient Mechanical Test Development for Accident Tolerant Cladding Candidates to Simulate PCMI-like Conditions

This study investigates the transient mechanical response of accident tolerant fuel (ATF) cladding candidates to provide information on potential PCMI-induced deformation and failure of candidate cladding materials. The mechanical testing is applied t to Gen-I FeCrAl and triaxially-braided SiC fiber SiC matrix composite materials under loading conditions like the PCMI during a postulated design basis reactivity insertion accident. To achieve the specific loading conditions, a modified burst test instrument with accurate mechanical pulse control has been designed and constructed. The mechanical response of FeCrAl, SiC-SiC composite tubes were compared with the pre-hydrided zirconium alloy cladding. The specimens subjected to mechanical pulses were investigated in a strain-controlled manner deformation.

Cinbiz, Nedim↗
DOE OSTI2013

Preliminary safety evaluation of an aircraft impact on a near-surface radioactive waste repository

The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a 'beyond design basis' event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a 'robust' target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircraft, even in presence of ongoing concrete progressive failure (some penetration and spalling of the concrete wall) of the impacted area. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1982

Technical and economic assessment of processes for the production of butanol and acetone

This report represents a preliminary technical and economic evaluation of a process which produces mixed solvents (butaol/acetone/ethanol) via fermentation of sugars derived from renewable biomass resources. The objective is to assess the technology of producing butanol/acetone from biomass, and select a viable process capable of serving as a base case model for technical and economic analysis. It is anticipated that the base case process developed herein can then be used as the basis for subsequent studies concerning biomass conversion processes capable of producing a wide range of chemicals. The general criteria utilized in determining the design basis for the process are profit potential and non-renewable energy displacement potential. The feedstock chosen, aspen wood, was selected from a number of potential renewable biomass resources as the most readily available in the United States and for its relatively large potential for producing reducing sugars.

Source record↗
DOE OSTI2013

Inhibition of Oxidation in Nuclear Graphite

Graphite is a fundamental material of high temperature gas cooled nuclear reactors, providing both structure and neutron moderation. Its high thermal conductivity, chemical inertness, thermal heat capacity, and high thermal structural stability under normal and off normal conditions contribute to the inherent safety of these reactor designs. One of the primary safety issues for a high temperature graphite reactor core is the possibility of rapid oxidation of the carbon structure during an off normal design basis event where an oxidizing atmosphere (air ingress) can be introduced to the hot core. Although the current Generation IV high temperature reactor designs attempt to mitigate any damage caused by a postualed air ingress event, the use of graphite components that inhibit oxidation is a logical step to increase the safety of these reactors. Recent experimental studies of graphite containing between 5.5 and 7 wt% boron carbide (B4C) indicate that oxidation is dramatically reduced even at prolonged exposures at temperatures up to 900°C. The proposed addition of B4C to graphite components in the nuclear core would necessarily be enriched in B-11 isotope in order to minimize B-10 neutron absorption and graphite swelling. The enriched boron can be added to the graphite during billet fabrication. Experimental oxidation rate results and potential applications for borated graphite in nuclear reactor components will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Development of the reactor core isolation cooling system model for the extended station black-out accident

In order to accurately predict the behavior of the Reactor Core Isolation Cooling System (RCIC) in the context of the FLEX strategies under the extended station black-out accident (SBO), we present a mechanistic model of the RCIC in this work. The control volume formulation of angular momentum conservation is employed to model the Terry turbine in the RCIC system. The two-phase nozzle flow model developed in the previous work is utilized to simulate the convergent-divergent nozzles in the Terry turbine. A nozzle flow table is generated to cover the operation range of the RCIC during beyond design basis accidents (BDBAs). The mechanistic RCIC turbine-pump model and the nozzle flow table are incorporated in a TRACE BWR model. Simulations of short-term and long-term SBO transients are performed to demonstrate dynamic RCIC responses. A system analysis code with the proposed mechanistic model of the RCIC system will have a better prediction of system safety response during SBO accidents. With the addition of such more realistic models, the system code will be capable of providing expanded understanding of safety margin, which can help help to form the technical basis for the successful implementation of the FLEX and SAMG measures under Extended Loss of AC Power (ELAP) conditions for BWR or PWR reactor designs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Degradation and Failure Phenomena of Accident Tolerant Fuel Concepts: FeCrAl Alloy Cladding

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of accident tolerant fuel (ATF) in United States commercial nuclear power reactors. Pacific Northwest National Laboratory is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs. This report focuses specifically on the iron-chromium-aluminum (FeCrAl) alloys being investigated to replace zirconium-based alloys for fuel cladding and provides current state-of-the-industry information on material properties and fuel performance considerations under reactor operating conditions and design basis accident conditions. Currently, Global Nuclear Fuels (GNF) is the only U.S.-based fuel vendor considering FeCrAl cladding as a near-term ATF design. GNF has tested several different FeCrAl alloys including Kanthal APMT, C26M, and MA956; fueled C26M rods and unfueled Kanthal APMT, C26M, and MA956 rods have been inserted in lead test assemblies at Plants Hatch and Clinton. This report will focus on those three alloys specifically, with broader information given regarding in-reactor and ex-reactor testing of FeCrAl alloys. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of FeCrAl cladding concepts, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Coupling of Pronghorn and RELAP-7 for a Pebble Bed Reactor

High temperature gas cooled reactors (HTGR) are a candidate for timely Gen-IV reactor technology deployment because of high technology readiness and walk-away safety. Among HTGRs, pebble bed reactors (PBRs) have attractive features such as low excess reactivity and online refueling. Pebble bed reactors pose unique challenges to analysts and reactor designers such as continuous burnup distribution depending on pebble motion and recirculation, radiative heat transfer across a variety of gas-filled gaps, and long design basis transients such as pressurized and depressurized loss of forced circulation. Modeling and simulation is essential for both the PBR’s safety case and design process. In order to verify and validate the new generation codes the Nuclear Energy Agency (NEA) Data bank provide a set of benchmarks data together with solutions calculated by the participants using the state of the art codes of that time. An important milestone to test the new PBR simulation codes is the OECD NEA PBMR-400 benchmark which includes thermal hydraulic and neutron kinetic standalone exercises as well as coupled exercises and transients scenarios. In this work, the reactor multiphysics code MAMMOTH and the thermal hydraulics code Pronghorn, both developed by the Idaho National Laboratory (INL) within the multiphysics object-oriented simulation environment (MOOSE), have been used to solve Phase 1 exercises 1 and 2 of the PBMR-400 benchmark. The steady state results are in agreement with the other participants’ solutions demonstrating the adequacy of MAMMOTH and Pronghorn for simulating PBRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1991

A mathematical basis for the design and design optimization of adaptive trusses in precision control

A mathematical basis for the optimal design of adaptive trusses to be used in supporting precision equipment is provided. The general theory of adaptive structures is introduced, and the global optimization problem of placing a limited number, q, of actuators, so as to maximally achieve precision control and provide prestress, is stated. Two serialized optimization problems, namely, optimal actuator placement for prestress and optimal actuator placement for precision control, are addressed. In the case of prestressing, the computation of a 'desired' prestress is discussed, the interaction between actuators and redundants in conveying the prestress is shown in its mathematical form, and a methodology for arriving at the optimal placement of actuators and additional redundants is discussed. With regard to precision control, an optimal placement scheme (for q actuators) for maximum 'authority' over the precision points is suggested. The results of the two serialized optimization problems are combined to give a suboptimal solution to the global optimization problem. A method for improving this suboptimal actuator placement scheme by iteration is presented.

Das, S. K.↗
NASA NTRS1992

Thermostructural responses of carbon phenolics in a restrained thermal growth test

The thermostructural response of carbon phenolic components in a solid rocket motor (SRM) is a complex process. It involves simultaneous heat and mass transfer along with chemical reactions in a multiphase system with time-dependent material properties and boundary conditions. In contrast to metals, the fracture of fiber-reinforced composites is characterized by the initiation and progression of multiple failures of different modes such as matrix cracks, interfacial debonding, fiber breaks, and delamination. The investigation of thermostructural responses of SRM carbon phenolics is further complicated by different failure modes under static and dynamic load applications. Historically, there have been several types of post-firing anomalies found in the carbon phenolic composites of the Space Shuttle SRM nozzle. Three major failure modes which have been observed on SRM nozzles are pocketing (spallation), ply-lift, and wedge-out. In order to efficiently control these anomalous phenomena, an investigation of fracture mechanisms under NASA/MSFC RSRM (Redesigned Solid Rocket Motor) and SPIP (Solid Propulsion Integrity Program) programs have been conducted following each anomaly. This report reviews the current progress in understanding the effects of the thermostructural behavior of carbon phenolics on the failure mechanisms of the SRM nozzle. A literature search was conducted and a technical bibliography was developed to support consolidation and assimilation of learning from the RSRM and SPIP investigation efforts. Another important objective of this report is to present a knowledge-based design basis for carbon phenolics that combines the analyses of thermochemical decomposition, pore pressure stresses, and thermostructural properties. Possible areas of application of the knowledge-based design include critical material properties development, nozzle component design, and SRM materials control.

Wang, C. Jeff↗
NASA NTRS2022

Evaluation of Multiple Methods for Calculating Gray Equivalent

The assessment of different algorithms to determine the Gray-Equivalent as defined in NASA-STD-3001 from a source of space radiation has been evaluated in this paper. The Gray-Equivalent applies an RBE (Relative Biological Effectiveness) to the dose seen at the organ of interest in a human phantom. The current design basis solar particle event was used in the assessments along with the August 1972 event modeled by J.H. King and with idealized spheres and two vehicle designs. Three different algorithms were used and compared. One of the algorithms was the current OLTARIS algorithm. This algorithm is astronaut orientation averaged, which is not what happens in a storm shelter during a solar particle event. Two other algorithms were proposed to eliminate this issue. It is clear that either algorithm will be adequate to satisfy NASA-STD-3001 requirements. This work recommends the algorithm which applies the RBEs to the phantom points instead of at the surface of the phantom as with the current OLTARIS algorithm.

Gray Equivalent↗
DOE OSTI2020

DEMONSTRATION OF MULTI-PHYSICS EVALUATION OF FUEL ROD BURST PROBABILITY WITH BURNUP EXTENSION UNDER LBLOCA CONDITIONS

Extending the fuel discharge burnup level, e.g., from the current limit of rod averaged discharge burnup limit of 62 GWD/MT to a proposed new limit of 75 GWD/MT, can provide significant economic benefits to the current fleet of operating light water reactors (LWRs). It allows for longer operating cycles and improved resource utilization. The major economic gain of longer operating cycles is attributable to the increased capacity factor resulting from decreased refueling time as a fraction of total operating time, as well as fewer assemblies to be discharged for a given amount of energy produced. The main licensing challenges for higher burnup fuel are to ensure fuel rod safety under design basis accident conditions, especially under large-break loss-of-coolant accident (LBLOCA) and reactivity insertion accident (RIA). In this work, two-year cycle core design for a typical 4-loop pressurized water reactor (PWR) is performed with enrichment increased up to 6% and burnup extended to 75 GWD/MT. The fuel rod burst potential evaluations under large-break loss-of-coolant accident (LBLOCA) conditions are subsequently performed using the multi-physics best estimate plus uncertainty analysis framework LOTUS (LOCA Toolkit for the U.S. LWRs) and the preliminary results are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

Progress on the NSTX Center Stack Upgrade

The National Spherical Torus Experiment (NSTX) will be upgraded to provide increased toroidal field, plasma current and pulse length. This involves the replacement of the so-called center stack, including the inner legs of the Toroidal Field (TF) coil, the Ohmic Heating (OH) coil, and the inner Poloidal Field (PF) coils. In addition the increased performance of the upgrade requires qualification of remaining existing components for higher loads. Initial conceptual design efforts were based on worst-case combinations of possible currents that the power supplies could deliver. This proved to be an onerous requirement and caused many of the outer coils support structures to require costly heavy reinforcement. This has led to the planned implementation of a Digital Coil Protection System (DCPS) to reduce design-basis loads to levels that are more realistic and manageable. As a minimum, all components must be qualified for the increase in normal operating loads with headroom. Design features and analysis efforts needed to meet the upgrade loading are discussed. Mission and features of the DCPS are presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗