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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Fuel Safety Research Plan for Accident Tolerant Fuels

This document outlines a strategic plan for a safety research program for accident tolerant fuels. Accident tolerant fuels (ATFs) are nuclear fuel materials and designs for existing and future light water reactors. This effort is under the Department of Energy Office of Nuclear Energy’s Nuclear Technology Research and Development Advanced Fuels Campaign. The development of ATF is an industry lead initiative with fuel designs being championed by Westinghouse, Framatome, and General Electric. Industry goals for ATF deployment include implementing the first batch reloads of coated cladding and doped fuel designs into commercial reactors by 2023 and having the first batch reloads of more advanced designs into commercial reactors by 2026. Follow on goals include implementing the first full cores of coated cladding and doped fuel designs into commercial reactors by 2026 and implementing full cores of more advanced designs by 2030. The objectives of the national laboratory research programs within the Advanced Fuels Campaign related to ATF are: 1. Develop and maintain of unique nuclear testing infrastructure, which includes test reactors, hot cell PIE equipment, and unique out of pile separate effects testing facilities. 2. Respond directly to industry requests for support by partnering with the ATF vendors and executing experiments and/or analysis on their behalf. 3. Conduct independent scientific investigations into the performance of the different ATF concepts that are relevant to Industry designs. To achieve these objectives, the safety research program will execute transient irradiation experiments at the TREAT facility at Idaho National Laboratory with the aim of identifying the relevant transient fuel behaviors for each fuel design in order to establish appropriate fuel safety criteria and thresholds. Two test series are described in this plan, (1) an Integral Test Series, and (2) a Focused Effects Test Series. In fulfillment of the first objective for the safety research program, Integral Effects Tests will be developed and demonstrated using standard UO2 Zircaloy fuel designs in both the fresh and previously irradiated case. The Integral Test Series will consist of three campaigns to evaluate transient fuel performance in Reactivity Initiated Accidents, Loss of Coolant Accidents, and in generalized Power Cooling Mismatch scenarios. Development of these test programs also enables the safety research program to fulfill its second objective by making these three test programs available to each industry team. Initially, priority will be given in the integral test series to the coated cladding and doped fuel concepts as their deployment schedule is the most aggressive. To fulfill the third objective of the safety research program, a Focused Effects Test Series is also being developed which will examine specific transient fuel behaviors that are of general interest to all the ATF designs under development. These tests look at transient fuel behavior related to pellet cladding mechanical interaction, coated cladding swelling, rupture, and time at temperature behavior, non oxide fuel melting and mechanical interaction with Silicon Carbide cladding, UO2 and doped UO2 fuel fragmentation and transient fission gas release, and finally fuel coolant interaction. This document provides a general overview of the objectives of the safety research program as well as providing background into how programs like this have been used in the past to develop the criteria necessary for the safe operation of the present fuel designs. An overview of the materials and methods used for safety testing is presented. Finally the Integral and Focused Effects tests series are introduced and described. This document will be maintained as a “living” document with annual updates to reflect evolving requirements and needs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Phase-field simulations of fission gas bubbles in high burnup UO2 to inform engineering-scale fuel performance modeling

To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). The cause of pulverization has been hypothesized to be the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel during a LOCA. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory (INL)'s Marmot application for phase-field simulation of nuclear materials. Simulations of the bubble response to steady-state and transient conditions were conducted. Simulation results were used to inform a mechanistic model of pulverization in BISON, INL’s fuel performance simulation code.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2 +/- x: Implications for nuclear fuel performance modeling

Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2x nonstoichiometrywere used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated. 2014 Elsevier B.V. All rights

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Evaluation of B&W UO2/ThO2 VIII experimental core: criticality and thermal disadvantage factor analysis

In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock & Wilcox (B&W) Spectral Shift Control Reactor (SSCR) critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO2-ThO2 fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated.

Criticality↗
DOE OSTI2023

Diffusion in undoped and Cr-doped amorphous UO 2

UO 2 fuel pellets are often doped with chromium oxide to obtain favourable properties such as higher density, improved thermal stability, large grain sizes, improved pellet-clad interaction margins, and increased fission gas retention during transients. Chromium has a low solubility limit in UO 2 , with past experimental work reporting solubility limits ranging between 0.004 to 0.06 wt.% Cr. Due to its low solubility, segregation of Cr ions to the grain boundary may occur. Further, the complexity of these boundaries may be high as observed in other ceramics resulting in disordered or amorphous regions along the boundary, affecting a range of material and operational properties of the fuel pellet. To assess these disordered regions, in this work we study amorphous undoped and Cr doped UO 2 systems (containing 10–50 at.% Cr 3+ ) that have been modelled using classical molecular dynamics methods incorporating Cr 3+ into the well-used CRG potential library. Diffusion coefficients, pre-exponential factors, and activation energies for diffusion were computed for oxygen ions, assessing the impact of structure and extrinsic species on migration. Oxygen diffusion was observed to be much faster in the undoped amorphous system compared to its crystalline counterpart. Oxygen diffusion in doped systems decreased with increasing Cr concentration, highlighting the importance of additives to retain fission products and other migratory species.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Modeling of Cr 2 O 3 -doped UO 2 as a near-term accident tolerant fuel for LWRs using the BISON code

The near-term Accident Tolerant Fuel (ATF) concepts consist of minor modifications of the standard UO 2 -Zr fuel rod materials to provide enhanced accident tolerance. Uranium dioxide fuel with chromia dopant, developed by AREVA, is of interest in view of its enhanced fission gas retention. The predictive capability of the BISON fuel performance code for chromia-doped UO 2 fuel is assessed through simulation of the Halden IFA-677.1 fuel rod experiment. A sensitivity analysis is conducted to interpret the deviation of code’s predictions from the experimental data. A power ramp test is also modeled with BISON and calculations are compared to the experimental database. Finally, a large-break LOCA (LBLOCA) case is simulated with BISON, leading to an initial assessment of the enhanced safety associated with chromia-doped fuel compared to the standard UO 2 fuel in LWRs

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sensitivity of UO 2 fuel performance to microstructural evolutions driven by dilute additives

Use of dilute additives to nuclear fuel is being considered to increase the security of commercial fuel management through traceability of fabricated fuel elements. Taggants, as additives are denoted when included for traceability purposes, may also improve fuel performance, as demonstrated in Cr-containing uranium dioxide as described in the literature, and they may also improve fuel safety. In fact, studies have shown that some additives affect fuel material properties such as grain size and density after sintering. Given the possible range of elements that could be used as additives, the impact of such fuel property variations on the fuel’s thermomechanical behavior becomes relevant. These effects can be evaluated through a sensitivity study of standard fuel models to analyze changes in these properties using a fuel performance code. In this work, the BISON code is being used to investigate these effects through a 2D axisymmetric model of smeared UO 2 fuel pellets and ZIRLO® cladding under realistic pressurized water reactor core irradiation conditions. Here, randomly sampled densities and grain sizes within specified ranges are used as input parameters in the simulations, and several fuel model-related outputs are evaluated. The thermomechanical response of the cladding is also addressed in this study. The simultaneous variation of both input parameters offers a more comprehensive path to identify key sensitivities. Outputs explored include temperature, fission gas release, creep, and radial stress. Results show that although most of these outputs are sensitive to grain size to a certain extent, density mainly affects fuel temperature and elastic strain. Furthermore, sensitivities can vary depending on the radial position within the fuel pellet.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Compare Mechanistic Predictions for Doped UO 2 Mechanical Response and Other Properties with Empirical Models and Experimental Measurements

The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation program develops predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity, high resolution fuel performance tool at the engineering scale. As part of its development, additional modeling capabilities and improvements have been developed for relevant fuel forms. In this work, a fuel creep deformation model for Cr-doped fuel has been implemented into BISON, along with improvements to the empirical UO 2 fuel creep model based on experimental data and improvements to the radial power factor calculation for doped fuels. This work allows for more accurate simulation analyses for both UO 2 and doped-UO 2 fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

In Situ Study of Particle Precipitation in Metal-Doped CeO2 during Thermal Treatment and Ion Irradiation for Emulation of Irradiating Fuels

Metallic particles formed in oxide fuels (e.g., UO 2 ) during neutron irradiation have an adverse impact on fuel performance. A fundamental investigation of particle precipitation is needed to predict the fuel performance and potentially improve fuel designs and operations. In this study, we report on the precipitation of Mo-dominant β-phase particles in polycrystalline CeO 2 (surrogate for UO 2 ) films doped with Mo, Pd, Rh, Ru, and Re (surrogate for Tc). In situ heating scanning transmission electron microscopy indicates that particle precipitation starts at ~1073 K with a limited particle growth to ~10 nm. While particle concentration increases with increasing temperature, particle size remains largely unchanged up to 1273 K. There is a dramatic change in the microstructure following vacuum annealing at 1373 K, probably due to phase transition of reduced cerium oxide. At the high temperature, particles grow up to 75 nm or larger with distinctive facets. The particles are predominantly composed of Mo with a body-centered cubic structure (β phase). An oxide layer was observed after storage at ambient conditions. In situ heating X-ray photoelectron spectroscopy reveals an increasing reduction of Ce charge state from 4+ to 3+ in the doped CeO 2 film at temperatures from 673 to 1273 K. In situ ion irradiation transmission electron microscopy with 2 MeV Al 2+ ions up to a dose of ~20 displacements per atom at nominally room temperature does not lead to precipitation of visible particles. However, irradiation with 1.7 MeV Au 3+ ions to ~10 dpa at 973 K produces ~2 nm sized pure Pd particles; Au 3+ irradiation at 1173 K appears to result in precipitates of ~6 nm in size. Some of the defects produced by ion irradiation could be nucleation sites for precipitation, leading to generation of smaller particles with a higher concentration.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Work updates for Prof. Lee

I want to update my work which I have done so far in INL, including matlab GUI for neutron radiography and some fuel EBSD works.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

AGR-2 TRISO Fuel Post-Irradiation Examination Final Report

An array of post-irradiation exams and experiments were employed to assess the irradiation performance and high-temperature behavior of AGR-2 UCO TRISO fuel produced in an engineering-scale coater. Work was also conducted on UO2 TRISO fuel produced for comparison with the AGR UCO fuel. Emphases we replaced on enumerating the frequency of SiC layer failures (failure of the SiC layer with at least one PyC layer remaining intact) and TRISO failures (failure of all three dense layers) and evaluating the retention of key fission products within the fuel. Examinations of the fuel kernel and coating morphologies were used to understand the features typical of normal intact fuel, degraded or failed fuel, and mechanisms of degradation. Radiochemical analyses were used to determine their retention within the fuel. The results confirm excellent performance of the fuel that exceeds historical HTGR design requirements with significant margin.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

SCCTE: An LEU NTP Concept with Tungsten Cermet Fuel

Presented in this paper is the Space Capable Cryogenic Thermal Engine (SCCTE), a Low Enriched Uranium (LEU) W-UO2 cermet fuel, ZrH1.8 moderated Nuclear Thermal Propulsion (NTP) Concept. The SCCTE project was undertaken by a collaboration with NASA Marshall Space Flight Center (MSFC) and Center for Space Nuclear Research (CSNR) to investigate the applicability of LEU cermet NTP technology for a human Mars mission in the 2030's time frame. This paper focuses on the nuclear design aspects of SCCTE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order to assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure of the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗