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Results for “Fuel performance analysis”

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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Preliminary analysis of FCM fuel thermal-mechanical performance under normal and transient conditions

Fully ceramic microencapsulated (FCM) fuel is a candidate fuel form to replace commercial UO{sub 2} fuel rod in LWR to enhance fuel reliability under accident conditions. The FCM pellet consists of TRISO particles embedded inside the dense SiC matrix. This fuel design is more complex than conventional LWR fuel, which needs new fuel performance models or analysis tools to assist fuel design and concept evaluation. In this paper an attempt was made to develop a finite element method to analyze FCM fuel pellet thermal-mechanical performance. Three idealized cases of normal operation, power ramp, and LOCA were analyzed. The results showed that the temperature distribution and stress distribution of FCM pellet exhibited a strong non-uniformity. The irradiation strain and creep of buffer layer and PyC (pyrolytic carbon) layers had a strong effect on the mechanical performance of FCM pellet. The hoop stresses of SiC layers and SiC matrix increased significantly during the LOCA and varied with radial positions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Micro heat pipe nuclear reactor concepts: Analysis of fuel cycle performance and environmental impacts

The demand for more practical and innovative nuclear reactor designs capable of providing clean energy to meet new global demands is revolutionizing the nuclear industry. Here, we present the fuel cycle and neutronics analysis of a design expected to be similar to the Westinghouse eVinci™ heat pipe (HP) reactor. The concept considered here uses low enriched urania rods and potassium liquid metal-cooled HPs to remove heat from the core. The fuel and HPs are contained in steel monolith to form the core which is surrounded by an alumina reflector. There is no need for forced circulation to remove heat from the active core region, thus eliminating the use of pumps, valves and tube piping. The concept is designed to operate safely and provide a reliable autonomous power supply in support of off-grid missions. This study shows that the HP reactor design can operate for more than 10 years without refueling. Further, we show that the current design of the HP reactor concept is best suited to serve as a nuclear battery rather than a centralized power source. We also studied the nuclear fuel cycle performance of the HP reactor concept in a once-through fuel cycle. The natural resource utilization, waste output, and environmental impact, as defined in a reference study from the literature, did not perform as well on a GWe per year energy basis compared to light water reactors with less than 5% enriched uranium in a once-through fuel cycle. Parasitic absorption in the steel monolith structure and neutron leakage lower the reactivity of the core, which decreases the discharge burnup of the fuel. Finally, analysis of modifications of the monolith material and size of the reference core configuration showed that the neutron leakage impact on the small sized HP core is the most limiting factor on the fuel cycle performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1984

Group-type hydrocarbon standards for high-performance liquid chromatographic analysis of middistillate fuels

A new high-performance liquid chromatographic (HPLC) method for group-type analysis of middistillate fuels is described. It uses a refractive index detector and standards that are prepared by reacting a portion of the fuel sample with sulfuric acid. A complete analysis of a middistillate fuel for saturates and aromatics (including the preparation of the standard) requires about 15 min if standards for several fuels are prepared simultaneously. From model fuel studies, the method was found to be accurate to within 0.4 vol% saturates or aromatics, and provides a precision of + or - 0.4 vol%. Olefin determinations require an additional 15 min of analysis time. However, this determination is needed only for those fuels displaying a significant olefin response at 200 nm (obtained routinely during the saturated/aromatics analysis procedure). The olefin determination uses the responses of the olefins and the corresponding saturates, as well as the average value of their refractive index sensitivity ratios (1.1). Studied indicated that, although the relative error in the olefins result could reach 10 percent by using this average sensitivity ratio, it was 5 percent for the fuels used in this study. Olefin concentrations as low as 0.1 vol% have been determined using this method.

Otterson, D. A.↗
DOE OSTI2016

Treatment of the uncertainties in fuel rod design analysis

This paper describes the results of the effort performed in order to systematically treat uncertainties in the fuel rod performance analysis with the aim at providing conservative results while also showing the uncertainty of the results (Best Estimate plus Uncertainty). 3 distinct groups of uncertainties affect the prediction of the fuel performance code: -) code models and material properties, -) fuel rod as fabricated parameters, and -) core power and power distribution determination. While the first two groups may be easily covered by statistical analysis, the last group requires good knowledge of the origin of the uncertainties. Especially in tall PWR and VVER-1000 cores where the possibility of fuel assembly bowing affects the inter-assembly gap width generating an uncertainty. In general, this parameter evolves during the residence of the fuel assembly in the core. Several methodological approaches to the core power and power distribution are compared using the TRANSURANUS code. TRANSURANUS is an integral '1.5D' code assuming perfect cylindrical symmetry of the fuel rod and only limited coupling in the axial direction. Core neutron transport, including pin-wise power distributions, are calculated by the nodal code ANDREA. As an example of a practical application, the study of the fuel rod response to the VVER-1000 Condition 2 transient (unintentional dilution of the boric acid in the primary circuit) is shown as well as an example of application of the same statistical approach to the validation of the TRANSURANUS code using the in-pile measurements in the Halden reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Status of Multiple Channel Fuel Performance Capabilities Within the SAS4A/SASSYS-1 Safety Analysis Software

SAS4A/SASSYS-1 (SAS) is a fast-running simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. It is a critical element of safety analysis capabilities for the U.S. Department of Energy and is utilized within industry to perform the transient safety analyses required to support the licensing of Liquid Metal-cooled Fast Reactors (LMFRs). Although SAS is exceptionally fast for most transient scenarios, fuel performance calculations, along with the associated pre-transient characterization of the fuel pin, may be required for transient scenarios where fuel pin failure is hypothesized. Both the pre-transient characterization and the transient fuel performance calculation are necessary to properly quantify margins to potential fuel failure and assess the time spent potentially exceeding such margins during events. While safety analysis calculations with fuel performance models provide a more detailed characterization of the reactor during a transient, the pre-transient characterization can be time-consuming and computationally expensive. Often, large numbers of fuel pins have been exposed to similar pre-transient irradiation conditions. Similarly, the same pre-transient fuel characterization may be applicable to numerous transient conditions. This provides an opportunity to optimize the SAS computational framework such that pre-transient fuel characterization can be shared across multiple channels (fuel pins) and across multiple simulations, thus dramatically reducing overall computational costs. This report summarizes progress toward enhancing the SAS computational framework to support shared, multiple channel fuel performance characterizations intended to significantly reduce computational costs. Preliminary testing has shown that the computational time saved by using the pre-transient sharing capability is approximately equal to the time it takes to perform the pre-transient characterization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Fuel performance uncertainty quantification and sensitivity analysis in the presence of epistemic and aleatoric sources of uncertainties

Fuel performance modeling and simulation includes many uncertain parameters from models to boundary conditions, manufacturing parameters and material properties. These parameters exhibit large uncertainties and can have an epistemic or aleatoric nature, something that renders fuel performance code-to-code and code-to-measurements comparisons for complex phenomena such as the pellet cladding mechanical interaction (PCMI) very challenging. Additionally, PCMI and other complex phenomena found in fuel performance modeling and simulation induce strong discontinuities and non-linearities that can render difficult to extract meaningful conclusions form uncertainty quantification (UQ) and sensitivity analysis (SA) studies. In this work, we develop and apply a consistent treatment of epistemic and aleatoric uncertainties for both UQ and SA in fuel performance calculations and use historical benchmark-quality measurement data to demonstrate it. More specifically, the developed methodology is applied to the OECD/NEA Multi-physics Pellet Cladding Mechanical Interaction Validation benchmark. A cold ramp test leading to PCMI is modeled. Two measured quantities of interest are considered: the cladding axial elongation during the irradiations and the cladding outer diameter after the cold ramp. The fuel performance code used to perform the simulation is FAST. The developed methodology involves various steps including a Morris screening to decrease the number of uncertain inputs, a nested loop approach for propagating the epistemic and aleatoric sources of uncertainties, and a global SA using Sobol indices. The obtained results indicate that the fuel and cladding thermal conductivities as well as the cladding outer diameter uncertainties are the three inputs having the largest impact on the measured quantities. More importantly, it was found that the epistemic uncertainties can have a significant impact on the measured quantities and can affect the outcome of the global sensitivity analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Performance and cost analysis of liquid fuel production from H 2 and CO 2 based on the Fischer-Tropsch process

Electro-fuels can be produced from concentrated sources of carbon dioxide and hydrogen using electricity generated from renewable sources; this process enables energy storage at high volumetric energy density. Among the electro-fuels options, FT (Fischer-Tropsch) fuel is attractive for heavy-duty trucks and non-road transportation applications. This study conducts a techno-economic analysis of FT liquid fuel production from H 2 and CO 2 using a detailed performance analysis. Minimum fuel selling price is estimated for a broad range of H 2 and CO 2 prices and for a range of potential CO 2 credits. The analysis indicates that H 2 price has the largest impact on the minimum selling price of FT fuel. FT fuel production with a CO 2 price of $17.3/metric ton requires an H 2 price of $0.8/kg to be cost-competitive with the pre-tax petroleum diesel price of $3.1/gal in 2050 (before the application of any CO 2 credits). When the H 2 price is $2.0/kg from central water electrolysis (2020 target), the minimum selling price of the FT fuel is $5.4–5.9/gal. A sensitivity analysis shows that future system optimization of FT fuel production could focus on improving the H 2 and CO 2 recycle contributions and FT fuel conversion ratio. The analysis results can be combined with various upstream systems for H 2 and CO 2 production.

08 HYDROGEN↗
DOE OSTI2019

A Thermal Model for the Performance Analysis of Large Scale Spent Fuel Cooling Pond - 19173

This paper presents a thermal model for the performance analysis of a large scale cooling pond for the long term storage of spent nuclear fuel. The model has been developed using the knowledge gained from fundamental investigation using CFD and has been validated by comparing against reliable experimental and/or numerical data. It was then applied in the context of operation of a large scale pond to ascertain the sensitivity of various operational and environmental conditions. To develop the model, the water body along with the heating elements was considered as one zone and humid air inside the building as another zone, both of which are considered as well-mixed zones. Various input conditions namely, total heat load from the heat source, environmental temperature and relative humidity, ventilation rate, efficiency of the cooling tower, recirculation and make-up water flow rates have been programmed in the model. The statistical method AVONA has been used to assess the influence of the above variables. The results reveal that the efficiency of the cooling tower is the most influential parameter on the cooling performance of cooling pond within the considered values of heat load. It has also been found that the majority of the heat is taken by the recirculating water for very high heat load while the other modes of heat transfer are small but not insignificant. The model is capable of simulating the scenario of loss of cooling water and suggests that it can be used for managing a large pond efficiently as well as recovery from an accident situation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Progress on ATF development of SiC for LWR

Since 2011, Toshiba has been participating in a study whose objective is the commercial utilization of SiC instead of Zircaloy as LWR fuel material. During the period (Phase-1) from 2012 to 2015, feasibility studies on material design and fabrication were performed. This paper reports the principal results of a comprehensive feasibility study on SiC involving experiments and trial fabrication in terms of material properties, core and fuel design, and analytical simulation for safety analysis. The principal results are summarized as follows. First, the fabrication trials for the reduced scales of channel box and cladding have been successfully performed. Secondly, with regard to fuel rod performance analysis, hoop stress during normal operation may be a potential issue in fuel mechanical design as it is relatively high. Thirdly, with regard to safety analysis, the hydrogen generation of SiC fuel tends to decrease significantly compared with that of Zircaloy fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

EVALUATION OF ENHANCED ACCIDENT TOLERANT FUELS

The U.S. Department of Energy Advanced Fuels Campaign (AFC) is currently funding research and development on a number of fuel technologies for light water reactors (LWRs) that could provide enhanced performance under accident scenarios relative to the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the accident tolerant fuel (ATF) research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. Evaluating the performance of candidate accident tolerant fuel concepts is a topic of significant discussion among researchers, fuel vendors and utilities to ensure that the developed concepts can meet minimum performance and economic requirements, can be developed within a reasonable time, and can be approved for insertion as a lead fuel rod in a commercial reactor by the 2022 goal. Characterizing the performance enhancements of candidate ATF first requires understanding the performance of the current Zr-UO2 system under equivalent operations and accident scenarios. Proper evaluation of each concept is dependent on development of data through focused out-of-pile and in-core experiments to support modeling of the fuel and cladding behavior in fuel performance and systems analysis codes; complimentary irradiation of ATF concepts will be discussed in a separate paper. This paper will provide an overview of the key evaluation tools and evaluation scenarios currently being considered within the LWR community for ATF. The evaluation toolset includes standard neutronic and thermal-hydraulic analysis for normal operating conditions and transient/accident conditions and analysis of severe accident behavior using a modified version of the MELCOR code. Development of advanced fuel performance analysis using the BISON application based on the Multi-physics Object-Oriented Simulation Environment (MOOSE) environment at Idaho National Laboratory (INL) has also been initiated. Evaluation scenarios (e.g. accident scenarios) currently being discussed for use across the international ATF development teams are also presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Demonstration of LOTUS multiphysics BEPU analysis framework for LB-LOCA simulations

The LOCA Toolkit for U.S. light water reactors (LOTUS) currently under development at Idaho National Labs (INL) is a plug and play, multiphysics environment to be used in support of system level plant transient analysis. New proposed rule changes in LOCA safety regulations (10 CFR 50.46c) require the inclusion of cladding hydrogen content in the evaluation of equivalent cladding reacted (ECR) and peak cladding temperature (PCT). The potential for a future implementation of said rule change motivates the current development of a unique union of core design, system analysis, and fuel performance codes within a framework allowing for uncertainty quantification (UQ). A demonstration of LOTUS capabilities to address this potential need is performed with an integration of the core design code PHISICS, the fuel performance code FRAPCON, and the system analysis code RELAP5-3D. UQ is performed via Wilks’ method to compute the one sided 95/95 confidence values of the aforementioned safety metrics. Results demonstrate the benefits of the Multiphysics Best Estimate Plus Uncertainty (MP-BEPU) methodology and provide useful visualization of the limiting cases.

10CFR50.46c↗
DOE OSTI2020

Review of the Technical Basis for Properties and Fuel Performance Data Used in HEU to LEU Conversion Analysis for U-10Mo Monolithic Alloy Fuel

This report provides the technical basis for properties and fuel performance data used in conversion analysis for U.S. High Performance Research Reactors (USHPRR) that will convert from highly enriched uranium (HEU) to low-enriched uranium (LEU) using a new U-10Mo monolithic alloy fuel that is being qualified. The conditions that the fuel experiences changes between an HEU and LEU fuel element design due to many causes, including the density of the fuel, the presence of U-238 resonant absorber, changes to the plate and coolant channel dimensions, and changes in fuel management due to reactivity or optimization. These types of changes have been documented in operational and safety analyses conducted for conversions over decades for over 70 reactors. These conversions have all, or almost all, required recalculation of safety-related values as well as establishing operational characteristics of the core, including power distribution, reactor core power level, and cycle length between required fuel management. An overall objective of conversion is to change the reactor core design as little as possible while maintaining the reactors’ scientific, isotope production, medical, and engineering missions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Impacts of Fast-Spectrum Molten Salt Reactor Characteristics on Fuel Cycle Performance

The System Analysis and Integration Campaign nuclear fuel cycle evaluation and screening study was conducted to outline the potential benefits and challenges of nuclear fuel cycle options. The study identified continuous recycle in fast critical reactors as a common characteristic shared among some of the most promising future nuclear fuel cycles. This study was technology agnostic to focus on the underpinning physics that drives fuel cycle outcomes. This process used an analysis example to generate a physics-based understanding of the fuel cycle, and it yielded performance metrics over ranges rather than comparing absolute numbers. While no fast-spectrum molten salt reactor (MSR) has ever been built, several concepts exist using different fuels and carrier salts, many of which target high-burnup objectives, such as lightwater reactor nuclear waste consumption. Putting the resulting fuel cycle impacts of these reactor choices (e.g., fuel and neutron spectrum) into perspective is informative for our current understanding of a given fuel cycle’s performance, and it also demonstrates the applicability of MSRs to different fuel cycles. It has been established that in a fuel cycle transition, performance of a given fast-spectrum MSR depends in part on the initial fuel loading requirements and processing methods.

Betzler, Benjamin↗