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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Preliminary irradiation effect on corrosion resistance of zirconium alloys - 2016-0067

In order to improve the composition and structure of zirconium alloys, it is important to know how changes in their structure-phase state under irradiation affect various properties, including corrosion. In this work, autoclave corrosion tests were carried out on fuel cladding samples made of zirconium-niobium and zirconium-niobium-tin-iron alloy systems. All samples were preliminarily irradiated in a BOR-60 nuclear reactor at a fluence of 2 X10{sup 26} m{sup -2} (E>0.1 MeV). The autoclave tests of irradiated and unirradiated samples were performed at 350 deg. C for 240 days in distilled water containing 10 ppm lithium and 1,600 ppm boron. The irradiation of all alloys resulted in a significant change in their initial structure-phase state. The content of niobium in b-niobium-phase precipitates was reduced from 80 % to - 90 % to 50 % to - 60 %. Iron atoms from the Laves-phase particles were diffused into a matrix, and their lattice transformed from hexagonal close-packed to body-centered cubic. An average size of the transformed second-phase precipitates in the irradiated alloys increased relatively to an average size of the initial precipitates. Elongated radiation-induced precipitates (RIPs) with an average length up to 7 nm were formed in all alloys under irradiation. In comparison with the unirradiated state, irradiation reduced the corrosion of zirconium-niobium alloys, and this effect became more apparent with increased niobium content in the alloy. For zirconium-niobium-tin-iron alloys, irradiation led to more severe corrosion, mostly because of increasing tin content. Niobium content increased up to 2.37 %, and iron content decreased up to 0.18 % in the alloys, contributing to higher corrosion resistance. An interconnection between the niobium/(tin + iron) alloying parameter, quantity of nanometric precipitates formed therein under irradiation, and their corrosion performance has been suggested. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

A Data-driven approach to Core Power distribution reconstruction in a Nuclear Reactor

This report presents the initial development of a data-driven approach for reconstructing the core power distribution in a nuclear reactor (power shape synthesis) using ex-core sensors. Traditional techniques rely on deploying a large number of detectors throughout the reactor core. However, this approach is not feasible for innovative reactor concepts like Advanced Reactors and Microreactors. First, the tight lattice pitch, designed to maximize power density, limits the space available for sensors. Secondly, the harsh operating conditions are not compatible with commercially available detectors. The method proposed in this work integrates high-fidelity modeling with data-driven techniques to accurately reconstruct power distribution across various reactor types, thereby reducing the reliance on in-core sensors. Purdue University Reactor One (PUR-1) was selected as the test case. The CAD model representing the latest configuration of the PUR-1 core was imported into the OpenMC simulation framework, and the model was built. Additionally, the previously developed MCNP6 model was updated. The two models were assessed against the data collected during an experimental campaign conducted in July 2024. Thirty gold foils were placed in three Irradiation Assemblies in PUR-1 core. Using the measured activity of the irradiated foils, the neutron flux at different core locations was estimated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

SCALE Lattice Physics Code Assessments of Accident Tolerant Fuel

This report highlights accident-tolerant fuel (ATF) code assessment activities performed under Project NRCHQ- 60-17-T-0017, Lattice Physics Enhancements and Assessment. ATF covers a broad range of advanced fuel and clad designs for light water reactors (LWRs) to enhance performance under several accident conditions. Several ATF concepts are anticipated to be deployed as lead test rods (LTRs) or lead test assemblies (LTAs) within the next five years. The purpose of this work is to assess the predictive capabilities of NRC neutronics codes that underpin various licensing calculations. ATF designs use different fuel and clad materials compared to standard UO 2 fuel and zirconium-alloy claddings (hereafter UO 2 -Zry). These new materials and geometrical designs need to be assessed to quantify the impact of nuclear data uncertainties on quantities of interest (QOIs) in licensing calculations and the impact of modeling approximations which may be valid for UO 2 -Zry but not for ATF. This report outlines a systematic approach for ATF neutronics code assessment which includes sensitivity and uncertainty (S/U) analysis of nuclear data, identification of experimental benchmark and gaps for code validation, investigation of modeling approximations, and code-to-code comparisons of calculated QOIs against high-fidelity reference continuous energy (CE) Monte Carlo (MC) calculations. This report focuses on the assessment of the SCALE/Polaris lattice physics code for reactor safety analysis. Polaris lattice physics calculations generate few group (FG) cross sections for PARCS full-core calculations. (Full-core analysis, spent fuel analysis, and severe accident analysis will be performed in future work.) The selected ATF concepts for this report include Cr 2 O 3 and Al 2 O 3 -Cr 2 O 3 -doped UO 2 fuel, U 3 Si 2 fuel, FeCrAl cladding, SiC cladding, and Cr-coated cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The Effects of Different Lattice Codes on Results of Sensitivity and Uncertainty Analysis

According to the UAM ('Uncertainty Analysis in Modeling') benchmark, uncertainties in the LWR system calculations at all stages should be determined and quantified. As the neutronics calculations are prerequisite for the predictions of the reactor system, and the uncertainties introduced by the neutronics calculations would be propagated to the subsequent calculations, hence uncertainty analysis has been firstly focused on the neutronics phase to quantify the uncertainties of the reactor-physics predictions in UAM. For the lattice physics, the objective of uncertainty analysis is focused on uncertainty quantifications of the few-group constants; for the core physics, uncertainties of the core steady-state stand-alone neutronics calculations are focused on. Based on the previous researches, the nuclear-data uncertainty has been treated as one of the most significant source of uncertainty and received the most focus. In the previous works, the domestic code UNICORN has the capability of performing sensitivity analysis and uncertainty analysis using the direct numerical perturbation (DNP) method and statistical sampling (SS) method. In this paper, the capability of the UNICORN code has been improved. The domestic lattice code NECP-CACTI has been added in UNICORN to perform the lattice calculations, with DRAGON5.0 and NECP-CACTI available for lattice calculations. For DRAGON5.0 and NECP-CACTI, all the methods and technologies are shared in UNICORN. Therefore, the effects of different lattice codes on the results of sensitivity and uncertainty analysis have been researched and determined in this paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Two-Step Analysis of Watts Bar Nuclear 1 Cycle 1 with SCALE/PARCS

A two-step approach to reactor core analysis was performed using SCALE and PARCS. Polaris is a new lattice physics module in SCALE 6.2 that provides light water reactor (LWR) modelling capabilities. The VERA (Virtual Environment for Reactor Applications) benchmark problems have been established by the Consortium of Advanced Simulation of Light Water Reactors (CASL). These benchmark problems were chosen to assess the performance of Polaris and PARCS. Homogenized few-group assembly cross sections were generated using Polaris and used in the United States Nuclear Regulatory Commission (NRC) core simulator PARCS. This paper will provide descriptions of the benchmark problems modelled, discuss the impact of transport cross section approximations in SCALE 6.2.0 and 6.2.1, and present the results obtained from the SCALE/PARCS benchmark calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

CFD modeling of turbulent air flow in self-heated gyroid TPMS structures: Thermal-hydraulic performance and validation

The application of mathematically derived geometries, such as triply periodic minimal surface (TPMS) lattices, has garnered significant interest across various fields, including the nuclear sector, due to their superior thermal-hydraulic characteristics for heat transfer compared to traditional plain or finned tubes. Here, this study validates a computational fluid dynamics (CFD) model, evaluates different turbulence models and CFD model settings, and performs uncertainty quantification to provide a comprehensive analysis. Despite extensive research on CFD modeling of TPMS lattices, such as gyroid and diamond geometries, there is a notable lack of publicly available literature providing comprehensive details on numerical analysis aspects, including convergence and methodological best practices. This study embarks on a benchmark analysis of a gyroid geometry to evaluate its thermal-hydraulic performance under turbulent flow conditions and scrutinize various CFD model configurations. The main contributions of this work include validating the CFD model, assessing and comparing different turbulence models, and enhancing pressure drop and temperature prediction capabilities. The results aim to support the development of methodologies needed to benchmark and enhance numerical analysis techniques for TPMS lattices. This work seeks to complement the existing body of knowledge, support the development of TPMS reactor concepts, and improve best practices for CFD modeling of TPMS lattices, ultimately advancing methodologies to support future applications in this domain.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Modeling and Simulation of Fuel Dispersal During the Loss-of-Coolant Accident

This document is the compilation of the milestone portion to a larger end of project NEUP report. The executive summary of the modeling portion is provided below: In the event of cladding rupture during a postulated LOCA in a pressurized water reactor, fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod, a phenomenon referred to as fuel dispersal. The initial stage of fuel dispersal is strongly influenced by the high-pressure ejection of fuel fragments, the size and geometry of the ruptured cladding, and the depressurization history of the fuel rod during the postulated LOCA transient. Depending on the location of the burst orifice relative to the quench front, the dispersal event represents an intricate three-phase flow and heat transfer phenomenon, where high-temperature fuel particles carried by the fission gases interact with the coolant within the narrow subchannels of the fuel assemblies, inducing localized phase change. Given the unique multiphysics nature of this phenomena, the current study develops a dedicated computational framework to predict the mass distribution and cooling of dispersing fuel particles, facilitating post-accident assessment and management of the fuel assemblies. Considering the scale of nuclear reactor applications, a continuum three-fluid model is proposed for simulating the transport of solids within the reactor core. With high-temperature fuel fragments within the liquid media, nucleation sites inducing phase changes are dispersed within the flow domain. Coupled with the fact that the transient dispersal event occurs on different time scales than other three-phase flow applications, this study derives a time-averaged three-fluid flow model without losing generality. The assumptions regarding the continuum treatment of the solid phase and the modeling of fuel dispersal behavior are incorporated to simplify the governing equations and derive applicable closure relations. The computational validation of the model was conducted using adiabatic experimental results obtained from ongoing research at Oregon State University, focusing on characterizing fuel dispersal behavior during simulated LOCA conditions. Settlement characteristics of the solids, quantified by the probability distribution of equivalent particles, closely matched the probability density functions reported in experimental studies. The transport of fuel particles within a scaled 5 × 5 lattice of a pressurized-water reactor rod bundle geometry was modeled through a two-fluid Eulerian framework. The required boundary conditions were evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considered solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum exchange between the participating fluids. The simulation results provided the volume fraction of the solids obtained at the bottom surface of the enclosing tank geometry. Postulated LOCA leading to fuel dispersal phenomena involves the strong coupling between fuel thermomechanics, cladding deformation, thermal-hydraulics, and fuel particle transport. Incorporation of such a strong coupling in numerical simulation is performed by coupling the multiphysics solvers. In the case of fuel dispersal, a strong coupled simulation can be performed by coupling the BISON code for fuel performance, the TRACE code for system-level thermal hydraulics, and fuel particle transport in Multiphysics Object-Oriented Simulation Environment (MOOSE). For such intricate infrastructure, the MOOSE Framework eases the data transfer between codes. The recent version of MOOSE has incorporated the Navier-Stokes module for the fluid flow. An exploratory exercise was done to gain familiarity with finite volume capabilities in the MOOSE framework to incorporate the Spalart-Allmaras (SA) turbulence model. New finite-volume and auxiliary kernels were introduced to assemble the SA transport equation, compute turbulent viscosity, and evaluate wall distance and diagnostic turbulence terms, fully integrated with existing Navier-Stokes modules. A turbulent lid-driven cavity at a Reynolds number of approximately 10,000 is used for verification. MOOSE shows the robust solver convergence and produces the turbulent features. But it underpredicts the velocity profile and turbulent quantities, emphasizing the need to develop improved SA near-wall treatments (e.g., low-Re corrections or wall functions) as a key direction for future work.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Application of Variance Reduction Techniques in RMC Burnup and Thermal-Hydraulic Coupled Calculation

In reactor analog simulation, the Monte Carlo method has relatively high fidelity since it can reduce calculation errors resulting from using approximation methods by describing complex geometry and neutron energy spectra. The Monte Carlo method is, in essence, a stochastic simulation of the neutron transport process, and statistical uncertainty of all the computed results must be provided. In large-scale reactor computation, this method can accurately estimate integral quantities such as the effective multiplication factor (K{sub eff}), and it can limit the uncertainty within an acceptable range via reasonably selecting simulation particles. Nevertheless, it faces great challenges in estimating local quantities like neutron flux and fission power, with uncertainty of estimators beyond acceptance. To decrease the uncertainty of local quantities by increasing simulation particles may cost much time and reduce efficiency. The main reasons are as follows. Firstly, the scale of reactors is so large that common pressurized water reactors have tens of thousands of lattice cells; secondly, uneven distribution of neutron flux and fission power causes uneven distribution of uncertainty. Smith K. puts forward the 95/95 principle to estimate whether uncertainty calculated with Monte Carlo Code satisfies the requirement. According to the principle, with a confidence level of 95%, statistical errors of power density in more than 95% of the regions should be less than 1%. To meet the requirement of the principle, the uniform fission site (UFS) method is raised currently. UFS aims at redistributing fission sources to uniformize the fission source sampling, which can make statistical variance distribution more uniform. This method, which is easy to use, will not obviously increase computation time. Many studies show that UFS has a good variance reduction effect in Monte Carlo criticality calculation, but less attention is paid to its performance in burnup calculation and thermal-hydraulic coupled calculation. Uncertainty arising from the Monte Carlo method has a large impact on burnup calculation and thermal-hydraulic coupled calculation. Variance propagates between burnup steps and between iterative steps of coupled calculation, which causes asymmetry of results of the physical symmetrical regions and instability of coupled iterative convergence errors. To further verify the effect of UFS, this study uses it in the RMC (the Reactor Monte Carlo code) burnup calculation module and the neutronics/thermohydraulics coupled calculation module, and conducts two-dimensional pressurized water reactor burnup calculation and BEAVRS benchmark thermalhydraulic coupled calculation. In this way, it verifies the improvement effect of UFS on variance propagation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Uncertainty Analysis in Critical Experiments of SORA Mockup

A series of static critical experiments were completed from September 1965 to August 1966 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) in support of the design of the Sorgente Rapida Reactor (SORA) under a cooperative agreement between EURATOM and the U.S. Atomic Energy Commission. The SORA which had been proposed by EURATOM and was described by Raievski and Larrimore was designed to be composed of rods of highly enriched uranium alloy latticed, as a fast, reflected, repetitively pulsed reactor to serve as a high-intensity source of neutrons for time-of-flight measurements. More information about the SORA design can be obtained in some references. The critical experiments, not including pulse testing, were performed with a mockup of the SORA reactor differing from the reactor principally in the use of iron in place of the coolant, of iron as reflector, and of uranium metal fuel in place of the alloy. A summary of the experiments for the mockup was compiled in report ORNL- 4263, in June, 1968, then published as a paper in 1969. Based on the available data from papers, reports mentioned above and the logbook, the typical 15 critical experiments were evaluated for experimental uncertainties for further benchmark analysis and inclusion in the ICSBEP Handbook. Through a series of rigorous procedures in analysis of the original experimental measurement data, the dimensions and displacement of individual components, and the isotopic contents, mass, and impurities of the individual components, the experimental uncertainties for the 15 configurations have been evaluated to be 0.00219 ∼0.00229 Δk{sub eff} for 11 configurations in type of the Reference Assembly, 0.00179 ∼0.00189 Δk{sub eff} for 4 configurations in type of the All-Iron Assembly. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Core Physics Characteristics of Extended Enrichment and High Burnup Boiling Water Reactor Fuel

This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with 235 U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRSDate not supplied

Lattice Confinement Fusion & Fast Fission for Space

Lattice Confinement Fusion (LCF) 1 • Fusion of hydrogen isotopes into helium isotopes • No tokamak magnets or laser power supplies • Nuclear fuel is confined and triggered within lattice • Nuclear reactions enhanced by electron screening • Extended Electrodynamics (EED) role • NASA published results in Physical Review C 2,3 • Locally hot but globally cold • Commercialized in 2025 to produce medical radioisotopes 4 Lattice Confinement Fusion Fast-Fission Hybrid Reactor • Fusion neutrons fission uranium, spent fuel rods or thorium • No enriched uranium, cleaner fission! • Demonstrated with US Navy and GEC • Supported by NASA STMD (NIAC) and NSF Application • LEO, Deep Space Power and High I sp Propulsion • (Terrestrial)

lattice confinement fusion↗
DOE OSTI2015

Uncertainty Quantification and Management for Multi-scale Nuclear Materials Modeling

Understanding and improving microstructural mechanical stability in metals and alloys is central to the development of high strength and high ductility materials for cladding and cores structures in advanced fast reactors. Design and enhancement of radiation-induced damage tolerant alloys are facilitated by better understanding the connection of various unit processes to collective responses in a multiscale model chain, including: dislocation nucleation, absorption and desorption at interfaces; vacancy production, radiation-induced segregation of Cr and Ni at defect clusters (point defect sinks) in BCC Fe-Cr ferritic/martensitic steels; investigation of interaction of interstitials and vacancies with impurities (V, Nb, Ta, Mo, W, Al, Si, P, S); time evolution of swelling (cluster growth) phenomena of irradiated materials; and energetics and kinetics of dislocation bypass of defects formed by interstitial clustering and formation of prismatic loops, informing statistical models of continuum character with regard to processes of dislocation glide, vacancy agglomeration and swelling, climb and cross slip.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Reactor physics benchmark of Westinghouse PWR core design suite for high burnup/high enrichment fuel - Part 1: Pin and lattice

Westinghouse has performed a comprehensive set of code-to-code benchmark comparisons to corroborate application of its state-of-the-art NEXUS-based code suite to high burnup and high enrichment (HB/HE) fueled PWR core design and reload analysis. This paper focuses on analysis of the predictions for pin and lattice designs representative of HB/HE core reloads from the Westinghouse advanced lattice code PARAGON2, the Monte-Carlo code SERPENT2 and MPACT, the deterministic neutronic component of VERA. A companion paper presented at this conference analyzes results obtained with the Westinghouse ANC9 core simulator, using lattice data provided by PARAGON2 with the NEXUS cross-section methodology, and MPACT for 2D and 3D core configurations incorporating HB/HE fuel reloads to achieve cycle length extension to 24-month in high-power density PWRs. The results of the pin and lattice benchmark presented in this paper show that the neutronic predictions from PARAGON2 for HB/HE fueled PWRs are in remarkable agreement with predictions from both SERPENT2 and MPACT including for depletion to high fuel burnup values, and employing a set of challenging lattices which feature combined use of burnable absorbers to obtain the required reactivity hold-down to support PWR transition to HB/HE fuel and 24-month operational cycles. As corroborated by this analysis and the analysis contained in the companion paper, the Westinghouse NEXUS-based PWR core analysis package can be confidently applied by industry for design analysis of PWR core reloads adopting HB/HE fuel and properly support the industry commercialization plans for in-reactor deployment. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

A study of fission product migration paths in irradiated TRISO particle SiC

Tristructural isotropic (TRISO) coated fuel particles used in high-temperature gas-cooled nuclear reactors (HTGRs) provide key advantages in terms of passive safety and reactor energy efficiencies. TRISO fuel consists of a number of special coating layers surrounding the fissile material in the kernel as shown in Fig. 1. The coating layers consist of a carbon buffer layer, inner pyrolytic carbon (PyC) layer, a silicon carbide (SiC) layer as well as an outer pyrolytic carbon layer. Together the layers act as a pressure vessel and barrier to the release of radioactive fission products into the reactor environment. The cubic 3C-SiC layer of thickness approximately 35 µm, acts as the primary barrier to the migration of metallic fission products out of the particle. The finding, more than three decades years ago, that silver can be released by intact TRISO particles has led to significant research to determine possible transport mechanisms of certain fission products in SiC. Out-of-reactor experiments did not reveal any significant lattice diffusion of silver in SiC1. Many earlier findings alternatively proposed a dominating role of fast diffusion paths such as grain boundaries and dislocations in enhancing fission product migration in SiC. Recently the role of the fission product palladium in enhancing the transport rate of silver in SiC2 was confirmed by transmission electron microscopy (TEM) of neutron-irradiated SiC from a TRISO coated particle from Compact 4-1-1, which was irradiated to an average burnup of 19.38% fissions per initial metal atom (FIMA); a time-averaged, volume-averaged temperature of 1072 °C; a time-averaged peak temperature of 1182 °C; and an average fast fluence of 4.13 x 1021 n/cm2 as part of the US Department of Energy's Advanced Gas Reactor (AGR)-1 experiment3. Although other studies found that Ag was found separately (without Pd) in TRISO coated particles irradiated at 11.4 %, 16.7 and 18.6% FIMA, Pd could have been present at lower than measureable concentrations4. In this study, SiC focussed ion beam (FIB) sections obtained from Compact 4-1-1 from the AGR-1 experiment were analysed in a double Cs-corrected JEOL ARM 200F using atomically resolved scanning transmission electron microscopy (STEM). By using atom resolved STEM in combination with electron energy loss spectroscopy (EELS) and large angle energy dispersive spectroscopy (EDS) a much better understanding of the migration routes and mechanisms of fission products in irradiated polycrystalline 3C-SiC was achieved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Wireless in-core acoustic telemetry and self-powered sensing

A promising wireless telemetry infrastructure based on acoustic transmission of in-pile measurements (information) is being developed at the Idaho National Laboratory for use in nuclear reactors. In this paper, we will demonstrate an approach to telemetry and sensing that exploits a reactor core's energy-rich environment to generate acoustic telemetry signals from self-powered telemeters that produce sounds. These sounds propagate through the reactor's coolant, whether it is gas, water, or liquid metal, through its mechanical structures (e.g., lattice work, pressure vessel, and piping), and can be detected outside the reactor. A thermo-acoustic (TAC) sensor was designed to be geometrically indistinguishable from the other Training Research Isotopes General Atomics (TRIGA) fuel elements in the Breazeale reactor's core. Heat was produced by fission of two Pathfinder {sup 235}UO{sub 2} fuel pellets contained within a finned heat exchanger. The acoustic signals from the TAC sensor can be picked up by hydro-phones and accelerometers anywhere in the reactor pool and from structures that penetrate the pool. Most of the signals we acquired had sufficient signal-to-noise ratios even with both the {sup 16}N diffusion pump and the pool coolant pump operating. The TAC sensor was tested during eight irradiation runs in the Breazeale Nuclear Reactor over the span of a week. We were able to show that the frequency of the thermo-acoustically generated sound provides an accurate determination of the reactor's coolant temperature. Such a sensor strategy might have provided useful information in a reactor accident like that which destroyed the Fukushima complex in March 2011.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Verification & Validation of High-Order Short-Characteristics-Based Deterministic Transport Methodology on Unstructured Grids

The research team has developed a practical, high-order, discrete-ordinates, short characteristics neutron transport code for three-dimensional configurations represented on unstructured tetrahedral grids that can be used for realistic reactor physics applications at both the assembly and core levels. This project will perform a comprehensive verification and validation of this new computational tool against both a continuous-energy Monte Carlo simulation (e.g. MCNP) and experimentally measured data, an essential prerequisite for its deployment in reactor core modeling. Verification is divided into three phases. The team will first conduct spatial mesh and expansion order refinement studies to monitor convergence of the numerical solution to reference solutions. This is quantified by convergence rates that are based on integral error norms computed from the cell-by-cell difference between the code’s numerical solution and its reference counterpart. The latter is either analytic or very fine- mesh numerical solutions from independent computational tools. For the second phase, the team will create a suite of code-independent benchmark configurations to enable testing the theoretical order of accuracy of any particular discretization of the discrete ordinates approximation of the transport equation. For each tested case (i.e. mesh and spatial approximation order), researchers will execute the code and compare the resulting numerical solution to the exact solution on a per cell basis to determine the distribution of the numerical error. The final activity comprises a comparison to continuous-energy Monte Carlo solutions for zero-power critical configuration measurements at Idaho National Laboratory’s Advanced Test Reactor (ATR). Results of this comparison will allow the investigators to distinguish between modeling errors and the above-listed discretization errors introduced by the deterministic method, and to separate the sources of uncertainty.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗